High-level Radioactive Liquid Waste Research Articles

Thermo-physical behavior of borosilicate glasses in the presence of high-level radioactive liquid waste constituents

Elemental composition of high-level radioactive liquid waste (HLW) from samples of various batches was assessed by using side-on-view inductively coupled plasma-atomic emission spectroscopy (ICP-AES). Under robust condition (Mg ratio, 10.2 ± 0.4 at RF power 1.3 kW), t he plasma excitation temperature (Texe, 9925 ± 1950 K) was calculated using the Boltzmann plot method. Thermal behavior of major constituents (> 1 g L−1) of HLW, such as nitrates of Na+, Fe3+, Cs+, Nd3+, and UO22+ , were studied using simultaneous thermal analyzer. Subsequently, the thermal stability of five and seven components borosilicate glasses was studied in detail with the above nitrates individually added as additives. Vapor pressure of sodium metaborate (NaBO2, one of the major vapor species from the glass melt in the off-gas system during vitrification of HLW) was estimated using transpiration thermogravimetry. Based on mass loss, the equilibrium vapor pressures (2.51–26.37 Pa) over a temperature span of 1233.15–1343.15 K were estimated. The molar enthalpy of vaporization (ΔvapH°) for sodium metaborate was calculated to be 287.84 ± 0.94 kJ mol−1 using the experimental data.

Cesium-137 Concentration in the Surface Waters of Eurasian Seas: Evidence from the Expedition Research of 2017

New data were obtained during Cruise 42 of the R/V Akademik Boris Petrov (January–March 2017) and Cruise 68 of the R/V Akademik Mstislav Keldysh (June–August 2017) on the concentration of the anthropogenic radionuclide cesium-137 in the surface waters of the East China Sea, South China Sea, equatorial Indian Ocean, Red Sea, Mediterranean Sea, Bay of Biscay, English Channel, Baltic Sea, North Atlantic, Norway Sea, and Barents Sea. Low cesium-137 concentrations were observed in the surface waters of the Eurasian seas. An exception is the Baltic Sea, where the cesium-137 content is much higher than the values obtained before the Chernobyl accident. Compared with previous estimates, higher cesium-137 concentration was observed in the surface water of the East China Sea, which could be related to the extensive discharge of high-level liquid radioactive waste after the Fukushima Daiichi nuclear accident of 2011.

Simulated studies on optimization and characterization of feed and product of melter for safe disposal of high-level radioactive liquid waste

Abstract The elemental composition of actual High-level Liquid Waste (HLW) received from reprocessing plant was estimated from analysis of several samples by using Inductively Coupled Plasma-Atomic Emission Spectrometry (ICP-AES). Based on the composition, the simulated waste (20–30 wt % based on oxide content) was prepared and vitrified in SiO2·Na2O·B2O3·TiO2·Fe2O3 glass system. Simulated Vitrified Waste Products (SVWPs) were characterized for their structural changes and relative thermal properties by using Infra-red (IR) spectrometry, X-ray Diffraction (XRD) and Differential Scanning Calorimetry (DSC) techniques. The measure of BO3, BO4, and Si−O−Si functional groups were found to be decisive in thermal and chemical durability of the products. Leach rate (LRNa) was found to be 10−5−10−6 g/cm2 day for ≤26% waste oxide (WO) loaded SVWPs. The relative increase in surface area by imparting 1 J of energy to the SVWPs was found to be

Leaching studies on borosilicate glasses for the immobilization of high-level radioactive waste in the pellet form subjected to aggressive test conditions

Pellet leaching and associated thermal and structural changes of sodium borosilicate (NBS) glass, used for the immobilization of high-level radioactive liquid waste, subjected to aggressive test conditions have been compared with international simple glass (ISG) subjected to the same leaching conditions. The crystalline phase getting separated out from NBS glass is found to be different for pellets and powder leaching experiments and this has been explained based on the difference in the extent of leaching occurring with glass samples in the two experiments. Based on Fourier transform infrared studies, it is inferred that, unlike in the ISG sample, Si–O–Si/B structural units become more ordered with the leaching in the case of NBS glass, and this is attributed to the partial network destruction occurring with NBS glass and crystallization of the \(\hbox {SiO}_{2}\) phase from the glass matrix. Both the NBS glass and ISG sample show L-centre emission and the emission intensity remained unaffected with leaching, confirming that the local environment around non-bridging oxygen atoms in the NBS glass and ISG sample are unaffected and leaching occurs through network dissolution.

