Intermediate Level Radioactive Waste Research Articles

Quantification of dissolved organic14C-containing compounds by accelerator mass spectrometry in a corrosion experiment with irradiated steel

ABSTRACTLow- and intermediate-level (L/ILW) radioactive waste produced in Switzerland consists of large amounts of14C-containing irradiated steel.14C will be released during the anoxic corrosion of the steel in the cementitious near field of an L/ILW repository. In this study, a corrosion experiment with irradiated steel was carried out to determine the speciation of14C released during the corrosion process in conditions similar to those anticipated in the near field of a cement-based repository. The development of the experimental setup, including installation of the reactor and development of suitable analytical methods based on compound-specific14C analysis with accelerator mass spectrometry (CSRA AMS) is reported. Time-dependent increase in the total content of14C-bearing organic compounds in solution (TO14C) was determined by AMS and the main organic corrosion products that are14C-bearing formate, acetate and lactate were identified by CSRA AMS after a pre-concentration step. The concentration of the14C-bearing organic compounds was found to be very low (fmol to pmol14C/L). Stable carbon compounds were identified and quantified while the source of stable carbon in the system has not yet been identified and the temporal evolution of the concentration of these carbon species is presently not understood.

Incorporation of strontium and europium in crystals of α-calcium isosaccharinate.

The final repository for short-lived, low and intermediate level radioactive waste in Sweden is built to act as a passive repository. Already within a few years after closure water will penetrate the repository and conditions of high alkalinity (pH 10.5-13.5) and low temperature (< 7 °C) will prevail. The mobility of radionuclides in the repository is dependent on the radionuclides distribution between solid and liquid phases. In the present work the incorporation of strontium (II) and europium (III) in α-calcium isosaccharinate (ISA) under alkaline conditions (pH ∼10) at 5 °C and 50 °C have been studied. The results show that strontium and europium are incorporated into α-Ca(ISA)2 when crystallized both at 5 °C and 50 °C. Europium is incorporated to a greater extent than strontium. The highest incorporation of europium and strontium at 5 °C rendered the phase compositions Ca0.986Eu0.014(ISA)2 (2.4% of Eu(ISA)3 by mass) and Ca0.98Sr0.02(ISA)2 (2.2% of Sr(ISA)2 by mass). XPS spectra show that both trivalent and divalent Eu coexist in the Eu incorporated samples. Strontium ions were found to retard the elongated growth of the Ca(ISA)2 crystals. The incorporation of Sr2+ and Eu3+ into the solid phase of Ca(ISA)2 is expected to contribute to a decreased mobility of these ions in the repository.

Influence of formulation parameters of cement based materials towards gas production under gamma irradiation

The release of radiolysis gas is a concern that may restrict the use of cement materials to condition intermediate level radioactive waste. Indeed, water naturally present in cement materials produces hydrogen gas (which can be explosive/flammable under some conditions) when it is exposed to ionizing radiation.The primary goal of the MATRICE (MAterials Resistant to Irradiation based on Cement) project is to identify and define formulations of cement materials in order to minimize the quantities of hydrogen gas released by radiolysis. The first approach is the minimization of water amount in standard Portland materials (calcium silicate-based cements) by addition of specific compounds (superplasticizers) to enable the preparation of wasteform. The second approach is to use “alternative” cement such as calcium sulfoaluminate cement. This cement was expected to release less hydrogen because the quantity of water needed for cement hydration is higher than Portland and moreover, their hydrates differ from those of hydrated calcium silicate mostly encountered in Portland based materials.Based on gamma irradiations with a60Co source, the results obtained demonstrate that the first approach is efficient but yet limited because the production of hydrogen of Portland pastes is about proportional to the total amount of water present in the materials. Thus, a tremendous drop of hydrogen production cannot be reach because rheological constraint does not allow a huge reduction of water, even with efficient superplasticizers. The second approach using calcium sulfoaluminate cements as an alternative binder provides results that are quite similar to Portland cement concerning the production of hydrogen under gamma irradiation.

