Intermediate-level Liquid Waste Research Articles

Nanocrystalline Ce(OH)4-based materials: ruthenium selective adsorbent for highly alkaline radioactive liquid waste

Abstract Cerium hydroxide, Ce(OH)4 (Ce), has been synthesised and assessed as a Ru-selective adsorbent for treating alkaline radioactive liquid waste. Infrared spectroscopy, thermal analysis, scanning electron microscopy, and energy-dispersive X-ray spectroscopy investigations confirmed the successful formation of nanocrystalline Ce from Ce(NO3)3·6H2O. Selective removal of 106Ru from the ion-exchange effluent of intermediate-level liquid waste (ILW) by Ce was assessed using a high-pure germanium (HPGe) gamma-ray spectrometer. The calculated average distribution coefficient (k D) was ∼200 mL/g. The percentage removal of 106Ru using Ce by varying time, [106Ru] and [Ce] was calculated. The adsorption of 106Ru on Ce follows pseudo-second-order and Freundlich isotherms. The calculated Q max was 93,584 Bq/g. Accelerated leaching studies of the Ru-laden Ce cement product were carried out and found suitable for transport and disposal. Further, Ce-Polyether sulphone (Ce-PES) and Ce-Chitosan (CeC) composites were prepared and assessed for their Ru-uptake capacity for engineering scale application.

Evaluation of cementation of intermediate level liquid waste produced from fission 99Mo production process and disposal feasibility of cement waste form

The Korea Atomic Energy Research Institute (KAERI) is planning the construction of the KIJANG Research Reactor (KJRR) for stable supply of 99Mo. The Fission 99Mo Production Process (FMPP) of KJRR produces solid waste such as spent uranium cake and alumina cake, and liquid waste in the form of intermediate level liquid waste (ILLW) and low level liquid waste (LLLW). This study thus established the operating range and optimum operating conditions for the cementation of ILLW from FMPP. It also evaluated whether cement waste form samples produced under optimum operational conditions satisfy the waste acceptance criteria (WAC) of a disposal facility in Korea (Korea radioactive waste agency, KORAD). Considering economic feasibility and safety, optimum operational conditions were achieved at a w/c ratio of 0.55, and the corresponding salt content was 5.71 wt%. The cement waste form samples prepared under optimum operational conditions were found to satisfy KORAD's WAC when tested for structural stability and leachability. The results indicate that the proposed cementation conditions for the disposal of ILLW from FMMP can be effectively applied to KJRR's disposal facility.

Ruthenium speciation in radioactive wastes and state-of-the-art strategies for its recovery: A review

• Recovery of 106 Ru from spent nuclear fuel is required for its use in ocular cancer therapy. • Ruthenium speciation decides its separation strategy. So it is discussed in detail. • Ru recovery from acidic and alkaline radioactive feeds is given in detail. • Ru recovery by solvent extraction, precipitation and ion exchange is summarized. • Special mention is made to ion-exchange based separation methods. Ruthenium, mainly 106 Ru, is one of the notorious fission products present in the radioactive wastes. In view of its complex chemistry, it gets partitioned into various processing streams during the radioactive waste management steps. Also, due to the volatile nature of its tetraoxide, RuO 4 , it gets deposited in the process pipelines and stacks. Attempts have been made by various researchers, over the years, to separate radio-ruthenium from radioactive feeds including the acidic high level liquid waste (HLLW) as well as the alkaline low level liquid waste (LLLW) or intermediate level liquid wastes (ILLW). Solvent extraction, precipitation, ion-exchange, etc are some of the commonly employed techniques used for the separation of radio-ruthenium from radioactive feeds. The present review article gives a state-of-the-art of the separation of radio-ruthenium from radioactive waste feeds based on Ruthenium speciation in the system. Most of the published literature involves synthetic feeds rather than actual radioactive wastes. Nevertheless, the published data gives very important information on how to separate radio-ruthenium from different types of radioactive feeds. The aqueous chemistry of Ruthenium is quite complex due to the co-existence of different ionic species of different oxidation states. In view of this, this review article begins with a brief summary on the aqueous chemistry of Ruthenium in acidic as well as alkaline feeds and subsequently its separation from these solutions were discussed using various techniques under the light of its speciation.