Chemical forms of molybdenum ion in nitric acid solution studied using liquid-phase X-ray absorption fine structure, Ultraviolet–Visible absorption spectroscopy and first-principles calculations

Abstract We have investigated chemical forms of molybdenum ion in nitric acid solution, using liquid-phase X-ray absorption fine structure, ultraviolet–visible absorption spectroscopy and first-principles calculations, from a viewpoint of disposal of high-level radioactive nuclear liquid wastes. The experimental and theoretical results indicated that Mo is a hexavalent ion and forms a hexa-coordination binuclear-structured complex in the 2 M nitic acid solution. The predominant chemical species of Mo complexes in the 2 M nitric acid solution (which is used for HLLW) are assigned to be [Mo2O5(H2O)6]2+ and [Mo2O4(OH)(H2O)6]3+. These species may play a crucial role of forming so-called yellow-phase in vitrified objects.

Preparation and characterization of tri-n-octylamine microcapsule (TOA-MC) used for selective separation of Re(VII) [chemical analogs of Tc(VII)] in the high level liquid radioactive waste (HLLW)

TOA-MC was developed to act as absorbent for selective separation of 99Tc(VII) from high-level radioactive liquid waste. The uptake (%) of Re(VII) a substitute for Tc(VII) due to similar chemical behavior by TOA-MC normally exceeding 95% in the presence of dilute HNO3 was strongly retained and decreased with higher HNO3 concentration by batch method. The surface morphology, thermal stability and diameter/size distribution of the microcapsules were examined by scanning electron microscopy, thermogravimetric analysis and digital microscope, respectively. The chemical structure was characterized using Fourier transform infrared spectroscopy and 1H nuclear magnetic resonance spectroscopy.

Liquid-Air Interface Corrosion of A537 Steel in High Level Liquid Radioactive Waste Simulant in a Sealed Container

Liquid-air interface (LAI) corrosion is regarded as a potential issue for interim storage tanks of high level liquid radioactive waste. Insightful studies have been carried out and several hypotheses have been proposed so far, while none has been directly confirmed experimentally. A sealed container has been developed to investigate the effect of atmosphere on LAI corrosion of partially immersed A537 steels, in which the composition and temperature of air can be changed accordingly, i.e. CO2 concentration varied between 0.05 wt% (natural air) and 1.97 wt%, and temperature varied between 25°C and 34°C. It has been found that the pitting initiation process under both anodic polarization and free immersion conditions can be largely accelerated by the increase of CO2 concentration. More importantly, a planar pH electrode has been applied to measure pH at the interface during anodic polarization and free immersion, and the occurrence of LAI corrosion can be correlated with the same interface pH at a certain temperature regardless of CO2 concentration, indicating a “critical pH” for LAI corrosion determined by this method. This finding suggests that LAI corrosion in high level liquid radioactive waste may be investigated through determination of a critical pH in certain circumstances.

Acid tolerant covalently functionalized graphene oxide for the selective extraction of Pd from high-level radioactive liquid wastes

Covalently functionalized graphene oxides were explored as a promising material for the recycle of severely depleting highly demanded palladium (Pd) from strongly acidic conditions.

Influence of Na-Al-Fe-P glass alteration in hot non-saturated vapor on leaching of vitrified radioactive wastes in water

Influence of preliminary heating in vapor on leaching of Na-Al-P-glass in water is investigated. The glass simulates industrial waste forms for solidification of high level liquid radioactive wastes used in Russia. Heating up to 300 °C in unsaturated vapor leads to transformation (devitrification) of pristine glass into four separate crystalline phases. The alteration takes place in whole volume of the monolithic glass sample. Experiments on leaching of altered and pristine glass in water are carried out. A ratio of estimated leaching rates of the altered and pristine glass can exceed two orders of magnitude. Fractions of radionuclide simulators in colloid of different size in leaching products are determined.