Radiotoxicity flux and concentration as complementary safety indicators for the safety assessment of a rock-cavern type LILW repository

This study presents a practical application of complementary safety indicators, which can be applied in a safety assessment of a radioactive waste repository by excluding a biosphere simulation and comparing the artificial radiation originating from the repository with the background natural radiation. Complementary safety indicators (radiotoxicity flux from geosphere and radiotoxicity concentration in seawater) were applied in the safety assessment of a rock-cavern type low and intermediate level radioactive waste (LILW) repository in the Republic of Korea. The natural radionuclide (40K, 226,228Ra, 232Th, and 234,235,238U) concentrations in the groundwater and seawater at the Gyeongju LILW repository site were measured. Based on the analyzed concentrations of natural radionuclides, the levels of natural radiation were determined to be 8.6 × 10−5 – 8.0 × 10−4 Sv/m2/yr and 6.95 × 10−5 Sv/m3 for radiotoxicity flux from the geosphere and radiotoxicity concentration in seawater, respectively. From simulation results obtained using a Goldsim-based safety assessment model, it was determined that the radiotoxicity of radionuclides released from the repository is lower than that of the natural radionuclides inherently present in the natural waters. The applicability of the complementary safety indicators to the safety case was discussed with regard to reduction of the uncertainty associated with biosphere simulations, and communication with the public.

Thermodynamic description of the plutonium – α–d–isosaccharinic acid system ii: Formation of quaternary Ca(II)–Pu(IV)–OH–ISA complexes

The quaternary system Ca(II)–Pu(III/IV)–OH–ISA was investigated in a series of solubility experiments where equilibrium conditions were reached from undersaturation conditions with PuO2(ncr,hyd) at T = (22 ± 2) °C under Ar atmosphere. Experiments were performed in 0.1 m NaCl–NaOH–NaISA–CaCl2 solutions at 8 ≤ pHm ≤ 12.5 (pHm = –log [H+], in molal units), 10−5 m ≤ m(ISA)tot ≤ 10−2 m and 10−3.5 m ≤ m(Ca)tot ≤ 0.02 m. Reducing conditions were adjusted and buffered with hydroquinone (HQ) or Sn(II), which allowed us to work in the (pe + pHm) region ≈ 9 and ≈ 1.5, respectively.Solid phase characterization (in-situ XRD, XANES and EXAFS) and analogy with the ternary system Pu–OH–ISA previously studied by our group indicate that PuO2(ncr,hyd) remains unaltered in the course of the solubility experiments at pHm ≤ 12 and controls the solubility of Pu under reducing conditions in the presence of ISA and Ca(II). In HQ-systems, the impact of ISA on the solubility of PuO2(ncr,hyd) is clearly enhanced compared to Ca(II)-free solutions of analogous composition, thus indicating the formation of quaternary complexes of the type Ca(II)–Pu(IV)–OH–ISA in the aqueous phase. The combination of slope analysis of the solubility data and DFT calculations support the predominance of the two aqueous quaternary complexes Ca(II)Pu(IV)(OH)3ISA–H+ and Ca(II)Pu(IV)(OH)3ISA–2H(aq) below and above pHm ≈ 11, respectively. The same speciation scheme explains properly the solubility of Pu in Sn(II)-buffered systems at pHm > 11. Below this pHm, solubility data under these very reducing conditions suggest the possible formation of ternary Pu(III)–OH–ISA aqueous complexes, although the scatter of the data does not allow a definitive conclusion to be drawn. The presence of Ca(II) in solution destabilizes Pu–ISA colloids, which significantly enhance Pu solubility in comparable, Ca-free systems. Based on these experimental and theoretical results, a thermodynamic model for the system Ca2+–Pu4+–OH––Cl––ISA––H2O(l) is derived.This study provides quantitative tools to evaluate the solubility behavior of Pu in reducing cementitious environments containing ISA, as expected in underground repositories for low- and intermediate-level radioactive waste (L/ILW) due to the degradation of cellulosic materials.