Современные сорбционные материалы для очистки жидких радиоактивных отходов от радионуклидов цезия и стронция

The paper describes the main types of sorption materials used to remove cesium and strontium radionuclides from low-level and intermediate-level liquid radioactive waste (LRW). It mainly focuses on sorbents produced at an industrial or pilot scale in Russia. Based on a developed standardized technique, it presents the experimental values for 137Cs and 90Sr distribution coefficients (Kd ) accounting for various types of sorbents. The study demonstrates that based on the developed technique, sorption characteristics of various types of sorption materials can be compared in an unbiased manner and most effective sorbents can be reasonably selected to treat liquid radioactive waste of various salt and radionuclide composition.

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).

Potential Use of Ambient-Cured Geopolymers for Intermediate Level Nuclear Waste Storage

AbstractThere is growing interest in reducing the use of ordinary Portland cement (OPC) owing to its high energy consumption and CO2 emissions. An environmentally-friendly alternative is the use of geopolymers, which can potentially reduce direct CO2 emissions through the appropriate choice of raw materials, mix design, and curing regimes. In this regard geopolymer mortars are also realistic candidates for the replacement of OPC mortars in nuclear waste immobilisation applications as they provide a more durable incorporation matrix as well as suppressing the formation of radiolytic hydrogen. The advantages of geopolymers over OPC for nuclear waste immobilisation include i) lower water content as alkaline activator is the main component that drives geopolymerisation, ii) higher thermal stability (<600°-800°C) compared to OPC concrete (<300°C), iii) higher compressive strength (50-80 MPa), and iv) lower leachability of radioactive ions when the mix design and curing temperature are appropriately balanced. UNSW and ANSTO have embarked on a long-term research program to investigate the possibility of using geopolymers for the immobilisation of Intermediate Level Liquid Waste (ILLW), the focus of which will be around the influence of gamma-irradiation on the durability.

Environmental Effects of Cavitation Technology for Radioactive Waste Management

Results from studies of the removal from storage and solidification of low- and intermediate-level liquid radioactive waste in inorganic binders (cementation) are presented. Cavitation technology for recycling radiochemical wastes from nuclear power plant decommissioning increases the extraction throughput and strength of the cement compound.

Advances in conditioning of low- and intermediate-level nuclear waste

ABSTRACTRadioactive waste with widely varying characteristics is generated from the operation and maintenance of nuclear reactors, nuclear fuel cycle facilities, research facilities and medical facilities and the through the use of radioisotopes in industrial applications. The waste needs to be treated and conditioned appropriately to provide wasteforms acceptable for safe storage and disposal. Conditioning of radioactive waste is an important step to prepare waste for long-term storage or disposal and includes the following processes:▪ Immobilization which may or may not also provide volume reduction, includinga) Low temperature processes andb) Thermal processes;▪ Containerization fora) Transport,b) Storage, andc) Disposal;▪ Overpacking of primary containersa) Prior to disposal andb) In a disposal facility as part of disposal process.Conditioning consists of operations that produce a waste package suitable for handling, transportation, storage and/or disposal and may be performed for a variety of reasons including standardization of practices and/or wasteforms, technical requirements for waste stability in relation to a repository design or safety case, technical requirements related to waste transportation, societal preferences, regulatory preferences, etc. This paper gives an overview of recent advances in conditioning of low- and intermediate-level radioactive waste.The paper is based on the new IAEA Handbook “Conditioning of Low- and Intermediate-Level Liquid, Solidified and Solid Waste” which is one of eight IAEA handbooks intended to provide guidance for evaluating and implementing various characterisation and radioactive waste processing and storage technologies before final disposal

Development of Industrial-Scale Fission 99Mo Production Process Using Low Enriched Uranium Target

