Intermediate-level Waste Disposal Research Articles

Evaluation of the Resistance to Crack Propagation of High-Performance Concretes by the Wedge Splitting Test Method

<p>This paper describes the key elements of the results of investigations on the resistance to crack propagation carried out as part of a large-scale project to develop high-performance concretes (HPCs) for the secondary lining of a low and intermediate-level waste disposal shaft. Four HPCs were investigated, in which the quantity of the binder component and the proportions of the two cements and the silica fume in it were varied. We have also added steel fibers to one HPC. The wedgesplitting test method was used to determine the resistance to crack propagation. The results obtained show that all the HPC investigated achieve good resistance to crack propagation. The addition of steel fibers further improves this resistance.</p>

RESISTANCE TO CRACK PROPAGATION OF HIGH-PERFORMANCE CONCRETE

<jats:p>This paper discusses the results of investigations on the resistance to crack propagation carried out as part of a large-scale project to develop high-performance concretes (HPCs) for the secondary lining of a low and intermediate level waste disposal shaft. Four HPCs were investigated, in which the quantity of the binder component and the proportions of the two cements and the silica fume in it were varied. We have also added steel fibers to one HPC. The wedge splitting test method was used to determine the resistance to crack propagation. The results obtained show that all the HPC investigated achieve good resistance to crack propagation. The addition of steel fibers further improves this resistance.</jats:p>

Deep borehole disposal of intermediate-level waste

Abstract. Around the world, deep borehole disposal is being evaluated for intermediate-level waste (ILW), high-level waste (HLW), spent nuclear fuel (SNF), separated plutonium waste and some very high specific activity fission product waste. In Australia, long-lived ILW from research reactors and radiopharmaceutical production represents the principal waste stream that requires deep geologic disposal. Whilst the Australian government has not yet made a decision on its preferred strategy for ILW disposal, deep borehole disposal of small volumes of ILW would be a more cost-effective and modular solution compared to a conventional geologic disposal facility (GDF). CSIRO, ANSTO and SANDIA have created an international partnership to execute a full-scale borehole research, development and demonstration (RD&D) project in Australia. The project will demonstrate the technical feasibility of the long-term safety of borehole disposal in deep geological formations. The execution of this project could also demonstrate options for nuclear waste disposal that would reduce proliferation risks, potentially up to the termination of compliance with international safeguards requirements. The RD&D includes demonstration of surface handling and waste/seal emplacement capabilities, basic research on foundational science areas, and full-scale field testing in both a deep characterization borehole and a larger-diameter (0.7 m or 27.5 inch) 2000 m deep demonstration borehole. The multi-barrier system designed for such a deep disposal borehole concept places much less reliance on engineered barriers at the disposal zone to achieve safety as compared to a conventional GDF. It rather relies on geological features for waste containment. The concept being explored uses disposal containers with primary waste packages, such as vitrified waste canisters, inside; to be both cost effective and fit for purpose, such a container could have a mild steel-based structural component with copper coating. A critical review of six coating technologies showed that cold spray has the greatest advantages, such as minimal porosity and compressive residual stress. The RD&D has delivered novel enabling tools that assist with site screening, borehole design and post-closure safety assessments. For instance, an automated geological fault mapping and meshing tool was developed that assists with ranking the suitability of potential disposal sites based on proximity to faults. New codes were developed for better representation of fault zones in 2D/3D numerical flow and transport models, while also being more efficient to execute. Post-closure safety assessments tested the sensitivity of long-term safety with respect to disposal depth, rock permeability and sorption. Heat transport calculations explored the sensitivity of temperature evolution within the borehole to parameters such as heat load, borehole depth, geothermal gradients and rock thermal conductivity. For verification of host rock tightness while also demonstrating the absence of recent groundwater, a new noble gas analytical facility has been established for measuring rare noble gases in mineral fluid inclusions as indicators of very old pore fluids.