Np(V) sorption and solubility in high pH calcite systems

Np(V) behaviour in alkaline, calcite containing systems was studied over a range of neptunium concentrations (1.62 × 10−3 μM–1.62 μM) in two synthetic, high pH, cement leachates under a CO2 controlled atmosphere. The cement leachates were representative of conditions expected in an older (pH 10.5, Ca2+) and younger (pH 13.3, Na+, K+, Ca2+) cementitious geological disposal facility. These systems were studied using a combination of batch sorption and solubility experiments, X-ray absorption spectroscopy, and geochemical modelling to describe Np behaviour. Np(V) solubility in calcite equilibrated old and young cement leachates (OCL and YCL) was 9.7 and 0.084 μM, respectively. In the OCL system, this was consistent with a Np(V)O2OH(am) phase controlling solubility. However, this phase did not explain the very low Np(V) solubility observed in the YCL system. This inconsistency was explored further with a range of pH 13.3 solubility experiments with and variable Ca2+(aq) concentrations. These experiments showed that at pH 13.3, Np(V) solubility decreased with increasing Ca2+ concentration confirming that Ca2+ was a critical control on Np solubility in the YCL systems. X-ray absorption near-edge structure spectroscopy on the precipitate from the 42.2 μM Np(V) experiment confirmed that a Np(V) dioxygenyl species was dominant. This was supported by both geochemical and extended X-ray absorption fine structure data, which suggested a calcium containing Np(V) hydroxide phase was controlling solubility. In YCL systems, sorption of Np(V) to calcite was observed across a range of Np concentrations and solid to solution ratios. A combination of both surface complexation and/or precipitation was likely responsible for the observed Np(V) reaction with calcite in these systems. In the OCL sorption experiments, Np(V) sorption to calcite across a range of Np concentrations was dependent on the solid to solution ratio which is consistent with the formation of a mono-nuclear surface complex. All systems demonstrated slow sorption kinetics, with reaction times of weeks needed to reach apparent equilibrium. This could be explained by slow recrystallisation of the calcite surface and/or the presence of Np(V) colloidal species. Overall, these data provide valuable new insights into Np(V) and actinide(V) behaviour in alkaline conditions of relevance to the disposal of intermediate level radioactive wastes.

The biogeochemical fate of nickel during microbial ISA degradation; implications for nuclear waste disposal

Intermediate level radioactive waste (ILW) generally contains a heterogeneous range of organic and inorganic materials, of which some are encapsulated in cement. Of particular concern are cellulosic waste items, which will chemically degrade under the conditions predicted during waste disposal, forming significant quantities of isosaccharinic acid (ISA), a strongly chelating ligand. ISA therefore has the potential to increase the mobility of a wide range of radionuclides via complex formation, including Ni-63 and Ni-59. Although ISA is known to be metabolized by anaerobic microorganisms, the biodegradation of metal-ISA complexes remains unexplored. This study investigates the fate of a Ni-ISA complex in Fe(III)-reducing enrichment cultures at neutral pH, representative of a microbial community in the subsurface. After initial sorption of Ni onto Fe(III)oxyhydroxides, microbial ISA biodegradation resulted in >90% removal of the remaining Ni from solution when present at 0.1 mM, whereas higher concentrations of Ni proved toxic. The microbial consortium associated with ISA degradation was dominated by close relatives to Clostridia and Geobacter species. Nickel was preferentially immobilized with trace amounts of biogenic amorphous iron sulfides. This study highlights the potential for microbial activity to help remove chelating agents and radionuclides from the groundwater in the subsurface geosphere surrounding a geodisposal facility.

An Experimental Study on Treatment of Typical Low and Intermediate Level Radioactive Wastes with Thermal Plasma Melting Technology

With the active development of the nuclear industry in China, it is extremely urgent to study and develop advanced technologies for the radioactive waste treatment and disposal. With features of high temperature, high energy density and broad applicability, the thermal plasma melting technology has been considered as one of the key subjects in the radioactive waste treatment research field. A pilot scale experimental study was carried out on the treatment of typical low and intermediate level radioactive wastes (LILWs) from the nuclear power plant in China with thermal plasma melting technology. Two representative wastes, thermal insulation waste and cotton, were selected and melted in the plasma melting furnance at 1250°C for ~ 100 minutes until they were completely vitrified. Meanwhile, tracers of Co2O3, CsCl and SrCO3 were added to the original wastes to simulate the containment effect of 58Co and 60Co, 134Cs and 137Cs as well as 90Sr in the vitrification during the treatment process. The XRD patterns of both vitrification samples displayed as the typical amorphous state. Performances of resultant glass waste-forms, including the physical property, the leachability and the mechanical capacity, were in accordance with the standard requirements of the hign level radioactive waste glass vitrificaton, and were better than those of the cement solidified waste. Further experimental work is in need to better understand the radioactive nuclide migration regularity and to improve the performance of the plasma melting system.