Abstract Molybdenum-99 (99Mo) is the most important isotope because its daughter isotope, technetium-99m (99mTc), has been the most widely used medical radioisotope for more than 50 years, accounting for > 80% of total nuclear diagnostics worldwide. In this review, radiochemical routes for the production of 99Mo, and the aspects for selecting a suitable process strategy are discussed from the historical viewpoint of 99Mo technology developments. Most of the industrial-scale 99Mo processes have been based on the fission of 235U. Recently, important issues have been raised for the conversion of fission 99Mo targets from highly enriched uranium to low enriched uranium (LEU). The development of new LEU targets with higher density was requested to compensate for the loss of 99Mo yield, caused by a significant reduction of 235U enrichment, from the conversion. As the dramatic increment of intermediate level liquid waste is also expected from the conversion, an effective strategy to reduce the waste generation from the fission 99Mo production is required. The mitigation of radioxenon emission from medical radioisotope production facilities is discussed in relation with the monitoring of nuclear explosions and comprehensive nuclear test ban. Lastly, the 99Mo production process paired with the Korea Atomic Energy Research Institute's own LEU target is proposed as one of the most suitable processes for the LEU target.

Joint U.S./Russian Studies of Population Exposures Resulting From Nuclear Production Activities in the Southern Urals

Beginning in 1948, the Soviet Union initiated a program for production of nuclear materials for a weapons program. The first facility for production of plutonium was constructed in the central portion of the country east of the southern Ural Mountains, about halfway between the major industrial cities of Ekaterinburg and Chelyabinsk. The facility, now known as the Mayak Production Association, and its associated town, now known as Ozersk, were built to irradiate uranium in reactors, separate the resulting plutonium in reprocessing plants, and prepare plutonium metal in the metallurgical plant. The rush to production, coupled with inexperience in handling radioactive materials, led to large radiation exposures, not only to the workers in the facilities, but also to the surrounding public. Fuel processing started with no controls on releases, and fuel dissolution and accidents in reactors resulted in release of ~37 PBq of I between 1948 and 1967. Designed disposals of low- and intermediate-level liquid radioactive wastes, and accidental releases via cooling water from tank farms of high-level liquid radioactive wastes into the small Techa River, caused significant contamination and exposures to residents of numerous small riverside villages downstream of the site. Discovery of the magnitude of the aquatic contamination in late 1951 caused revisions to the waste handling regimes, but not before over 200 PBq of radionuclides (with large contributions of Sr and Cs) were released. Liquid wastes were diverted to tiny Lake Karachay (which today holds over 4 EBq); cooling water was stopped in the tank farms. In 1957, one of the tanks in the tank farm overheated and exploded; over 70 PBq, disproportionately Sr, was blown over a large area to the northeast of the site. A large area was contaminated and many villages evacuated. This area today is known as the East Urals Radioactive Trace (EURT). Each of these releases was significant; together they have created a unique group of cohorts with their chronic, low dose-rate radiation exposure. The 26,000 workers at Mayak were highly exposed to external gamma and inhaled plutonium. A cohort of individuals raised as children in Ozersk is under evaluation for their exposures to radioiodine. The Techa River Cohort consists of over 30,000 people who were born before the start of exposure in 1949 and lived along the Techa River. The Techa River Offspring Cohort consists of ~21,000 persons born to one or more exposed parents of this group, many who also lived along the contaminated river. The EURT Cohort consists of ~18,000 people who were evacuated from the EURT soon after the 1957 explosion and another 8,000 who remained. These groups together are the focus of dose reconstruction and epidemiological studies funded by the United States, Russia, and the European Union to address the question, "Are doses delivered at low dose rates as effective in producing health effects as the same doses delivered at high dose rates?"Introduction of Joint U.S. and Russian Studies of Population Exposures (Video 2:13, http://links.lww.com/HP/A28).

The Intermediate Level Liquid Molybdenum-99 Waste Treatment Process at the Australian Nuclear Science and Technology Organisation

ANSTO manufactures 99Mo for radiopharmaceutical use. Alkaline Intermediate Level Liquid Waste (ILLW) from this process, plus legacy acidic waste, are planned to be treated by converting both wastes into stable, solid, waste forms with oxide-basis loadings of 25-35 wt% and 30-50 wt%, respectively. The hot-cell plant design utilises the same unit process steps to treat both wastes. Hot-Isostatic Pressing (HIP) is employed to consolidate the processed waste and achieve substantial waste volume reductions compared to a cementation option. In this paper an overview of the treatment process and selected waste forms for ANSTO's 99Mo production ILLW is given.

