Borate Waste Research Articles

Waste acceptance criteria testings of a phosphate-based geopolymer waste form to immobilize radioactive borate waste

Immobilization of radioactive borate waste (BW) is an urgent issue in South Korea because of the lack of appropriate waste forms. This study aimed to investigate whether phosphate-based geopolymers could satisfy the waste acceptance criteria established by Korean government. Geopolymers with 40 wt% BW did not exceed the criterion of compressive strength (3.445 MPa) after the water immersion and thermal cycling tests, whereas geopolymers containing 30 wt% BW still showed higher than 3.445 MPa after waste acceptance criteria tests. In particular, the compressive strength increased and the total porosity significantly decreased after the thermal cycling and gamma irradiation tests. The leachability index of boron calculated according to the ANSI/ANS 16.1 test was found to be higher than the criterion value (>6.0), and the controlling leaching mechanism was diffusion. The characteristic analyses indicated that water-soluble borate phases were leached from the geopolymer during the water immersion and leaching tests. However, the structurally incorporated 4-coordinated boron was also present as a sparingly soluble phase. Consequently, phosphate-based geopolymers can be used as stable binder materials to immobilize radioactive BW.

Effect of Si/Al molar ratio and curing temperatures on the immobilization of radioactive borate waste in metakaolin-based geopolymer waste form

Immobilization of radioactive borate waste (RBW) using a geopolymer with a high Si/Al ratio has been challenging because boron-silicon networks lower the compressive strength and delay the setting time. In this study, metakaolin-based geopolymer waste form to immobilize simulant RBW was fabricated using different Si/Al ratios (1.0–1.4) and curing temperatures (26 and 60 ℃). The 7-day compressive strength results revealed that a certain amount of silicon and an elevated curing temperature are required to achieve high compressive strength and waste loading. Following waste acceptance criteria tests, all geopolymers exhibited compressive strengths higher than 3.445 MPa. The leachability index of boron was higher than 6.0, and the leaching mechanism was identified as diffusion. No significant structural changes in the geopolymer were observed after thermal cycling and gamma irradiation tests. The physically bound or unincorporated RBW was leached out of the geopolymer during water immersion and leaching tests; however, boron, which was chemically connected with silicon, was present as an inert phase together with a geopolymer binder. Consequently, immobilizing RBW using a geopolymer with a low Si/Al ratio (1.4) is beneficial in terms of RBW loading and structural durability.

Immobilization mechanism of radioactive borate waste in phosphate-based geopolymer waste forms

The solidification of radioactive borate waste (BW), which contains a high concentration of boron ions using cement waste forms is challenging because soluble borate hinders the cement hydration reaction. Here, BW was immobilized using a phosphate-based geopolymer, and the phase change of BW in the geopolymer was investigated. The geopolymers could solidify BW up to 50 wt% depending on the water content and heat curing conditions. In geopolymers cured at high temperatures (60 and 90 °C), the 7-day compressive strength increased as the BW waste loading increased up to 40 wt%. Heat curing promoted geopolymerization and the precipitation of amorphous boron phosphate, leading to an increase in the compressive strength. The formation of a new amorphous boron phosphate phase was confirmed by performing XRD, MAS NMR, and SEM analyses of the geopolymers and reaction products. These results can improve our understanding of the immobilization mechanism of BW in phosphate-based geopolymers.

Immobilization Mechanism of Radioactive Borate Waste in Phosphate-Based Geopolymer Waste Forms

Immobilization Mechanism of Radioactive Borate Waste in Phosphate-Based Geopolymer Waste Forms

Development of geopolymer waste form for immobilization of radioactive borate waste

Radioactive borate waste containing a high concentration of boron (B) is problematic to be solidified using cement because soluble borate such as boric acid hinders the hydration reaction. In this study, borate waste was used as a raw material for metakaolin-based geopolymer according to the characteristic that B replaces a part of Si. Geopolymers using KOH alkaline activator (K-geopolymers) showed higher compressive strength than geopolymers using NaOH alkaline activator (Na-geopolymer). In addition, the compressive strength increased proportionally to the Si/(Al+B) ratio regardless of the alkaline cation species. These variations in compressive strength might be due to the viscosity of the geopolymer mixture, atomic size of alkaline cations, and the increase in Si content. The characteristic analyses (XRD, FT-IR, and solid state 11B MAS NMR) indicated that B was incorporated into the geopolymer structure. Thus, the K-geopolymer has a dense and homogeneous microstructure. In a semi-dynamic leaching test, less B leached from the geopolymers compared to the cement waste form. Consequently, borate waste can be solidified using metakaolin-based geopolymer, and the use of a KOH alkaline activator is advantageous in terms of mechanical property and structural durability.

