A Comparison of the Performance of Nuclear Waste Glasses by Modeling

1987 ◽  
Vol 112 ◽  
Author(s):  
B. Grambow ◽  
D. M. Strachan
Keyword(s):  
Nuclear Waste ◽  
Nuclear Reactors ◽  
Liquid Waste ◽  
Spent Fuel ◽  
High Level Liquid Waste ◽  
Geologic Repository ◽  
Waste Forms ◽  
The Earth ◽  
Soluble Solid ◽  
High Level

The reprocessing of spent fuel from nuclear reactors and processing of fuels for defense purposes have generated large volumes of high-level liquid waste that need to be immobilized prior to final storage. For immobilization, the wastes must be converted to a less soluble solid, and, although other waste forms exist, glass currently appears to be the choice for the transuranic-containing portion of the reprocessed waste. Once produced, this glass will be sent in canisters to a geologic repository located some 200 to 500 m below the surface of the earth.

Characterization of Mechanical Strengths for Simulated High-Level Waste Forms

1989 ◽  
Vol 176 ◽  
Author(s):  
Hiroshi Igarashi ◽  
Takeshi Takahashi
Keyword(s):  
Fracture Toughness ◽  
Glass Ceramic ◽  
Bending Strength ◽  
Liquid Waste ◽  
High Level Waste ◽  
Spent Fuel ◽  
High Level Liquid Waste ◽  
Waste Forms ◽  
High Level ◽  
Mechanical Strengths

ABSTRACTWaste forms have been developed and characterized at PNC (Power Reactor and Nuclear Fuel Development Corporation)to immobilize high-level liquid waste generated from the reprocessing of nuclear spent fuel.Mechanical strength tests were excecuted on simulated solidified highlevel waste forms which were borosilicate glass and diopside glass-ceramic. Commercial glass was tested for comparison. Measured strengths were three-point bending strength,uniaxial compressive strength,impact strength by falling weight method,and Vickers hardness. Fracture toughness and fracture surface energy were also measured by both notch-beam and indentation technique.The results show that mechanical strengths of waste glass form are similar and that the glass ceramic form has the higher fracture toughness.

The Application Research of Concentration & Denitration Technology in Advanced Reprocessing Process

2012 ◽  
Vol 560-561 ◽  
pp. 637-643
Author(s):  
Yong Li ◽  
Xue Gang Liu ◽  
Jin Chen
Keyword(s):  
Nitric Acid ◽  
Nuclear Power ◽  
Liquid Waste ◽  
Fission Products ◽  
Extraction Process ◽  
Purex Process ◽  
Current Status ◽  
Spent Fuel ◽  
High Level Liquid Waste ◽  
High Level

The proper management of spent fuel arising from nuclear power production is a key issue for the sustainable development of nuclear energy. While conventional reprocessing process, PUREX process, was successful to recover uranium and plutonium, in recent years some countries have turned to focus on advanced reprocessing process, which features of partitioning of minor actinides (MA) and long-lived fission products(LLFP). Most advanced reprocessing processes under development involve new extractants and additional extraction cycles. In China, TRPO extraction process has been developed to partition MA/LLFP from high-level liquid waste(HLLW) since early 1980’s. In parallel to R&D work on separation technologies, studies on concentration & denitration process have been evolved to prepare feed solutions to suit qualifications of extraction. Industrially, concentration & denitration is the internationally recognized standard to treat HLLW released from PUREX before vitrification. It enables to minimize the volume of interim storage, to restrain the corrosion of storage tank, to recover nitric acid in HLLW and to reduce the required evaporation duty of the vitrification process. Generally, the constitution of concentrated HLLW has little impact on the following vitrification process. But when concentration & denitration acts as pretreatment process of partitioning, the composition of actinides, fission products, and nitric acid in concentrated HLLW solution plays significant role in extraction process. A series of technical issues relevant to the connection between concentration ﹠denitration and extractions should be solved. This paper describes current status of concentration & denitration technology utilized in industry and under reprocessing plants. The specific separation requirements in advanced reprocessing process and challenges to apply concentration & denitration process are addressed. Besides, concentration & denitration process was tested in laboratory to adjust feed solutions for TRPO and Cyanex301 partitioning. Results demonstrate its promising prospect in advanced reprocessing process.