Inorganic sorbents based on modified microspheres for treatment of liquid radioactive waste

Sorbents based on hollow microspheres of entrained ash, with the surface modified by various chemical compounds (ferrocyanides, phosphates, oxides, etc.), were synthesized. Their physical and ion-exchange characteristics were examined. These sorbents show promise for treatment of low-and intermediate-level liquid waste of various origins in the dynamic and static sorption modes. The use of microspheres as matrices is not only technically but also environmentally efficient, because waste from thermal power engineering is utilized for disposal of nuclear waste.

Removal of actinides and fission products activity from intermediate alkaline waste using inorganic exchangers

Hydrated iron oxide or amorphous-Fe2O3·3.5 H2O (HFeO), hydrated titanium oxide (HTiO) and hydrated thorium oxide (HThO) were synthesized and their applicability for the decontamination of intermediate level liquid wastes (ILLW) was tested. The sorption of a few actinides like plutonium and americium on HFeO, 137Cs and 106Ru on HTiO and 90Sr on HThO was investigated as a function of pH, time and loading capacity of the hydrous oxide with metal ions. The influence of the total dissolved salt content was also monitored. Some of these parameters influenced the sorption behavior significantly. The radiation stability of these inorganic sorbents were studied by irradiating them up to 48 Mrad. Adsorbed actinides and fission products were successfully eluted from HFeO and from the mix-bed of HTiO and HThO by 0.5M nitric acid.

Radiation-Chemical Aspects of Radioactive Waste Management

Studies of radiation-chemical processes in model liquid radioactive waste are summarized. These processes are radiolytic transformations of various substances present in the waste (actinide ions (primarily, neptunium and plutonium ions), inorganic matter (for example, nitrate ions), and organic compounds (acetic acid, EDTA, etc.)) and radiolytic gas evolution. Attention was focused on the latter process. The volumes and compositions of gases formed in the γ-radiolysis of aqueous solutions that simulated intermediate-level and high-level liquid waste were determined. It was found that the nature and the radiation-chemical yields of released gases depend on the absorbed dose, the composition and concentration of solutions, the irradiation temperature, the presence of solid substances, etc. In a number of cases, data obtained with model solutions were compared to the results of studies of real liquid radioactive waste.

Scaling-up experience of actinides partitioning from intermediate level high salted alkaline waste streams of Purex process using Versatic-10 extractant

Partitioning of minor alpha-emitting actinides, especially U, Pu and Am from medium active alkaline waste is possible from intermediate level liquid wastes (ILLW) produced during spent fuel reprocessing following Purex process. This paper deals with the efficient removal of alpha-activity from ILLW by solvent extraction process. Counter current batch extraction with O/A ratio 2:1 as well as multistage mixer settler has demonstrated that most of the alpha-activity was removed from the alkaline effluents using 20% Versatic-10 (V-10) in dodecane after giving 3 to 4 contacts, thus converting alkaline waste as non-alpha waste. Under the optimum conditions (pH 9.0-9.5 and VA-10), both Pu(IV) and Am(III) are highly extractable whereas U(VI) is relatively poorly extracted. To assess the applicability of this process for regular treatment of the waste, a feasibility study on pilot plant scale using six stage mixer settler was operated to treat the ILLW. The results indicated that almost >99.90% alpha-emitting actinides are removed. Dilute nitric acid (0.5M HNO3) served as the most efficient strippant for all these actinides. This facilitate an easy regeneration of the extractant which can be recycled. This method is useful in obtaining alpha-free wastes and had positive impact on ease and safety aspects during subsequent waste treatment and long term storage.

