Low-level Radioactive Solid Waste Research Articles

Modeling Cs-137 Migration in Groundwater due to Low Probability Event of Leaching from Near Surface Disposal Facility for Low Level Solid Radioactive Waste

In this paper, we present impact of Cesium-137 (Cs-137) disposal in a Near Surface Disposal Facility (NSDF) with respect to groundwater migration and its consumption for drinking. We employ the Multiple Area Source Model for simulating the fate of radionuclide in groundwater to evaluate its potential contamination due to extremely low probability event of leaching from the NSDF. The assessment conservatively assumes direct interaction between disposal trenches and the underlying aquifer neglecting the barrier provided by the unsaturated zone. Temporal profiles of concentration of Cs-137 at different down flow distances in groundwater are obtained over a period extending several years post-disposal. The results indicate that even under such a worst-case conservative scenario and for an annual disposal rate of 3 TBq/y for a period of 50y, the maximum Cs-137 concentration in groundwater remains significantly below the safe drinking water threshold of 10 Bq/L recommended by the World Health Organization. It is found that amount of Cs-137 generally disposed in NSDF is significantly lower than the allowable disposal limit corresponding to concentration defined for safe drinking water limit.

Site-specific characterisation of hydrogeological parameters for low-level solid radioactive waste disposal facility at Tarapur, India

Site-specific characterisation of hydrogeological parameters for low-level solid radioactive waste disposal facility at Tarapur, India

Safety verification tests of low and intermediate level radioactive solid waste drum

Steel drums or boxes are used to transport, store, and dispose of low- and intermediate-level waste. According to relevant standards, before the solid waste container design is finalized, a series of type tests should be performed, including leakage test, water spray test, stacking test, penetration test, free-drop test, etc. This article focuses on the acceptance test and test methods for the LID- II b type waste drum finalization test. The tests in accordance with the standard requirements are the first time reported in China. It found the inherent design defects, and had high guiding significance to improve design and test verification operation.

Activation studies for the decommissioning of PET cyclotron bunkers by means of Monte Carlo simulations

During the operational life of a PET cyclotron, the concrete walls of the vault are activated by secondary neutrons. For the dismantling of such accelerator facilities, a considerable amount of low level solid radioactive waste has to be characterized and disposed. To decrease future dismantling costs and complexity, the amount of radioactive waste has to been prospectively evaluated in the design phase, then confirmed at the time of planning decommissioning. In this work, the Monte Carlo code Fluka was used for the assessment of the activation of the bunkers of two different facilities: the 16.5 MeV GE PETtrace at S. Orsola-Malpighi Hospital in Bologna and the 18 MeV IBA Cyclone 18/18 HC at the Bern University Hospital (Inselspital). The simulations were validated by means of experimental measurements performed in our previous works: non-destructive, in field measurements using a portable CZT detector were performed in Bologna; while core drilling samples were extracted from the bunker and measured in laboratory with an HPGe detector in Bern. The activity of the most important radionuclides in the concrete walls of the bunker, namely Eu-152, Mn-54, Co-60, Sc-46, Zn-65 and Cs-134 resulted within the range of 0.01 – 2 Bq/g. The consistency between Monte Carlo results and experimental measurements was within a factor 2 - 3 for most radionuclides, except for Eu-152, Sc-46, Zn-65. The activity concentrations estimated at each position considered exceeds the clearance levels of the new Directive 2013/59/Euratom.The results of this work demonstrate that Monte Carlo simulations based on FLUKA are adequate to assess the residual activation levels, a fundamental information to foresee, plan and optimize the decommissioning of a cyclotron based PET centers.