Advanced ORIENT Cycle: Progress on Fission Product Separation and Utilization

2010 ◽  
Author(s):  
Isao Yamagishi ◽  
Masaki Ozawa ◽  
Hitoshi Mimura ◽  
Shohei Kanamura ◽  
Koji Mizuguchi
Keyword(s):  
Fuel Cycle ◽  
Liquid Waste ◽  
Fission Products ◽  
Spent Fuel ◽  
High Level Liquid Waste ◽  
Pt Electrode ◽  
Acid Media ◽  
Electrolytic Hydrogen ◽  
Product Separation ◽  
High Level

Fission reaction of U-235 and/or plutonium generates more than 40 elements and 400 nuclides in the spent fuel. Among them, 31 elements are categorized as rare metals. In a conventional fuel cycle U and Pu are reused but others are vitrified for disposal. Adv.-ORIENT (Advanced Optimization by Recycling Instructive Elements) Cycle strategy was drawn up for the minimization of radio-toxicity and volume of radioactive waste as well as the utilization of valuable elements/nuclides in the waste. The present paper describes the progress on Fission Products (FP) separation in this Cycle. Highly functional inorganic adsorbent (AMP-SG, silica gel loaded with ammonium molybdophosphate) and organic microcapsule (CE-ALG, alginate gel polymer enclosed with crown ether D18C6) were developed for separation of heat-generating Cs and Sr nuclides, respectively. The AMP-SG adsorbed more than 99% of Cs selectively from a simulated High-level Liquid Waste (HLLW). The ALG microcapsule adsorbed 0.0249 mmol/g of Sr and exhibited the order of its selectivity; Ba > Sr > Pd >> Ru > Rb > Ag. The electrodeposition is advantageous for both recovery and utilization of PGMs (Ru, Rh, Pd) and Tc because PGMs are recovered as metal on Pt electrode. Among PGMs, Pd was easily deposited on the Pt electrode. In the presence of Pd or Rh the reduction of Ru and Tc was accelerated more in hydrochloric acid media than in nitric acid. In the simulated HLLW, the redox reaction of Fe(III)/Fe(II) disturbed deposition of elements except for Pd. The deposits on Pt electrode showed higher catalytic reactivity on electrolytic hydrogen production than the original Pt electrode. The reactivity of deposits prepared from the simulated HLLW was higher than that from solution containing only PGM.

Solids Formation Behavior of Simulated High-Level Liquid Waste During Long-Term Storage

2010 ◽  
Author(s):  
Jin Chen ◽  
Xuegang Liu ◽  
Yanchao Zhang ◽  
Qian’ge He ◽  
Jianchen Wang
Keyword(s):  
Solvent Extraction ◽  
Liquid Waste ◽  
Chemical Separation ◽  
Ion Concentration ◽  
Spent Fuel ◽  
High Level Liquid Waste ◽  
Long Term Storage ◽  
High Level ◽  
Term Storage

High-level liquid waste (HLLW) generated from reprocessing process contains actinides, lanthanides, fission products (FP) and a significant amount of nitrate ion. The partitioning and transmutation concept has been introduced for reducing the long-term hazards of HLLW. Several chemical separation processes mainly based on solvent extraction methods have been proposed to treat HLLW. However, solids consisting mainly Mo and Zr are known to form in HLLW during its long-term storage, Solid formations influence the composition of HLLW and the downstream solvent extraction process. To understand the precipitation behavior and stability of HLLW during its long-term storage, simulated HLLW (prepared as raffinate solution from LWR spent fuel reprocessing, 1AW solution) was prepared. Preliminary studies on solid formation behaviors with regard to the precipitation formation during refluxing and aging (representing a long-term storage) were carried out. Precipitation kinetics of major FPs such as Zr, Mo, Ru, rare earth elements, and etc. have been studied; The effect of phosphate ion concentration and temperature on solids formation were also experimentally examined. The formation conditions and the mechanism of solids were discussed.

Selective Uptake of Palladium From High-Level Liquid Wastes by Hybrid Microcapsules Enclosed With Insoluble Ferrocyanides

2010 ◽  
Author(s):  
Hitoshi Mimura ◽  
Takashi Sakakibara ◽  
Wu Yan ◽  
Yuichi Niibori ◽  
Shin-ichi Koyama ◽  
...  
Keyword(s):  
Sol Gel ◽  
Liquid Waste ◽  
Selective Separation ◽  
Spent Fuel ◽  
High Level Liquid Waste ◽  
Hno3 Concentration ◽  
Liquid Wastes ◽  
Gel Polymers ◽  
High Level ◽  
Cellulose Filter

Fine crystalline powders of KCuFC were immobilized with alginate gel polymers by sol-gel methods. The uptake properties of KCuFC-microcapsules (KCuFC-MC) were examined by batch and column methods. The size of KCuFC-MC particle was estimated to be about 1 mm in diameter, and KCuFC powders were uniformly dispersed in KCuFC-MC particles. The uptake rate of Pd2+ for KCuFC-MC was attained within 3 d, and the uptake of Pd2+ was found to be independent of the temperature and coexisting HNO3 concentration. As for the breakthrough properties of Pd2+ through a column packed with KCuFC-MC, a breakpoint of 5% breakthrough was enhanced with lowering of flow rate and independent of coexisting HNO3 concentration. The Pd2+ ions were selectively adsorbed in the KCuFC crystal phase, while other metal ions such as Ru(NO)3+ and ZrO2+ were absorbed in the alginate phase. High uptake percentage of 98.6% was obtained by using the dissolved solutions of spent fuel from FBR-JOYO (119 GWd/t, JAEA). The alginate film enclosing KZnFC was further prepared by using the support of cellulose filter paper, where the Pd2+ ions were selectively adsorbed on the KZnFC-MC film. The alginate film enclosing insoluble ferrocyanides are predicted for the selective separation of Pd2+ as an ion-exchange filter. Thus, the microcapsules enclosing insoluble ferrocyanides are effective for the selective separation of Pd2+ from high-level liquid waste (HLLW).