PRELIMINARY DESIGN OF POTENTIAL STORAGE FACILITY FOR LOW AND INTERMEDIATE LEVEL RADIOACTIVE WASTE

An issue concerning the management of radioactive waste (RAW) in the Republic of Croatia continuously rises due to the lack of systematic solutions. In this paper, emphasis is being placed on the storage of RAW, as one of the important phases of RAW management. The current RAW storage facilities are not sufficient and there is a need for a new facility for low and intermediate level radioactive waste (LILW), since such waste accounts for the largest part of RAW. For this reason, a preliminary estimate for a one-storey structure was carried out, a potential storage facility of LILW that would be constructed at the location of 􀀬erkezovac, Trgovska Gora, in the Republic of Croatia, as the currently most prominent location for the realization of the RAW storage facility. The geological characteristics of the site were analysed and, in accordance with the relevant regulations, a static design of the facility was made in the ETABS and SAFE program packages. The construction is a typical reinforced concrete structure, rectangular in plan (10.0 x 30.0 m) with a height of 5.9 m. The selected location meets the given parameters of geological characteristics, while the construction meets the given load and serviceability conditions de􀆤 ned by Eurocode (ultimate limit state and serviceability limit state). However, a more detailed analysis would be necessary if the decision to construct such a storage facility would be made.

Study of the effect of Kaolin in the mortar of cement matrices by confinement of ion exchange resins

Radioactive waste arising as a result of nuclear activities should be safely managed from its generation to final disposal in an appropriate conditioned form to reduce the risk of radiation exposure of technical personnel and of the public and to limit contamination of the environment. The immobilization of low and intermediate level radioactive wastes in cementitious matrices is the most commonly used technique to produce inexpensive waste matrix that complies with regulatory requirements in order to protect humans and the environment against nuisance caused by ionizing radiation. Cement based materials are used in radioactive waste management to produce stable waste forms. This matrix constitutes the first build engineering barrier in disposal facilities. In this work, the kaolin is used to enhance the mechanical performance of the matrix of confinement of ion exchange resins by gradually replacing the sand in mortar with kaolin clay. The Kaolin clay sample was a special pure product, sourced from a foreign country. The maximum quantity of resins that can be incorporated into the mortar formulation without the packages losing their strength is 13.915% which results in a better mechanical strength at 6.7686 MPA compression with kaolin.

Uncertainty Management on Human Intrusion Scenario Assessment of the Near Surface Disposal Facility for Low and Intermediate-Level Radioactive Waste: Comparative Analysis of RESRAD and GENII

In order to manage the uncertainty about the evaluation and analysis of the human intrusion scenario of the Gyeongju Low and Intermediate Level Radioactive Waste(LILW) disposal facility, the calculation result by the GENII code was assessed using the RESRAD code, which was developed to evaluate the radiation effects of contaminated soil. The post-drilling scenario was selected as a human intrusion scenario into the near-surface disposal facility to analyze the uncertainty of the modeling by identifying any limitations in the simulation of each code and comparing the evaluation results under the same input data conditions. The results revealed a difference in the migration of some nuclides between the codes, but confirmed that the dose trends at the end of the post-closure control period were similar for all exposure pathways. Based on the results of the dose evaluation predicted by RESRAD, sensitivity analysis on the input factors was performed and major input factors were derived. The uncertainty of the modeling results and the input factors were analyzed and the reliability of the safety evaluation results was confirmed. The results of this study can be applied to the implementation ‘Safety Case Program’ for the Gyeongju LILW disposal facility.

The Status and Experiences of LILW Disposal Facility Construction

This paper presents the status and experiences of Low and Intermediate-level radioactive Waste(LILW) disposal facility in Korea. The LILW disposal facility of 800,000 drums (200 L/drum) is being constructed in the site of KORAD's LILW Disposal Center(206 m2, Gyeongju-si). The first stage of the LILW facility construction was completed in June 2014, and approved for use by Nuclear Safety and Security Commission (NSSC) in December 2014. The first stage of the LILW facility is a type of underground silo with disposal capacity of 35,200 m3 (100,000 drums). The second stage of the LILW facility is under construction to be completed by December 2020. Site construction is currently underway. The second stage of the LILW facility is a type of near-surface disposal with disposal capacity of 32,875 m3 (125,000 drums)

Impact of an Alkaline Solution on the Chemistry, Mineralogy, and Sorption Properties of a Typic Rhodudult Soil

Abstract The preferred option for disposal of short-lived low and intermediate level radioactive wastes is a near surface disposal facility in which soil is one of the barriers that avoid radionuclide migration outside the controlled area. For construction of that kind of facility, concrete is widely used, and its interaction with water induces its degradation, resulting in a high pH solution. The alkaline solution may affect the near-field environment of radioactive waste repositories, including the soil, promoting mineralogical alterations that [...]