Leaching behaviors and mechanisms of vitrified forms for the low-level radioactive solid wastes

Leaching behaviors and mechanisms of commercialized glass wasteforms to sequester low-level solid-wastes were investigated: SG glass for resin waste and DG-2 glass for dry active waste. After ANS 16.1 leaching test, leachabilities of the nuclides, Co, Cs, and Sr, were all lager than 14, which met the requirement of the US-Nuclear Regulatory Commission. Holes of diameters 5–10 μm remained on the surface of the SG and crevices of lengths 10–50 μm were observed on the surface of the DG-2. We analyzed elemental compositions of the SG and the DG-2 with depths. For the SG, Si, Al, Ca, and Mg were accumulated and Na was depleted up to nearly 1.5 μm compared to an internal glass. For the DG-2, concentrations of B, Na, Al, Ca and Sr started to decrease from 2.5 μm even though other minor elements are still remained their concentrations. We suggested leaching mechanisms: alkali elements including H would diffuse through the holes on the SG, while most of the elements including Si and Al would diffuse through the crevices on the DG-2.

Activation studies of a PET cyclotron bunker

During the operational life of a PET Cyclotron, the concrete walls of its vault are activated by secondary neutrons. For this reason when considering dismantling of such accelerators, a considerable amount of low level solid radioactive waste has to be characterized and disposed. To decrease future dismantling costs, the amount of radioactive waste has to be evaluated in advance to define an optimum decommissioning strategy. Aim of this work is to develop methodologies for the preliminary assessment of activation induced in a bunker via experimental measurements. Two different facilities were studied and different approaches were adopted: non-destructive, in field measurements using a portable CZT detector were performed in the bunker of a 16.5 MeV GE PETtrace (S. Orsola-Malpighi Hospital of Bologna); while core drilling samples were extracted from the bunker of an 18 MeV IBA Cyclone 18 H C (Inselspital in Bern) and measured in laboratory with an HPGe detector. The principal long-lived radionuclides found in concrete were Eu-152, Mn-54, Co-60, Sc-46, Cs-134 with activity concentrations comprised between 0.01 and 0.6 Bq/g. In all performed measurements the total activity concentration exceed the clearance level of Directive (2013)/59/Euratom. Considering a hypothetical decommissiong of the two facilities, the amount of low lever radioactive waste estimated was about 93 m3 in the case of S. Orsola-Malpighi, while about 58 m3 for Insespital facility. Both the methodologies have proved to be adequate for the purposes.

Composite materials with primary lead slag content: Application in gamma radiation shielding and waste encapsulation fields

This paper presents a study on the valorization of primary lead slag as substituent of aggregate in concrete for gamma radiation shielding and as a component of cementitious binders with application in low level radioactive solid waste encapsulation. The first part of the paper outlines the impact of primary lead slag on the compressive strength and gamma radiation shielding properties of concrete. The presence of the primary lead slag as aggregate substituent in concrete, along with steel shot and barite, enhances the density and shielding efficiency against gamma radiation for Ir192, Cs137, Co60 radioactive elements with average gamma energies of 0.37 MeV, 0.662 MeV and 1.25 MeV. The results showed that even the presence of primary lead slag led to a decrease in compressive strength and all the tested concretes exhibited compressive strengths higher than 50 MPa. The utilization of primary lead slag could be recommended as aggregate in heavyweight concrete. The second part of the paper focuses on the experimental investigation of the physic-mechanical properties and low level radioactive solid waste encapsulation capacity of grout based on binders with primary lead slag content. Two binders (30% and 50% lead slag content) were used in cementitious grouts with different binder:sand ratio (1:1, 1:1.5 and 1:2), two types of sand (river sand or crushed concrete sand), different grain size distribution of sand and water/binder ratio of 0.3. It was observed that the grouts exhibit compressive strengths higher than 30 MPa, regardless of the binder:sand ratio. The increase in the binder:sand ratio influenced negatively the flow ability and positively the drying shrinkage. The low level radioactive solid waste encapsulation capacity of studied grout was lower than that of the reference grout with CRT glass waste.