Long-Term Experiments for Demonstrating Durability of a Concrete Barrier and Gas Generation in a Low-and Intermediate-Level Waste Disposal Facility

Long-Term Experiments for Demonstrating Durability of a Concrete Barrier and Gas Generation in a Low-and Intermediate-Level Waste Disposal Facility

Complexation of Sn with Boom Clay natural organic matter under nuclear waste repository conditions

126Sn is a long-lived fission product and as such a waste relevant radioisotope. Understanding its fate under deep-geological repository conditions is a requirement for a complete safety assessment. Yet, due to a limited knowledge of its hydrogeochemical properties, the solubility and speciation of tin in a nuclear waste repository remain uncertain. In Belgium, Boom Clay (BC) is investigated as a potential host rock for high-level and intermediate level waste disposal. It contains a significant amount of dissolved organic matter (DOM), which was shown to strongly influence the mobility of several radionuclides and notably of tetravalent actinides. By chemical analogy, the speciation, sorption and mobility of Sn(IV) in BC is also assumed to be controlled by its binding to DOM. However, the Sn-DOM complexation has never been quantified under relevant repository conditions. In this work, the complexation of Sn with BC DOM was investigated under present-day BC conditions and in synthetic seawater as representative of more saline conditions. The ultrafiltration technique was used to quantify the amount of Sn bound to DOM. The two investigated repository conditions yielded nearly identical binding constants, evidencing a low influence of salinity on the level of binding of Sn to BC DOM. BC DOM forms strong complexes with Sn that would significantly increase Sn solubility in BC pore water. The investigated range of Sn and DOM concentrations (DOMtot/Sntot = 2.8 × 102 to 5.7 × 103) also showed that the binding constants are dependent on the metal loading (5.0 < logK(Sn−DOM) < 6.0). A two-site Langmuir isotherm was used to describe this dependency and highlighted the binding of Sn on both strong and weak sites present on DOM. Overall, the experimental data show that the binding of Sn with BC DOM controls the Sn speciation under BC repository conditions, which reveals the importance of taking it into account in safety assessment calculations.

Modelling Long-Term Durability Performance of Cementitious Materials under Sodium Sulphate Interaction

Cementitious materials are one of the essential components for low- and intermediate-level waste disposal sites. Low-level nuclear waste from power plants consists of highly concentrated (~25 wt %) Na2SO4, and the wastes are solidified with cementitious materials. Degradation of cementitious materials that result from chemical and physical sulphate attack is a major concern in the safety of the waste disposal. In this study, hydration and reactive transport models, developed in previous works by the authors, were applied with Pitzer interactions coefficients to evaluate the long-term performance of Portland cement (PC) solidified with high concentration of Na2SO4. Expansive sulphate-bearing products of ettringite and mirabilite were formed and filled the pores in the hydrating PC with 25% of Na2SO4 by weight, but they were destabilised as temperature increased. Influence of Na2SO4 concentration and temperature on mineralogical changes is discussed. The simulation results from the reactive-transport model showed that the degradation of solidified Na2SO4 waste by cementitious materials exposed to 10% Na2SO4 for 1000 years is due to dissolution of mirabilite and secondary formation of ettringite, but not Na2SO4 crystallisation. The phases and porosity became stable close to exposure surface after 10 years, although the deterioration progressed from the surface to core with exposure time.

Preliminary analysis of gaseous radiocarbon behavior in a geological repository hosted in salt rock

ABSTRACTA preliminary evaluation of gaseous radiocarbon (14C) behavior under geological repository conditions for Italian radioactive high level waste-long-lived and intermediate level waste disposal has been performed. Although in Italy there is still no defined project for a geological disposal facility, current work may support future safety assessment studies for a hypothetical future repository in deep salt rock. In the Italian context of radioactive waste, the percentage of 14C bearing waste to be disposed in a possible geological repository is low; irradiated graphite is the most important radiological source. Data about the radiological inventory has been collected to simulate production and migration of gaseous 14C in a hypothetical geological repository. Three different conceptual models have been developed and simulated. The first model has considered a preliminary evaluation of the radiological impact referred to the whole inventory; the second and third model have evaluated the impact only due to the irradiated graphite. A preliminary sensitivity analysis has been carried out, highlighting the importance of geometry and of distribution coefficients (Kd) in materials used to seal the disposal underground facility. Results show the possibility to correlate the Kd values, the volume and the location of the sealing materials to the amount of 14C migrating toward the surface.