Assessment of strontium transport in clay medium by using a reactive transport modelling approach – an HP1 case study model

Modelling of radionuclide transport in the geosphere is a key issue of the evaluation of safety of low- and intermediate-level radioactive waste (LILW) repositories. In Bulgaria, due to the operation of Kozloduy Nuclear Power Plant (NPP), a building of such facility is ongoing. The migration of 90Sr was evaluated by simulation of flow and chemical modelling, performed with coupled numerical code HP1. The adsorption properties of the clays were implemented in the code by surface complexation reactions between strontium and iron oxides. The results were compared with another model performed with computer code HYDRUS 1-D, in which the adsorption properties of the medium were evaluated with the relevant for strontium distribution coefficient, Kd.

Geotechnical parameters of loess-cement mixture for construction of compacted soil-cement cushion

Soil-cement cushions are compacted and stabilized layers of the soil base, built under the foundation. Usually, they are constructed with local soil from the excavation, mixed with Portland cement. In Bulgaria, this soil improvement technique has been applied in foundation works in collapsible loess ground, aiming to replace a part of the collapsible layer, to increase the bearing capacity of the soil base, and/or to play a role of engineering barrier against migration of harmful substances in the geoenvironment. A multi-barrier near-surface short-lived low- and intermediate-level radioactive waste repository is under construction in Bulgaria. A loess-cement cushion beneath repository cells is going to be built by in-situ compacted mixture of local loess and Portland cement. Based on the results from classification and physico-mechanical tests of a set of loess-cement mixtures, it was proposed optimum cement content of the loess-cement cushion beneath the radioactive waste repository to be 5% of Portland cement. The present paper aims to assess the following geotechnical parameters of the selected loess-cement mixture after proper curing: unconfined compressive and flexural strength; shear strength parameters; static and dynamic elastic constants; and hydraulic conductivity. The results obtained prove that the mixture prepared at Wopt and ρds of local loess and 5% (by the dry weight of soil) of Portland cement type CEM I 42.5 N – SR 5 possesses strength and deformation characteristics that completely meet the design stress-strain requirements to the soil-cement cushion beneath the repository foundation.

Determination of Radionuclide Concentration Limit for Low and Intermediate-Level Radioactive Waste Disposal Facility II: Application of Optimization Methodology for Underground Silo Type Disposal Facility

2014년 12월 사용 승인된 경주 중저준위 방사성폐기물 동굴처분시설은 중저준위 방사성폐기물의 처분을 위해 운영중이나 중준위 방사성폐기물을 처분할 수 없다. 왜나하면 기존 중준위 방사성폐기물이 원자력안전위원회 고시 2014-003호에 따라 방사성폐기물 준위가 세분화되었으며, 기존의 중저준위 방사성폐기물 핵종별 처분농도제한치 값이 변경되었으나 이를 고려 하지 못하였기 때문이다. 중준위 방사성폐기물의 안전한 처분을 위해 IAEA에서 제시한 방법론과는 달리 방사능량 산출 시 적용된 가용데이터를 기반으로 기존의 설정된 극저준위 및 저준위 방사성폐기물의 처분농도제한치를 고려하여 1단계 동굴 처분시설의 중준위 방사성폐기물에 대한 처분농도제한치를 설정하였다. 단, 14C의 경우 처분농도제한치 외에 추가적인 방사 능량 제한이 필요함을 확인하고 우물이용시나리오를 통해 1단계 동굴처분시설의 총방사능량을 제한하였다. 설정된 중준위 방사성폐기물 처분농도제한치와 14C의 총방사능량이 적용된 방사능량에 대해 운영 중 및 폐쇄 후 시나리오의 평가결과가 모 두 성능목표치를 만족함을 확인하여, 도출된 중준위 방사성폐기물 처분농도제한치가 1단계 동굴처분시설의 중준위 방사성 폐기물 처분농도제한치로 사용할 수 있음을 확인하였다. 처분 안전성 증진을 위해 방사성폐기물 발생기관의 데이터를 추가 확보하며, 14C의 누적방사능량을 관리해 나갈 계획이다.