Regeneration of pilot-scale ion exchange columns for hexavalent chromium removal

Due to stricter regulations, some drinking water utilities must implement additional treatment processes to meet potable water standards for hexavalent chromium (Cr(VI)), such as the California limit of 10 μg/L. Strong base anion exchange is effective for Cr(VI) removal, but efficient resin regeneration and waste minimization are important for operational, economic and environmental considerations. This study compared multiple regeneration methods on pilot-scale columns on the basis of regeneration efficiency, waste production and salt usage. A conventional 1-Stage regeneration using 2 N sodium chloride (NaCl) was compared to 1) a 2-Stage process with 0.2 N NaCl followed by 2 N NaCl and 2) a mixed regenerant solution with 2 N NaCl and 0.2 N sodium bicarbonate. All methods eluted similar cumulative amounts of chromium with 2 N NaCl. The 2-Stage process eluted an additional 20–30% of chromium in the 0.2 N fraction, but total resin capacity is unaffected if this fraction is recycled to the ion exchange headworks. The 2-Stage approach selectively eluted bicarbonate and sulfate with 0.2 N NaCl before regeneration using 2 N NaCl. Regeneration approach impacted the elution efficiency of both uranium and vanadium. Regeneration without co-eluting sulfate and bicarbonate led to incomplete uranium elution and potential formation of insoluble uranium hydroxides that could lead to long-term resin fouling, decreased capacity and render the resin a low-level radioactive solid waste. Partial vanadium elution occurred during regeneration due to co-eluting sulfate suppressing vanadium release. Waste production and salt usage were comparable for the 1- and 2-Stage regeneration processes with similar operational setpoints with respect to chromium or nitrate elution.

Fate of Radium in Marcellus Shale Flowback Water Impoundments and Assessment of Associated Health Risks.

Natural gas extraction from Marcellus Shale generates large quantities of flowback water that contain high levels of salinity, heavy metals, and naturally occurring radioactive material (NORM). This water is typically stored in centralized storage impoundments or tanks prior to reuse, treatment or disposal. The fate of Ra-226, which is the dominant NORM component in flowback water, in three centralized storage impoundments in southwestern Pennsylvania was investigated during a 2.5-year period. Field sampling revealed that Ra-226 concentration in these storage facilities depends on the management strategy but is generally increasing during the reuse of flowback water for hydraulic fracturing. In addition, Ra-226 is enriched in the bottom solids (e.g., impoundment sludge), where it increased from less than 10 pCi/g for fresh sludge to several hundred pCi/g for aged sludge. A combination of sequential extraction procedure (SEP) and chemical composition analysis of impoundment sludge revealed that Barite is the main carrier of Ra-226 in the sludge. Toxicity characteristic leaching procedure (TCLP) (EPA Method 1311) was used to assess the leaching behavior of Ra-226 in the impoundment sludge and its implications for waste management strategies for this low-level radioactive solid waste. Radiation exposure for on-site workers calculated using the RESRAD model showed that the radiation dose equivalent for the baseline conditions was well below the NRC limit for the general public.

Determination of 63Ni in the low level solid radioactive waste

Nickel is a constituent of wide range of con- struction materials in nuclear industry and their corre- sponding activation products are often encountered in low level solid radioactive waste and effluents. The article presents the radiochemical analytical method for determi- nation of 63 Ni in the concrete samples originated from Nuclear Power Plant Mochovce, Slovakia. The sample of concrete was ashed and leached with aqua regia thereafter hydroxides were precipitated with ammonia solution, leaving Ni in aqueous phase. The next purification steps included a precipitation as Ni-dimethylglyoxime (DMG) complex, an anionic exchange column and an extraction chromatography on Nickel resin. The purified 63 Ni fraction was mixed with the scintillation cocktail and the mixture was measured by liquid scintillation counting (LSC). The radiochemical yield of the separation procedure was determined by atomic absorption measurement of stable Ni in the range of 49-71%. Up to the sample amount of 50 g the detection limit was 0.002 Bq g -1 at counting time 180 min and 95% confidence level.