Immobilization of radioactive borate liquid waste using natural diatomite

In this article, influence of diatomite as a natural adsorbent in immobilization of radioactive borate liquid waste in the cement-diatomite matrix was investigated. Cylindrical concrete mixtures were prepared with different amounts of diatomite additives at three different w/c values. Mechanical strength tests were carried out to determine the uniaxial compressive strength of the samples. A new dynamic column type leaching tests were applied to the solidified waste forms, and leach rates of 137Cs and 60Co were determined according to different diatomite additives. The experimental results indicated that the increasing of the diatomite additive amounts resulted high isolation performance of the solidified waste form. However, the results revealed that mechanical strength decreases with increasing borate diatomite amount. Results show that 9% amount of natural diatomite is effective as an admixture of the cementation process for achieving long-term stability of solidified borate waste. Optimum strength and leach values were determined as 11.5 MPa with leaching rates of 7.5 × 10−4 and 3.2 × 10−5 cm/d for 137Cs and 60Co, respectively.

Immobilization of Simulated Borate Radioactive Waste Solution in Cement-Poly(methylmethacrylate) Composite: Mechanical and Chemical Characterizations

Solidified cured cement blocks containing different concentrations of the borate waste simulate were impregnated using methyl methacrylate monomer having 0.3% by weight benzoyl peroxide as initiator. Then, the polymerization process was completed by heating the impregnated blocks at ≈40°C. To define the suitability, range of applicability, and compatibility of the inert matrix used, relevant to the borate waste immobilized, the final radioactive borate waste forms are evaluated on the basis of their mechanical and chemical properties. Some factors that may affect the characterization of the obtained final waste form, for example, concentration of the incorporated borate waste solution simulate, aging of the solid waste form, before immersion, immersion in various leachants, and time of immersion, were followed systematically. It could be stated that cement-poly(methylmethacrylate) composite exhibited better mechanical and chemical characters compared to the cement as a unique matrix for immobilization of borate waste concentrates.

Microporous glass ceramics from combination of silicate, borate and phosphate wastes

Borate mineral wastes and phosphate ash resulting from the incineration of meat and bone meal represent two particularly abundant inorganic wastes. This paper is dedicated to the combination of such wastes, together with kaolin clay, focused on the development of highly porous ceramic bodies. Borate waste has a multiple effect, providing liquid phase at the sintering temperature (1050°C), gas release (from the decomposition of its calcite fraction) and a CaO source, which reacts with residues from clay and promotes the formation of anorthite crystals as a newly formed phase. Control of the heating rate, i.e. adoption of fast heating (20°C min−1) and, above all, introduction of recycled soda–lime–silica glass as secondary additive, allowed obtaining lightweight microporous bodies (density below 0·45 g cm−3) with uniform pore structure that could be useful for thermal and acoustic insulations.

Performance of aged cement–polymer composite immobilizing borate waste simulates during flooding scenarios

An advanced composite of cement and water extended polyester based on the recycled Poly(ethylene terephthalate) waste was developed to incorporate the borate waste. Previous studies have reported the characterizations of the waste form (cement–polymer composite immobilizing borate waste simulates) after 28 days of curing time. The current work studied the performance of waste form aged for 7 years and subjected to flooding scenario during 260 days using three types of water. The state of waste form was assessed at the end of each definite interval of the water infiltration through visual examination and mechanical measurement. Scanning electron microscopy, infrared spectroscopy, X-ray diffraction and thermal analyses were used to investigate the changes that may occur in the microstructure of the waste form under aging and flooding effects. The actual experimental results indicated reasonable evidence for the durable waste form. Acceptable consistency was confirmed for the waste form even after aging 7 years and exposure to flooding scenario for 260 days.

Assessment study for multi-barrier system used in radioactive borate waste isolation based on Monte Carlo simulations

Radioactive waste generated from the nuclear applications should be properly isolated by a suitable containment system such as, multi-barrier container. The present study aims to evaluate the isolation capacity of a new multi-barrier container made from cement and clay and including borate waste materials. These wastes were spiked by (137)Cs and (60)Co radionuclides to simulate that waste generated from the primary cooling circuit of pressurized water reactors. Leaching of both radionuclides in ground water was followed and calculated during ten years. Monte Carlo (MCNP5) simulations computed the photon flux distribution of the multi-barrier container, including radioactive borate waste of specific activity 11.22KBq/g and 4.18KBq/g for (137)Cs and (60)Co, respectively, at different periods of 0, 15.1, 30.2 and 302 years. The average total flux for 100cm radius of spherical cell was 0.192photon/cm(2) at initial time and 2.73×10(-4)photon/cm(2) after 302 years. Maximum waste activity keeping the surface radiation dose within the permissible level was calculated and found to be 56KBq/g with attenuation factors of 0.73cm(-1) and 0.6cm(-1) for cement and clay, respectively. The average total flux was 1.37×10(-3)photon/cm(2) after 302 years. Monte Carlo simulations revealed that the proposed multi-barrier container is safe enough during transportation, evacuation or rearrangement in the disposal site for more than 300 years.