Risk analysis of high level radioactive waste storage tank based on HAZOP

High-level radioactive liquid waste tanks have the possibility of releasing radioactivity, and thus it is of great importance to identify hazards in such storage tanks and take measures to improve their safety. In this paper, a hazard and operability (HAZOP) methodology is used to identify the hazards in a double-shell stainless-steel tank for high-level radioactive liquid waste storage for which a risk matrix was implemented to assess the risks posed by the hazard. First, during the HAZOP process, appropriate process parameters and guidewords were selected to describe the deviation of the storage tank, and the causes and possible consequences of deviations as well as existing protection measures were analyzed. The risk matrix then was used to evaluate the risk level of the deviation identified, combined with the consequences of the deviation as well as the associated likelihood of its occurrence. Results showed that primary tank leaks and hydrogen combustion or explosion in tanks are undesirable risks, and thus corresponding countermeasures were put forward. The analysis results could be used to provide guidance for safe operation of, and rectification of hidden dangers in, high-level radioactive liquid waste storage tanks.

Spectroscopic and first-principles calculation studies of the chemical forms of palladium ion in nitric acid solution for development of disposal of high-level radioactive nuclear wastes

We have investigated the chemical forms of palladium (Pd) ion in nitric acid solution, using XAFS/UV-vis spectroscopic and first-principles methods in order to develop the disposal of high-level radioactive nuclear liquid wastes (HLLW: radioactive metal ions in 2 M nitric acid solution). The results of theoretical calculations and XAFS/UV-vis spectroscopy indicate that Pd is a divalent ion and forms a square-planar complex structure coordinated with four nitrate ions, [Pd(NO3)4]2-, in nitric acid solution. This complex structure is also thermodynamically predicted to be most stable among complexes [Pd(H2O)x(NO3)4-x]x-2 (x = 0-4). Since the overall feature of UV-vis spectra of the Pd complex was independent of nitric acid concentration in the range 1–6 M, the structure of the Pd complex remains unchanged in this range. Furthermore, we examined the influence of γ-ray radiation on the [Pd(NO3)4]2- complex, using UV-vis spectroscopy, and found that UV-vis spectra seemed not to be changed even after 1.0 MGy irradiation. This implies that the Pd complex structure will be still stable in actual HLLW. These findings obtained above are useful information to develop the vitrification processes for disposal of HLLW.

New route for synthesis of Synroc-like ceramic using non-selective sorbent LHT-9

ABSTRACT:Immobilization of bulk liquid high level radioactive waste (HLW) in Synroc ceramic is well-known reliable way for final isolation of dangerous long-lived radionuclides from biosphere. The alternative method of Synroc-like ceramic synthesis has been proposed. Radionuclide incorporation into crystalline titanate host-phases can be provided as a result of direct radionuclide sorption from liquid HLW using non-selective sorbent – layered hydrazinium titanate (LHT-9). Such an approach allows excluding expensive multi-stage procedure of precursor preparation. The precipitate obtained after sorption can be easily transformed into Synroc-like ceramic by cold pressing followed with sintering in air at 1000-1200°C. The highly radioactive samples of titanate ceramic loaded with real HLW have been synthesized and preliminary studied at KRI hot-cell facility. Chemical durability of this sample has been studied using static leach test in distilled water at 90°C and the leach rates for 154Eu, 241Am, 244Cm were (in g·m-2·day-1) from 2·10-4 to 5·10-3. Normalised 137Cs mass loss was 0.3 g·m-2 for 110 days at the same conditions.