Carbon-14 analysis in solidified product of non-metallic solid waste by a combination of alkaline fusion and gaseous CO2 trapping

In order to establish a simple and rapid analytical method for (14)C in solidified products made from non-metallic low-level radioactive solid wastes such as concrete, mortar and glass by melting treatment, a radiochemical analysis in combination with alkaline fusion as a sample decomposition method was examined. A simulated solidified product containing (14)C, which was prepared by using nuclear reaction (14)N(n, p)(14)C with thermal neutron irradiation, was analyzed by the present method to compare with a conventional radiochemical analysis using oxidizing combustion. The reproducible and quantitative recovery of (14)C from the simulated solidified product indicates that the present method is more efficient for (14)C analysis in solidified products than the conventional method using oxidizing combustion.

Mineralogy of selected geological deposits from the United Kingdom and the Republic of Ireland as possible capping material for low-level radioactive waste disposal facilities

Solid low-level radioactive waste (LLW) is currently being disposed at a number of facilities in the United Kingdom (UK). The safety of these facilities relies to some extent on the use of engineered barriers, such as a cap and a liner, to isolate the waste and protect the environment. Generally, the material used as the barrier layer within such a cap should be of low permeability and it should retain this property over long timescales (beyond a few decades normally required for facilities containing non-radioactive wastes). The objective of this research is to determine the mineralogy of selected geological deposits from the UK and Republic of Ireland as part of a larger project to examine their suitability as a capping material, particularly on LLW sites. Mineral transformations, from weathering, may impact on the long-term performance of the cap and even the disposal facility. These transformations could be brought on or enhanced by climate change. X-ray diffraction (XRD) was carried-out on the sand, silt and clay fractions of the London Clay, Belfast Upper Boulder Clay, Irish Glacial Till, Belfast Sleech, and Ampthill Clay geological deposits. Minerals were present that could pose both positive and negative effects on the long-term performance of the cap. Smectite, which has a high shrink swell potential, may produce cracks in London Clay, Belfast Upper Boulder Clay and Ampthill Clay capping material during dry, hotter periods as a possible consequence of future climate change; thus, resulting in higher permeability. Ampthill Clay and Belfast Sleech had elevated amounts of organic matter (OM) at 5.93% and 5.88%, respectively, which may also contribute to cracking. Over time, this OM may decompose and result in increased permeability. Gypsum [CaSO 4] in the silt and sand fractions of Ampthill Clay may reduce the impact of erosion during wetter periods if it is incorporated into the upper portion of the cap. There are potential negative effects from the acidity created by the weathering of pyrite [FeS 2] present in the silt and sand fractions of Belfast Sleech and Ampthill Clay that could impede the growth of grasses used to stabilize the surface of the capping material if this material is used as part of the vegetative soil layer. Additionally, acidic waters generated from pyrite weathering could negatively impact the lower lying capping material and the disposal facility in general, because they could react with contaminants and weaken vessels storing contaminants. However, calcium carbonate [CaCO 3] present in the silt and sand fractions of these deposits, and dolomite [CaMg(CO 3) 2] in Belfast Sleech, may counter act the acidity.

Experience in Progressing the Planning Application for New LLW Disposal Facilities for Dounreay

AbstractAn integral part of decommissioning the Nuclear Decommissioning Agency's (NDA) Dounreay site is the management of the solid low level radioactive waste (LLW). The United Kingdom Atomic Energy Authority (UKAEA) has developed and progressed a technical and stakeholder programme that has enabled it to submit a robust Planning Application to Highland Council (HC) for New LLW Disposal Facilities at Dounreay and to submit substantive preliminary safety and environmental cases to the Nuclear Installations Inspectorate (NII) and the Scottish Environment Protection Agency (SEPA). To UKAEA’s knowledge this is the most advanced project in the UK for new LLW disposal facilities. Experience has been gained in progressing the Best Practicable Environmental Option (BPEO) Study, working with regulators in unfamiliar areas, and undertaking groundbreaking stakeholder consultation. Key lessons learnt are that stakeholders should be engaged in dialogue on the project as early as possible, documentation must be high quality and tailored to its audience, and internationally respected and credible consultants must be involved.