Evaluation of efficient cementation of fission 99Mo production process waste and feasibility disposal of cement waste form

The Korea Atomic Energy Research Institute (KAERI) is planning the construction of the KIJANG Research Reactor (KJRR), which is currently undergoing licensing procedures. The fission 99Mo production process (FMPP) of the KJRR generates intermediate-level liquid waste (ILLW) and alumina residue (Al2O3 cake), which is a solid waste. In the present study, an efficient method of reducing the amount of generated cement waste form by mixing the ILLW and Al2O3 cake together, into low-level waste, was developed. The purpose of this study was to determine if the cement waste form produced by mixing cement with the waste from the FMPP would satisfy the acceptance criteria of a low- and intermediate-level waste disposal facility in Korea. The process operating range and the optimum operating conditions were found by investigating the mixability of the cement waste form, by considering the presence of free standing water after curing, and by performing a water immersion test depending on the mixing ratios of Portland cement (type I), ILLW, and Al2O3 cake. The feasibility of treating the cement waste form produced under the optimum operating conditions was evaluated in relation to the acceptance criteria of a low- and intermediate-level waste disposal facility in Korea. The results showed that the cement waste form satisfied the acceptance criteria with regard to structural stability (compressive strength test, thermal cycling test, water immersion test, and irradiation test) and leachability (leaching test).

Large scale in-situ experiments on sealing constructions in underground disposal facilities for radioactive wastes – Examples of recent BfS- and GRS-activities

This paper shows examples of in-situ constructions and laboratory tests that have been designed and implemented by BfS and GRS in order to demonstrate the technical feasibility of special constructions for the sealing of drifts and shafts in different salt formations.Since it is often not possible to directly verify the long-term functionality (for times scales envisaged in long-term safety assessments) of such constructions, an overall understanding of the main chemical and physical processes involved needs to be developed. Such an understanding is required in order to extrapolate the evolution of a sealing system with a sufficient degree of reliability. Laboratory tests and large-scale in-situ tests are necessary and integral parts of the procedures for enhancing the process understanding.Based on a safety case, BfS has developed a robust decommissioning concept for the closure of the low- and intermediate-level waste disposal facility Morsleben (ERAM) which also takes into account the retention of radionuclides by combined effects of different sealing components.GRS has contributed to this concept with laboratory tests and modelling of the long-term behaviour of the sealing material “salt concrete” foreseen for the sealing constructions in Morsleben. Salt concrete is based on the cement Portlandite i.e., on CaO. Furthermore GRS has also investigated the MgO based “sorel concrete” employed by BfS for the Asse project. The chemical corrosion path and the subsequent permeability changes for both concretes have been investigated with test procedures developed in the GRS laboratory. GRS has also developed a new Self-Sealing material “Selbst Verheilender Versatz – SVV” (SVV, German for Self Sealing Backfill) and tested in laboratory and in-situ experiments. With this material it is possible to achieve an instantaneous and long-lasting permeability reduction within the complex system consisting of the sealing construction and the Excavation Demaged Zone (EDZ) upon brine intrusion.The presentation demonstrates details of the concepts and highlights results from the BfS in-situ experiments in Morsleben and the preceding GRS laboratory and in-situ tests.

보조지표를 활용한 중·저준위 처분시설 성능평가: 방사성 핵종 플럭스 사례연구

The use of complimentary indicators, other than radiation dose and risk, to assess the safety of radioactive waste disposal has been discussed in a number of publications for providing the reasonable assurance of disposal safety and convincing the public audience. In this study, the radionuclide flux was selected as performance indicator to appraise the performance of engineered barriers and natural barrier in the Wolsong low- and intermediate-level waste disposal facility. Radionuclide flux showing the retention capability by each compartment of the disposal system is independent of assumptions in biosphere model and exposure pathways. The scenario considered as the normal scenario of disposal facility has been divided into intact or degraded silo concrete conditions. In the intact silo concrete, the radionuclide flux has been assessed with respect to the radionuclide retardation performance of each engineered barrier. In the degraded silo concrete, the radionuclide flux has been explored based on the performance degradation of engineered barriers and the relative significance of natural barrier quantitatively. The results can be used to optimally design the near-surface disposal facility being planned as the second project phase. In the future, additional complimentary indicators will be employed for strengthening the safety case for improving the public acceptance of low- and intermediate-level waste disposal facility.

Experiment on gas entry pressure and gas permeability of concrete silo for a low- and intermediate-level waste disposal facility in Korea

After closure of an underground disposal facility, it would be saturated with ground water and gases would be generated by various mechanisms. As the generated gases are accumulated, internal pressure of the concrete silo would increase and integrity of the silo could be damaged by overpressure. Therefore, an effective gas permeable seal is necessary to prevent the overpressurization of the concrete silo. For designing the gas permeable seal, experimental study and computer modeling on the gas migration through the concrete should be performed. In this study, in order to obtain the specific characteristics of the concrete silo which are necessary for computer modeling, gas entry pressure and gas permeability of the concrete silo were measured by using specimens which have a same composition with the concrete silo. The results of this study could be utilized to the simulation of gas migration and design of the gas permeable seal for a LILW disposal facility in Korea.