Chemical stability of seven years aged cement–PET composite waste form containing radioactive borate waste simulates

Different samples of radioactive borate waste simulate [originating from pressurized water reactors (PWR)] have been prepared and solidified after mixing with cement–water extended polyester composite (CPC). The polymer–cement composite samples were prepared from recycled poly (ethylene terephthalate) (PET) waste and cement paste (water/cement ratio of 40%). The prepared samples were left to set at room temperature (25 °C ± 5) under humid conditions. After 28 days curing time the obtained specimens were kept in their molds to age for 7 years under ambient conditions. Cement–polymer composite waste form specimens (CPCW) have been subjected to leach tests for both 137Cs and 60Co radionuclides according to the method proposed by the International Atomic Energy Agency (IAEA). Leaching tests were justified under various factors that may exist within the disposal site (e.g. type of leachant, surrounding temperature, leachant behavior, the leachant volume to CPCW surface area…). The obtained data after 260 days of leaching revealed that after 7 years of aging the candidate cement–polymer composite (CPC) containing radioactive borate waste samples are characterized by adequate chemical stability required for the long-term disposal process.

Immobilization of borate waste simulate in cement–water extended polyester composite based on poly(ethylene terephthalate) waste, part 3: Behavior of solid waste form during a scenario of flooding accident in the disposal site

Laboratory scale testing program was conducted to investigate the feasibility of using water extended polyester cement composite (PCC) based on poly- (ethylene terephthalate) waste (PET) as immobilization matrix for borate waste, originated from pressurized water reactors (PWRs), during prolonged term of disposal process and in scenario of flooding accident. Twenty-eight days cured solid blocks prepared as a premix of 3% by weight of the water extended polyester (WEP) and ordinary Portland cement (OPC) past having 40% by weight water were subjected to static immersion in three type of leachants, namely tap, sea, ground water. Weight changes, compressive strength variations, porosity, water absorption, and some other physical properties of the solid PCC waste forms were evaluated during 260 days of the static immersion. X-ray diffraction (XRD), infrared analysis (IR), and electron scanning microscope were performed for the waste forms at the end of immersion prolonged period to follow the internal changes in the solid waste forms that may occur due to the corrosion action of the different leachants. The results obtained from the different measurements confirm the performance of the PCC solidified waste form, based on the mentioned formulation, as an acceptable matrix in the various disposal sites and for extended disposal time even at the worst flooding environment. © 2007 Wiley Periodicals, Inc. J Appl Polym Sci 104: 4113–4120, 2007

Immobilization of Borate Waste Simulate in Cement-Water Extended Polyester Composite Based on Poly(Ethylene Terephthalate) Waste: 2-Frost Resistance of the Polymer Modified Cement Composite

Modified composite from water extended polyester (WEP); based on the recycled poly(ethylene terephthalate), (PET), and ordinary Portland cement (OPC) is a candidate matrix for processing borate waste solutions originating from the primary coolant circuit of pressurized water reactors (PWRs). This part of the work discusses the frost resistance of this polymer-modified cement composite (PCC) during the disposal course. Frost resistance is assessed for the candidate PCC after 115 cycles of freezing and thawing treatment (FTC) by evaluating their durability index, compressive strength, apparent porosity, volume of open pores, water absorption, and bulk density. Infrared (IR) and X-ray diffraction (XRD) as non-destructive analyses were performed for the candidate PCC before and after the freezing/thawing treatment to follow the changes that may take place during the hydration regime. The results obtained indicate that the candidate-modified cement composite exhibits acceptable resistance to freezing and thawing treatment even at very exaggerating conditions (−19°C and +60°C).

Immobilization of Borate Waste Simulate in Cement-Water Extended Polyester Composite Based on Polyethylene Terephthalate Waste 1-Mechanical Properties of the Final Waste Forms

ABSTRACT This paper discusses the suitability of Ordinary Portland Cement (OPC), water-extended polyester (WEP), based on recycled polyethylene terephthalate (PET), composite for stabilization/solidification (S/S) of borate waste solution originating from the primary coolant circuit of pressurized water reactors (PWRs). This treatment is a new and attractive economical process that aims at keeping man and his environment healthy through controlling the dispersion of radioactive wastes (i.e., borate waste solution) and the solid plastic waste materials (i.e., PET). Mechanical integrity of the WEP as well as its polymer composite (PC) with OPC was studied under different experimental conditions. Compressive strength measurements for the final waste forms of PC containing borate waste simulate were also evaluated using various compositions of the PC (e.g., water of hydration, polymer:cement ratio, borate content, etc.) Based on the data obtained so far, it could be concluded that the candidate PC with the formulation of polymer:cement ratio 3%, water:cement ratio 40%, polymer:water of emulsion 80/20, and containing up to 3% by weight borate waste simulate shows compressive strength values acceptable for different transportation, storage, and disposal processes.