<sup>107</sup>PdのICP-MS測定のためのレーザー誘起光還元法による非接触・選択的パラジウム分離 ─分離条件とPd回収率の関係─

Palladium-107 can be found in high-level radioactive liquid wastes (HLLW). Owing to its long half-life (6.5 million years), determining the 107Pd contents in HLLW is essential to evaluate the long-term safety of HLLW repositories. However, the 107Pd contents in HLLW have not been reported because of difficulties in a pretreatment for the measurement. Treating highly radioactive samples, such as HLLW, may cause radioactive contamination and radiation exposure to workers. Therefore, a simple and non-contact separation procedure is desirable. In this study, we applied a simple and non-contact separation technique based on laser-induced photoreduction to a simulated HLLW solution. The solution, which consisted of 14 major elements (Rb, Sr, Zr, Mo, Ru, Rh, Pd, Cs, Ba, La, Ce, Pr, Nd, Sm) in a 3 mol L−1 HNO3 solution, was used to evaluate the separation performance. The results showed that the recovery of Pd from a simulated HLLW solution depended on the solvent, concentration of ethanol, concentrations of Pd ion, irradiation time and pulse energy. The results indicate that the coexisting elements in a sample hampered the photoreduction of Pd ions or the formation of a Pd precipitate and the loss of a Pd precipitate at centrifugation caused a decrease of the Pd recovery. By considering the conditions at laser irradiation, the results showed that the Pd recovery reached 59 %, while about 99 % of the other 13 elements was removed. This indicates that the non-contact and high-selective separation of Pd is achievable with the proposed separation technique under a suitable separation condition.

高レベル放射性廃液模擬環境でのステンレス鋼腐食に及ぼす減圧沸騰の影響

高レベル放射性廃液模擬環境でのステンレス鋼腐食に及ぼす減圧沸騰の影響

Analysis on the characteristics of geologic disposal waste arising from various partitioning and conditioning options

Siting a nuclear geological repository entails high economic, social, and political costs. The concept of expanding the capacity of repository is of great interest to the nuclear industry. Partitioning and conditioning (P&C) technology of high level radioactive liquid waste (HLLW) is expected to be effective to mitigate the burden of repository by removal of some elements from disposal waste and reducing the repository size necessitated to emplace the waste forms. The impacts of P&C technology on the characteristics of disposal waste are discussed in this paper. Four P&C options are assumed in the analysis: (1) Spent fuel once through, to dispose spent fuel directly; (2) Conventional reprocessing, to dispose vitrified HLLW glass; (3) Reprocessing plus long-lived nuclides partitioning, to dispose waste streams without minor actinides (MAs), rare earth elements (REs), Tc and Pu; (4) Reprocessing plus high thermal output nuclides separation, to dispose waste streams without Sr&Cs. The partitioning scheme is based on China’s TRPO full partitioning process and its experimental results. The waste conditioning and disposal is in accordance with the specifications of Yucca Mountain Repository. The content of disposal waste stream in each option is presented and its characteristics, such as mass, volume, linear loading capacity, thermal power and linear thermal power, are identified.

The Application of Nitrogen Laser on Extraction of Uranium in The Long Life of High Level Radioactive Liquid Waste Using TBP-Kerosene Solvent

The radionuclide of uranium (U) is major radionuclide contained in the long life of high level radioactive liquid waste (HLLW) generated from reprocessing of spent nuclear fuel. The radioactive waste have to be treated to ready for long term disposal. Separation of the U at high efficiencies greatly lowers the volume of the long life alpha radioactive waste to be disposed and decreases the hazard level of the waste. The technology assessment of selective separation of U was carriedout as alternative and strategy for HLLW management in the future. The selective separation technology of U from fission products at very high efficiencies was developed by the extraction process using TBP-kerosene solvent and increasing the separation by exposure of nitrogen (N2) laser radiation at wavelength 337.1 nm. In the extraction process for simulation waste containing U and Zr in 5 M HNO3 (Zr as one of fission product which difficult to be separated from U) using 30 % TBP-kerosene solvent and by exposure of nitrogen laser radiation shows that increasing of distribution coefficient of U (Kd U) can obtains 135 % and increasing of separation factor of U and Zr (SF(U/Zr)) is 189 %. The increasing of Kd U using N2 laser higher than using CO2 laser (at the wave-number 944 cm-1 ) which increase of Kd U only 100 %. In Indonesia, assessment for adaptation of the separation technology by extraction for separation of U process using 30 % TBP-kerosene solvent should be carried-out as alternative for treatment the HLLW generated from 99Mo radioisotope production and from post irradiation examination of nuclear fuel.