Trial Operation of the Advanced Volume Reduction Facilities for LLW at JAEA

The Japan Atomic Energy Agency (JAEA) constructed the Advanced Volume Reduction Facilities (AVRF), in which volume reduction techniques are applied and achieved high volume reduction ratio, homogenization and stabilization by means of melting or super compaction processes for low level radioactive solid wastes. It will be able to produce waste packages for final disposal and to reduce the volume of stored wastes by operating the AVRF. The AVRF consist of the Waste Size Reduction and Storage Facilities (WSRSF) and the Waste Volume Reduction Facilities (WVRF); the former has cutting installations for large size wastes and the latter has melting units and a super compactor. Cutting installations in the WSRSF have been operating since July 1999. Radioactive wastes treated so far amount to 750m3 and the volume reduction ratio is from 1.7 to 3.7. The WVRF has been operating with non-radioactive wastes since February 2003 for the training and the homogeneity investigation in the melting processes. The operation of the pretreatment system in the WVRF with radioactive wastes has partly started in FY2005.

Trial Operation of the Advanced Volume Reduction Facilities for LLW at JAEA

The Japan Atomic Energy Agency (JAEA) constructed the Advanced Volume Reduction Facilities (AVRF), in which volume reduction techniques are applied and achieved high volume reduction ratio, homogenization and stabilization by means of melting or super compaction processes for low level radioactive solid wastes. It will be able to produce waste packages for final disposal and to reduce the volume of stored wastes by operating the AVRF. The AVRF consist of the Waste Size Reduction and Storage Facilities (WSRSF) and the Waste Volume Reduction Facilities (WVRF); the former has cutting installations for large size wastes and the latter has melting units and a super compactor. Cutting installations in the WSRSF have been operating since July 1999. Radioactive wastes treated so far amount to 750m3 and the volume reduction ratio is from 1.7 to 3.7. The WVRF has been operating with non-radioactive wastes since February 2003 for the training and the homogeneity investigation in the melting processes. The operation of the pretreatment system in the WVRF with radioactive wastes has partly started in FY2005.

Estimation of vertical diffusivity from seasonal fluctuations in groundwater pressures in deep boreholes near Sellafield, NW England

Abstract In 1998, as part of the process to assess the suitability of a site near Sellafield, Cumbria, for a deep geological repository for solid intermediate level and some solid low-level radioactive waste, United Kingdom Nirex Limited (Nirex) commissioned a study into the seasonal fluctuations in groundwater pressure observed in many of the monitoring boreholes in the area. Many of the monitoring zones in the deep boreholes at Sellafield show some response in groundwater pressure to annual variations in recharge. These seasonal fluctuations were quantified in terms of amplitude and lag over two full recharge cycles (1994–1996). The extremely detailed monitoring array installed by Nirex at the potential repository zone gives a unique opportunity to observe in detail the attenuation of the recharge signal, as it is propagated vertically downwards through the Sherwood Sandstone Group and into the underlying basement rocks. Use of an analytical approach to model the data provides constraints on values of the vertical diffusivity of the strata. The values of hydraulic formation parameters derived by this methodology are broadly consistent with results from borehole testing, though somewhat higher. This may in part be due to the large scale associated with the cyclic recharge signal. The attenuation of the seasonal fluctuations in three dimensions (3D) throughout the study area provides information at a scale suitable for use in constraining regional flow models.