Gas threshold pressure and gas permeability of silo concrete specimens for a low- and intermediate-level waste disposal facility in Korea

In order to simulate gas migration at an underground disposal facility, the characteristics of a medium such as the gas threshold pressure and gas permeability need to be measured in advance. In this study, the gas threshold pressure and gas permeability of silo concrete specimens for a Korean LILW (Low- and Intermediate-Level Waste) disposal facility were measured. The concrete specimens had the same composition as the concrete used in the construction of the silo. The gas threshold pressure was measured by injecting a constant gas flow into cross sections of the specimens. To measure the gas permeability, selected pressures were applied to the specimens and the apparent permeability was calculated using the Hagen-Poiseuille equation. The intrinsic permeability was calculated with the Klinkenberg empirical equation. The gas threshold pressure and gas permeability ranged from 30 to 40bar and from 10−17 to 10−18m2, respectively.

Modeling of concrete carbonation in deep geological disposal of intermediate level waste

Simulations of atmospheric carbonation of Intermediate-Level Long-lived radioactive Waste (ILLW) concrete packages were conducted to evaluate their possible chemical degradations. Two-phase liquid water-air flow is combined with gas component diffusion processes leading to a progressive drying of the concrete.Complete drying of the 11 cm thick waste disposal package wall occurs over a period ranging from 2 years for the low-performance concrete to 10 years for the high-performance concrete. The drying process slows down when transport characteristics of concretes are enhanced. Carbonation depths in the order of 2 to 3 cm in 100 years are predicted for this cementitious component. However, these values are slightly overestimated compared to experimental data. Also the kinetic model of mineral reactivity requires improvements with respect to the protective effect of secondary carbonates and to thermodynamic data.

The potential impact of anaerobic microbial metabolism during the geological disposal of intermediate-level waste

AbstractMicrobial metabolism has the potential to control the biogeochemistry of redox-active radionuclides in a range of geodisposal scenarios. In this study, sediments from a high pH lime workings site were incubated under carefully controlled anaerobic conditions, at a range of alkali pH values with added electron donors and electron acceptors, to explore the limits and rates of bioreduction in a sediment system analogous to intermediate-level nuclear waste. There was a clear succession in the utilization of electron acceptors (in the order nitrate &gt; Fe(III)-citrate &gt; Fe(III) oxyhydroxide &gt; sulfate), in accordance with calculated free energy yields and Eh values over the pH range 10–12. The rate and extent of bioreduction decreased at higher pH, with an upper limit for the processes studied at pH 12. The biochemical limits for such processes are discussed, alongside the potential impact of such forms of microbial metabolism on the solubility of a range of redox active radionuclides that will feature heavily in the safety case for the geological disposal of intermediate-level nuclear waste.

Long term leachate evolution during flow-through leaching of a vault backfill (NRVB)

AbstractSome of the illustrative concepts for the disposal of intermediate-level waste in a geological disposal facility in the UK employ a cementitious backfill around the waste packages. The concept for higher strength rocks would use a highly alkaline backfill composed of Portland cement (now known as CEM I), hydrated lime and limestone flour, referred to as Nirex reference vault backfill (NRVB).This paper reports a study of the extensive leaching of cured NRVB in a range of generic leachant compositions (deionized water, 0.1 M and 1 M NaCl solutions) under flow-through conditions using a flexible wall permeameter. The experiments were designed to run for up to two years and to pass at least 1000 volumes of leachant (defined as the cumulative leachate volume produced/NRVB solid specimen volume) through the NRVB samples. Results for the pH evolution profiles of the leachates and the microstructural analysis of the unleached and leached samples are presented.

Radiological and geochemical characteristics of an ultramafic massif (NE Portugal) regarding the site aptness to host a near surface repository for low and intermediate level radwaste