Utilization of Coal Fly Ash as CO Gas Adsorbent

The radionuclide of uranium (U) is major radionuclide contained in the long life of high level radioactive liquid waste (HLLW) generated from reprocessing of spent nuclear fuel. The radioactive waste have to be treated to ready for long term disposal. Separation of the U at high efficiencies greatly lowers the volume of the long life alpha radioactive waste to be disposed and decreases the hazard level of the waste. The technology assessment of selective separation of U was carriedout as alternative and strategy for HLLW management in the future. The selective separation technology of U from fission products at very high efficiencies was developed by the extraction process using TBP-kerosene solvent and increasing the separation by exposure of nitrogen (N2) laser radiation at wavelength 337.1 nm. In the extraction process for simulation waste containing U and Zr in 5 M HNO3 (Zr as one of fission product which difficult to be separated from U) using 30 % TBP-kerosene solvent and by exposure of nitrogen laser radiation shows that increasing of distribution coefficient of U (Kd U) can obtains 135 % and increasing of separation factor of U and Zr (SF(U/Zr)) is 189 %. The increasing of Kd U using N2 laser higher than using CO2 laser (at the wave-number 944 cm-1 ) which increase of Kd U only 100 %. In Indonesia, assessment for adaptation of the separation technology by extraction for separation of U process using 30 % TBP-kerosene solvent should be carried-out as alternative for treatment the HLLW generated from 99Mo radioisotope production and from post irradiation examination of nuclear fuel.

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).

Indian Program for Vitrification of High Level Radioactive Liquid Waste

Nuclear Power Programme in India is based on “closed fuel cycle”. Closed fuel cycle involves reprocessing and recycling of Spent Nuclear Fuel (SNF) coming out of nuclear reactors. During reprocessing, uranium &plutonium, constituting bulk of the SNF are separated and subsequently recycled. The remaining small portion constitutes high level radioactive waste containing most of the fission products and minor actinides. A three-step strategy involving immobilization, interim storage followed by ultimate disposal has been adopted in India for management of High Level Waste (HLW). Borosilicate glass matrix has been identified for immobilization of HLW owing to optimal waste loading, adequate leach resistance and long term stability of the product. Glass formulations are developed by suitable addition of modifiers to accommodate compositional variation in the waste. Detailed characterization studies are carried out to understand the structural modifications in the three dimensional network of incorporation of waste constituents followed by product characterization to ascertain the properties of the conditioned vitrified waste product (VWP). On the strength of these indigenous developments, Indian vitrification facilities are in operation at Trombay, Tarapur and Kalpakkam. Different types of melters like metallic, Joule Heated Ceramic Melter (JHCM) have been successfully deployed on industrial scale for vitrification of HLW. A Cold Crucible Induction Melter (CCIM) has also been developed. An interim storage facility is in operation for storage and surveillance of VWP. A site selection programme has been initiated to identify a few suitable geological domains for identifying a candidate disposal site for high level and long lived radioactive wastes. Presently granites have been studied extensively as a host rock & as a natural barrier. Long term evaluation of vitrified high level waste under geological conditions and its comparison with natural analogues (basaltic & obsidian) is being pursued in India.

Evaluation of Nitrate and Nitrite Reduction Kinetics Related to Liquid-Air-Interface Corrosion

Liquid-air interface (LAI) corrosion has been a concern for causing leaks in the carbon steel tanks used for holding high-level radioactive liquid waste. To assist in understanding the mechanism of LAI corrosion, the kinetics of nitrate and nitrite reduction reactions were investigated electrochemically. Cyclic voltammetry and cathodic polarization measurements indicated that the nitrite reduction reaction exhibited faster kinetics than the nitrate reduction reaction at higher cathodic overpotential. However, the primary reduction reaction at the open circuit potential under aerated conditions was the oxygen reduction reaction. The reduction of residual oxygen was also the dominant cathodic reaction at open circuit potential in deaerated conditions. Moreover, the kinetics of oxygen reduction on steel electrodes were significantly influenced by the sample immersion conditions (partial vs. full) for aerated liquid nuclear waste simulants, but not for deaerated conditions. Lastly, the gaseous products formed during LAI corrosion were analyzed using the gas detector tube method and gas chromatography-mass spectrometry and found to contain NH3, NO2 and NO. However, the results suggested that these products were caused by the local acidification generated by the hydrolysis of cations after LAI corrosion underwent extensive propagation, instead of being directly reduced in alkaline conditions. Thus, the results in this work showed that the kinetics of nitrate and nitrite reduction could not generate a salt concentration cell in the meniscus region to cause LAI corrosion.