Determination of Radionuclide Concentration Limit for Low and Intermediate-level Radioactive Waste Disposal Facility I : Application of IAEA Methodology for Underground Silo Type Disposal Facility

Determination of Radionuclide Concentration Limit for Low and Intermediate-level Radioactive Waste Disposal Facility I : Application of IAEA Methodology for Underground Silo Type Disposal Facility

폐밀봉선원 처분시스템 예비 폐쇄후 안전성평가

An optimum disposal plan of disused sealed radioactive sources (DSRSs) should be established to ensure long-term disposal safety at the low- and intermediate-level radioactive waste (LILW) disposal facility in Gyeongju. In this study, an optimum disposal system was suggested and preliminary post-closure safety assessment was performed. The DSRSs disposal system was composed of a rock cavern and near surface disposal facilities at the Gyeongju LILW disposal facility. The assessment was conducted using GoldSim program, and probabilistic assessment and sensitivity analysis were implemented to evaluate the uncertainties in the input parameters of natural barriers. Deterministic and probabilistic calculations indicated that the maximum dose was below the regulatory limits (<TEX>$0.1mSvyr^{-1}$</TEX> for the normal scenario, <TEX>$1mSvyr^{-1}$</TEX> for the well scenario). It was concluded that the DSRSs disposal system would maintain environmental safety over a long-time. Moreover, the partition coefficient of Np in host rock, Darcy velocity in host rock, and density of the host rock were the most sensitive parameters in predicting exposure dose in the safety assessment.

Impact of alkaline alterations to a Brazilian soil on cesium retention under low temperature conditions

To be used as backfilling materials in radioactive waste disposal facilities, a natural material must have a suitable permeability, mechanical properties and a high sorption capacity for radionuclides. Also important when considering a material as a backfill is the effect of its interaction with the alkaline solution generated from concrete degradation. This solution promotes mineralogical alterations that result in significant changes in the material key properties influencing its performance as a safety component of the repository. This paper presents results of an investigation on the effect of alkaline interaction under a low temperature on cesium retention properties of a local soil being considered suitable as a backfill for the Brazilian near surface disposal facility. A sample of the Brazilian soil was mixed with an alkaline solution, simulating the pore water leached in the first stage of cement degradation, during 1, 7, 14 and 28 days. The experiments were conducted under low temperature (25 °C) aiming to evaluate similar conditions found on a low and intermediate level radioactive waste disposal installation. A non-classical isotherm sorption model was fitted to sorption data obtained from batch experiments, for unaltered and altered samples, providing parameters that allowed us to assess the effect of the interaction on material quality as Cs sorbent. The sorption parameters obtained from the data-fitted isotherm were used then to estimate the corresponding retardation factor (R). Alkaline interaction significantly modified the soil sorption properties for Cs. The parameter Q, related to the maximum sorption capacity, as well as the affinity parameter (K) and the retardation coefficients became significantly smaller (about 1000 times for the R coefficient) after pretreatment with the simulated alkaline solutions. Moreover, the increase in n-values, which is related with the energy distribution width and heterogeneity of surface site energies, demonstrated that the adsorbent surface became more homogenous as a consequence of the alkaline alteration. Together these results suggest that cementitious leachate has a profound effect on Cs retention and should be accounted for estimating radionuclide retention in radioactive waste disposal systems containing cementitious materials.

A rapid method for determining 90Sr in leaching solution from cement solidification of low and intermediate level radioactive wastes

A rapid, low cost method for the determination of 90Sr in leaching solution of cement solidification was developed using triple-to-double coincidence ratio (TDCR) Cerenkov counting of 90Y. In this method, 90Y was separated from sample matrix and interfering radionuclides through two co-precipitation steps with hydrous titanium oxide followed by yttrium fluoride (YF3) precipitation. After dissolution of the YF3 precipitate, 90Y was measured by TDCR Cerenkov technique. The performance of the method was examined using a series of the samples spiked with various activities of 90Sr ranging from 1.75 to 32 Bq L−1, along with the procedural blanks. The mean chemical recovery of Y was 86 ± 8%. The measured 90Sr agreed well with the expected 90Sr within an acceptable deviation of ~10%. The minimum detectable concentration was about 0.3 Bq L−1 for 0.2 L sample with a counting time of 5400 s by a Hidex 300 SL TDCR liquid scintillation counter. The overall decontamination coefficients were determined to be >5400 for 60Co, >5300 for 137Cs, >14,500 for 40K and 4500 ± 580 for 210Pb/210Bi, respectively.