The search for suitable sites to install near-surface repositories for low to intermediate level radioactive wastes has increased in the recent years, reflecting both the continuous rise in radioactive materials applications by the society and the increasing awareness of public perception regarding radwaste management. International recommendations towards the application of higher safety standards to low and intermediate level waste disposal sites being developed by the International Atomic Energy Agency (IAEA), and the European Commission decision that each Member State must safely dispose its own radwaste in its territory, according to national radwaste management plans, has leading to new soundly scientific studies. In Portugal, possible locations for that purpose are under study since 2006, following the pre-requisites indicated by IAEA. One of these sites overlaps a significant portion of an ultramafic massif, corresponding to an allochthonous and metamorphosed ophiolite complex. In the surveyed area of this massif, the prevailing rock type consists of (serpentinised) peridotites, occasionally containing irregular bodies of (amphibolitised) gabbros, over which an irregular/ill-developed regolith exists. The activities of 40K, 137Cs, 210Pb, 226Ra and 232Th measured in samples representing three consecutive levels of the regolith upper portion in 13 stations confirm the heterogeneity put in evidence through in situ γ-ray surveys previously performed. The available dataset also demonstrates that radionuclide contents decrease prominently with depth, becoming closer to the bedrock values (up to ca. 41 Bq kg−1 for 40K in flaser-gabbros and near zero for all the remaining cases) but never reaching them. Therefore, the low/moderate radionuclide activity in the regolith upper portion reflect the contribution of sources that are independent of the lithological background; i.e. they indicate external and miscellaneous contributions to the “soil”-forming components, most of them coming from (dry or wet) atmospheric deposition. The chemical composition of the (serpentinised) peridotites and (amphibolitised) gabbros in the surveyed area, as well as their alteration products included in regolith, is homogeneous enough to be used as a faithful reference in monitoring studies of any kind. In addition, the significant 137Cs activities measured in some top “soil” samples (up to 92 Bq kg−1) strongly suggest that there are mineralogical specificities of the peridotite-derived regolith able to fix that radionuclide. This particular feature reinforces the existing criteria that favour the peridotites of Morais as a suitable host for a near surface repository of low to intermediate level radwaste.

Ageing management program for the Spanish low and intermediate level waste disposal and spent fuel and high-level waste centralised storage facilities

The generic design of the centralised spent fuel storage facility was approved by the Spanish Safety Authority in 2006. The planned operational life is 60 years, while the design service life is 100 years. Durability studies and surveillance of the behaviour have been considered from the initial design steps, taking into account the accessibility limitations and temperatures involved. The paper presents an overview of the ageing management program set in support of the Performance Assessment and Safety Review of El Cabril low and intermediate level waste (LILW) disposal facility. Based on the experience gained for LILW, ENRESA has developed a preliminary definition of the Ageing Management Plan for the Centralised Interim Storage Facility of spent Fuel and High Level Waste (HLW), which addresses the behaviour of spent fuel, its retrievability, the confinement system and the reinforced concrete structure. It includes tests plans and surveillance design considerations, based on the El Cabril LILW disposal facility.

Detritiation of the JET Carbon Tiles by Flame Heating

Operation with tritium plasma led to contamination of the JET in-vessel components with tritium at a level exceeding 12kBq/g, which is the Low Level Waste (LLW) threshold in the UK. Carbon tiles used at JET for protecting the pumped divertor and inner wall against heat and neutron flux create one of the Intermediate Level Waste (ILW) streams to deal with during the JET decommissioning in the future. To reduce quantity and cost of ILW disposal from JET, the study has been initiated for development of detritiation techniques. This paper presents a brief description of the study of the JET carbon tiles detritiation using flame heating.

Assessment of Redox Conditions in the Near Field of Nuclear Waste Repositories: Application to the Swiss high-level and intermediate level waste disposal concept

ABSTRACTRedox conditions were assessed for a spent fuel and high-level waste (SF/HLW) and an intermediate-level waste (ILW) repository. For both cases our analysis indicates permanently reducing conditions after a relatively short oxic period. The canister-bentonite near field in the HLW case displays a high redox buffering capacity because of expected high activity of dissolved and surface-bound Fe(II). This is contrary to the cementitious near field in the ILW case where concentrations of dissolved reduced species are low and redox reactions occur primarily via solid phase transformation processes.For the bentonite-canister near field, redox potentials of about -100 to -300 mV (SHE) are estimated, which is supported by recent kinetic data on U, Tc and Se interaction with reduced iron systems. For the cementitious near field, redox potentials of about -200 to -800 mV are estimated, which reflects the large uncertainties related to this alkaline environment.

Large-Scale Model Testing of a Concrete Silo EBS for Low- and Intermediate-Level Waste Disposal

ABSTRACTThis study focuses on demonstrating the feasibility of constructing a large reinforced concrete silo for the disposal of low- and intermediate-level radioactive waste. In a large-scale model test, the results of temperature (from heat of hydration) and thermal stress analyses were confirmed, as was the practicality of countermeasures to prevent thermal cracking. After one year, the average permeability of the model silo was estimated to be below 1×10−11m/s.