High Level Liquid Waste Solidification Using a “Cold” Crucible Induction Melter

2000 ◽  
Vol 663 ◽  
Author(s):  
Andrei V. Demine ◽  
Nina V. Krylova ◽  
Pavel P. Polyektov ◽  
Igor N. Shestoperov ◽  
Tatyana V. Smelova ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Liquid Waste ◽  
Cold Crucible ◽  
High Level Liquid Waste ◽  
Two Stage ◽  
One Stage ◽  
Waste Forms ◽  
High Level ◽  
Ceramic Melter

ABSTRACTAt the present time the primary problem in a closed nuclear fuel cycle is the management of high level liquid waste (HLLW) generated by the recovery of uranium and plutonium from spent nuclear fuel. Long-term storage of the HLLW, even in special storage facilities, poses a real threat of ecological accidents. This problem can be solved by incorporating the radioactive waste into solid fixed forms that minimize the potential for biosphere pollution by long-lived radionuclides and ensure ecologically acceptable safe storage, transportation, and disposal. In the present report, the advantages of a two-stage HLLW solidification process using a “cold” crucible induction melter (CCIM) are considered in comparison with a one-stage vitrification process in a ceramic melter.This paper describes the features of a process and equipment for a two-stage HLLW solidification technology using a “cold” crucible induction melter (CCIM) and identifies the advantages compared to a one-stage ceramic melter. A two-stage pilot facility and the technical characteristics of the equipment are described using a once-through evaporator and cold-crucible induction melter currently operational at the IA.Mayak. facility in Ozersk, Russia. The results of pilot-plant tests with simulated HLLW to produce a phosphate glass are described. Features of the new mineral-like waste form matrices synthesized by the CCIM method are also described. Subject to further development, the CCIM technology is planned to be used to solidify all accumulated HLLW at Mayak – first to produce borosilicate glass waste forms and then mineral-like waste forms.

Effects of Noble Metal Elements on Viscosity and Electrical Resistivity of Simulated Vitrified Products for High-Level Liquid Waste

1991 ◽  
Vol 257 ◽  
Author(s):  
Hiroshi Igarashi ◽  
Kazuhiro Kawamura ◽  
Takeshi Takahashi
Keyword(s):  
Electrical Resistivity ◽  
Noble Metal ◽  
Nuclear Waste ◽  
Noble Metals ◽  
Liquid Waste ◽  
Waste Glass ◽  
High Level Liquid Waste ◽  
Metal Elements ◽  
Nuclear Waste Glass ◽  
High Level

ABSTRACTThe effects of noble metal elements such as ruthenium, rhodium and palladium on the viscosity and electrical resistivity of simulated nuclear waste glass were studied. The glass enriched with noble metals showed the viscosity of a non-Newtonian fluid. The viscosity of the waste glass with 10 wt% Ru0 2 was 3 to 7 times higher than that of glass without noble metals. The RuO2 was mainly responsible for the increase in viscosity for the glass enriched with noble metals. Electrical resistivity of the glass with 15 wt% RuO2 was one seventh to two orders of magnitude lower than that of glass without noble metals. The three noble metals contributed to the decrease in resistivity. The quantitative effects of noble metals on these properties were obtained.

scholarly journals Ab Initio Investigation of Helium Mobility in La2Zr2O7 Pyrochlore

Crystals ◽  
2021 ◽  
Vol 11 (6) ◽  
pp. 667
Author(s):  
Yanxia Lu ◽  
Qing Peng ◽  
Chenguang Liu
Keyword(s):  
Nuclear Waste ◽  
Density Functional ◽  
Minimum Energy ◽  
Elastic Band ◽  
Waste Forms ◽  
Band Method ◽  
And Migration ◽  
High Level ◽  
Migration Pathways ◽  
Nuclear Waste Forms

The α-decay of incorporated actinides continuously produces helium, resulting in helium accumulation and causing security concerns for nuclear waste forms. The helium mobility is a key issue affecting the accumulation and kinetics of helium. The energy barriers and migration pathways of helium in a potential high-level nuclear waste forms, La2Zr2O7 pyrochlore, have been investigated in this work using the climbing image nudged elastic band method with density functional theory. The minimum energy pathway for helium to migrate in La2Zr2O7 is identified as via La–La interstitial sites with a barrier of 0.46 eV. This work may offer a theoretical foundation for further prospective studies of nuclear waste forms.

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