High-resolution stratigraphic modeling of the vadose zone at the Savannah River Site low-level radioactive waste trenches disposal facility

The principal disposal facility for solid, hazardous, and mixed low-level radioactive wastes is located in the central part of the Savannah River Site. These wastes are generally disposed of in engineered, unlined trenches and excavations. The trenches vary in length and width but are typically constructed in the upper 7–8 m (21–24 ft) of the local sediments and in the vadose zone. Water from the surface must pass through the waste trenches and unsaturated vadose zone to reach the saturated water table and may contribute to groundwater contamination. The vadose zone underlying the study area extends to depths between approximately 20 and 24 m (65 and 78 ft). To comply with regulatory requirements, an investigation program was developed to characterize the vadose zone underlying two active disposal trench areas. A high-resolution vadose zone lithostratigraphic model was required before the installation and placement of experimental equipment to measure soil hydraulic properties. A detailed mathematical model of flow and transport could then use the stratigraphic model and the water flux information from the instrumentation, for example, to constrain a finite-element grid and arrive at the contaminant flow characteristics in the vadose interval. The two high-resolution stratigraphic models were developed and interpreted from field and laboratory data collected from continuous split-spoon borings, Shelby tube soil samples, piezocone penetration soundings using cone penetrometer technology (CPT), and borehole geophysical logs from nearby shallow monitoring wells. Exposed stratigraphy in open trenches and nearby cuts was also used. These models revealed (1) detailed lithostratigraphic and soil moisture conditions in the vadose zone, including three variations in depositional character roughly corresponding to previously defined formational contacts; (2) that indications that the lateral extent of many vadose zone bedforms is on the order of 20 m (66 ft) or less, therefore requiring vadose zone sampling points to be placed on spacings of less than 20 m (66 ft) for modeling; and (3) that once correlation and modeling have been performed, it is possible to use CPT data, matched to the model, to precisely place new sensors and instrument.

A Comparison of Site Characterisation Data and Modelling Results from a Radiological Assessment of the Drigg Low Level Radioactive Waste Disposal Site

ABSTRACTThe biogeochemical processes controlling release of radionuclides from solid low level radioactive waste (LLW) disposed at the Drigg site have been represented in a computational model in order to determine the effects of varying major element chemical parameters and solubility limits of radionuclides. The model is used to provide a source term for radiological safety assessment calculations for the groundwater pathway. Site data has been compared to the results for the initial stages of the model in order to build confidence in the conceptual and mathematical model used to derive the source term for 234U, the key dose contributor, and other radionuclides which are influenced by the biogeochemical processes. Measured aqueous concentrations of radionuclides whose release from LLW is sorption controlled are compared to the model results as a basis to discuss the uncertainty in sorption input parameters used.

Development of the Post-Closure Safety Case for the Low Level Waste Disposal Site at Drigg, United Kingdom

ABSTRACTThis 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.

Interpretation of pumping tests in the Sherwood Sandstone Group, Sellafield, Cumbria, UK

In 1995 as part of the process to assess the suitability of a site near Sellafield, Cumbria, for a deep geological repository for solid intermediate level and some solid low level radioactive waste, United Kingdom Nirex Limited (Nirex) commissioned a series of long duration pumping tests in the various stratigraphic horizons at the site. Two of the pumping tests were conducted in vertically adjacent sections of the lower Sherwood Sandstone Group (SSG) whilst pressures were monitored at discrete intervals in surrounding monitoring boreholes. Responses were observed in over twenty monitoring zones at various levels around the pumped sections during both tests. In order to arrive at an understanding of the hydraulic system in the SSG, a systematic methodology was adopted to analyse the data sets from the pumping tests. Preliminary analysis of the data from each responding zone used the standard 2D radial flow analytical approach. This indicated that, whilst responses conformed broadly to this model, there were important deviations. Three different analytical modifications were adopted to assess the significance of these deviations; a partial penetration model, a bounded aquifer model and a generalized radial flow model. The use of the different models allowed a conceptual model of flow processes occurring in the SSG to be developed.