Vitrification of borate waste from nuclear power plant using coal fly ash. (I) Glass formulation development

Vitrification of borate waste from nuclear power plant using coal fly ash. (I) Glass formulation development

Vitrification of borate waste from nuclear power plant using coal fly ash. (II) Leaching behavior of the FA30 glass

The leachability of the FA30 glass was investigated using TCLP and M-PCT tests. The heavy metals, Pb, Zn, Ba, Cr, Cd and Ni were successfully immobilized in the FA30 glass, as determined by the TCLP test. The results of M-PCT tests indicated that the ion-exchange process dominated the glass corrosion. The releases of Al and Fe were due to the glass network dissolution in the FA30 glass.

A comparison of HLW-glass and PWR-borate waste glass

Glass can incorporate a wide variety of wastes ranging from high level wastes (HLW) to low and intermediate level wastes (LILW). A comparison of HLW-Glass and PWR-borate waste glass is given in this paper. The HLW glass formulation named GC-12/9B and 90-19/U can incorporate 16–20 wt% HLW at 1100°C or 1150°C. The borate waste glass named SL-1 can incorporate 45 wt% borate waste generated from PWR. Their physical properties, characteristic temperatures, chemical durability and leach behavior are summarized here. The comparison indicates: the PWR-glass SL-1 can incorporate up to 45 wt% waste oxides at lower melting temperature (1000°C) in agreement with minimum additive waste stabilization (MAWS) approach; owing to the PWR-borate glass contain less Si and more B and Na, its mass loss is higher than HWR-glass; both HLW-glass and PWR-borate glass have favorable chemical durability and the same leaching phenomena, i.e., Na is mostly depleted, but Ca, Mg, Al and Ti are enriched in the leached surface layer.

Vitrification of borate waste from nuclear power plant using coal fly ash. (I) Glass formulation development

Borate waste, which is the main liquid waste from daily operation of PWR nuclear power plants, contains a large amount of Na and B. The coal fly ash is a by-product of the coal combustion power plants. The high content of SiO 2 and Al 2O 3 of coal fly ash makes it a desirable glass network former additive; therefore, using the coal fly ash to vitrify the borate waste without other additives is of great interest from the view point of economic and environment. In this paper, we evaluate the suitability for vitrifying borate waste using coal fly ash. Upon to 40 wt% of borate waste oxides could be successfully vitrified. Glass with 30 wt% of borate waste and 70 wt% of fly ash contain no crystals and show good chemical durability and suitable viscosity. The coal fly ash is a promising material for use in the immobilization of borate waste.

Utilizing the KEP-A glass frit to vitrify low-level radioactive waste from Korean NPPs

Vitrification has been considered as an integrated, cost effective, and environmentally sound solution for low-level radioactive waste (LLW) generated from nuclear power plants (NPPs) in Korea. Korea is in the process of preparing for its first ever vitrification plant to handle LLW from her NPPs. As a part of this project, Nuclear Environment Technology Institute (NETEC) is developing glass formulae for the different waste streams. KEP-A is the glass frit which has been developed to vitrify polyvinyl chloride (PVC) from Korean NPPs. Since 1996, glass formulae have been studied in lab-scale using simulated wastes by considering leacheability of the waste glass and viscosity of the glass melt. Based on the database of 55 glasses, on the ternary diagram of Si/Al-B/alkali–other components, the acceptance regions of the waste glasses have been presented considering its durability and viscosity. Utilizing one glass frit to vitrify different waste types with variable compositions is a sub-topic of this project. This paper presents the test results of glass formulation to vitrify other low-level radioactive waste from Korean NPPs. The waste streams include three categories, combustible Dry Active Wastes (DAW), borate waste, and spent resin. The combustible DAW in this research contain PVC, paper, and cotton. Considering chemical durability and viscosity, it is found that the KEP-A glass frit can be used to vitrify mixtures of combustible DAW up to 50 wt% of ash loading. The KEP-A allows wide composition range of combustible DAW while maintaining good durability. However, with more than 15 wt% of ash loading of borate waste or spent resin, the KEP-A could not produce a durable glass.

Leaching of borate waste cement matrices: pore solution and solid phase characterization

The chemical composition of the pore solution expressed from matrices of simulated borate liquid waste (containing H3BO3, Na2SO4, Na2HPO4 and Ca3(PO4)2) and pozzolanic cement (Spanish type IV-35) and the changes caused by leaching over 180 days in deionized water at temperatures of 20 and 40°C were studied. The leaching was carried out according to the ANSI/ANS-16.1-1986 test. In addition, the change of the solid phase was characterized by X-ray diffraction, infrared spectroscopy and scanning electron microscopy.