Use of local soil for the decontamination of liquid radioactive wastes at PINSTECH

The use of locally available mineral clay for the removal of radionuclides from the low and intermediate level liquid waste has been studied. Adsorption behavior of this naturally available inorganic adsorbent is reported. The factors such as contact time, nature of the waste, pH and adsorption capacity have been considered. These optional physico-chemical conditions suggest an effective use of this locally available adsorbent for the decontamination of the liquid radioactive waste at Pakistan Institute of Nuclear Science and Technology (PINSTECH), Islamabad.

Separation of Cesium from Intermediate Level Waste Solutions Using Zirconyl Tungstate Ceramic-Supported Membranes

Zirconyl tungstate ceramic-supported membranes were developed and characterized. The transport process of Cs(I), Sr(II) and La(III) from the anode compartment to the cathode was investigated. The cationic fluxes for Cs(I), Sr(II) and La(III) were found to be 9·10−10, 2.7·10−9 and 2.4·10−9 g·eq·cm−2·s−1, respectively. The transport numbers of these cations were found to be 0.12, 0.07 and 0.09, respectively. Separation of Cs(I), Sr(II) and La(III) was achieved using EDTA as a chelating agent in the anode compartment. Cs(I) was successfully separated with a decontamination factor of 97% from simulated intermediate level liquid waste (ILLW) containing Cs(I) and Sr(II).

Removal of137Cs from an intermediate level liquid waste by using various ion-exchange media

An intermediate level liquid waste (ILLW or Category III waste) stream was treated for the removal of radioactive Cs in pilot scale experiments, using three different ion exchange media. Results indicated that the polyacrylic fibre coated with cupric ferric hexacyanoferrate (CuFeHCF-fibre composite) was the most efficient, followed by cupric hexacyanoferrate loaded ion exchange resin (CuHCF-resin) and mixed zeolites (AR-1, 4A and 13X in the ratio 6:1:1). The mixed zeolites column and the CuFeHCF-fibre column were used in series in order to get a high total decontamination factor (DF). Leaching studies on the CuFeHCF-fibre composite loaded with137Cs, in demineralised (DM) water, tap water and ground water media indicated a release of about 19.3%, 25.5% and 41.3% of137Cs, respectively, in a period of about 8 months. XPS studies with CuFeHCF-fibre composite indicated some chemical interaction between the CuFeHCF precipitate and the polyacrylic fibre. Some of the possible disposal options for the CuFeHCF-fibre composite have also been discussed.

Separation of plutonium from intermediate level liquid waste by precipitation with alkyl-pyridinium nitrates

Separation of plutonium from intermediate level liquid waste by precipitation with alkyl-pyridinium nitrates

Leaching of Chlorine, Cesium, Strontium and Technetium from Cement-Fixed Intermediate Level Liquid Waste

ABSTRACTThe leaching behaviour of Cl-36, Cs-137, Sr-90 and Tc-99 from cement-fixed intermediate level liquid waste was investigated. Bentonite, MicroSili-ca®, acrylic resin and sodium sulfide were used as additives to a blast furnace cement matrix. In some cases portland cement was used. The liquid (water or waste solution) to cement ratio applied was mostly 0.47. Samples were leached with water or quinary brine (Q-brine).After 250 days the sequence of leachability from additive-free specimen in Q-brine was Cl>Cs>Sr>Tc with little modification in water: Cs>Cl>Sr>Tc. Cl and Cs leaching are diffusion-controlled, Sr and Tc are chemically fixed.Bentonite improves the retardation of Cs significantly, but has no significant effect on Cl, Sr and Tc. MicroSilica® deteriorates the retardation of Cs, but has no effect on Cl, Sr and Tc release.Sulfide was expected to reduce the leachability of Tc by forming insoluble TC2S7. No significant effect was observed, however, because either TC2S7 does not form at the pH value of the cement slurry or, due to a solubility competition, insoluble Tc species also form in the absence of sulfide ions.Acrylic resin does not reduce the leachability of Cs significantly. Irradiation with doses up to 4.4–105 Gy slightly deteriorates Cs retardation.Applying a liquid to cement ratio of 0.35 instead of 0.47 improves the retardation of Cs by a factor of 1.5–4, which may be due to a reduction of porosity in the specimen.