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.

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.

Adsorption Materials Development for the Separation of Actinides and Specific Fission Products from High Level Waste

2014 ◽  
Vol 94 ◽  
pp. 103-110 ◽  
Author(s):  
Yue Zhou Wei ◽  
Shun Yan Ning ◽  
Qi Long Wang ◽  
Zi Chen ◽  
Yan Wu ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
Selective Adsorption ◽  
Packed Column ◽  
Liquid Waste ◽  
Separation Method ◽  
High Level Waste ◽  
Small Scale ◽  
Fast Diffusion ◽  
High Level Liquid Waste ◽  
High Level

The long-term radiotoxicity of high level liquid waste (HLLW) generated in spent nuclear fuel reprocessing is governed by the content of several long-lived minor actinides (MA) and some specific fission product nuclides. To efficiently separate MA (Am, Cm) and some FPs such as Cs and Sr from the HLLW, we have been studying an advanced aqueous partitioning process, which uses selective adsorption as separation method. In this work, we prepared different types of porous silica-based organic/inorganic adsorbents with fast diffusion kinetics, improved chemical stability and low pressure drop in a packed column. So they are advantageously applicable to efficient separation of the MA and specific FP elements from HLLW. Adsorption and separation behaviors of the MA and some FP elements such as Cs and Sr were studied. Small scale separation tests using simulated and genuine nuclear waste solutions were carried out and the obtained results indicate that the proposed separation method based on selective adsorption is essentially feasible.

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.

The Strategy of the Belgian Nuclear Research Centre in the Area of High-Level Waste Form Compatibility Research

2007 ◽  
Author(s):  
Karel Lemmens ◽  
Christelle Cachoir ◽  
Elie Valcke ◽  
Karine Ferrand ◽  
Marc Aertsens ◽  
...  
Keyword(s):  
Nuclear Research ◽  
Waste Form ◽  
Research Centre ◽  
High Level Waste ◽  
Spent Fuel ◽  
National Agency ◽  
Waste Forms ◽  
In Situ Experiments ◽  
High Level

The Belgian Nuclear Research Centre (SCK•CEN) has a long-standing expertise in research concerning the compatibility of waste forms with the final disposal environment. For high level waste, most attention goes to two waste forms that are relevant for Belgium, namely (1) vitrified waste from the reprocessing of spent fuel, and (2) spent fuel as such, referring to the direct disposal scenario. The expertise lies especially in the study of the chemical interactions between the waste forms and the disposal environment. This is done by laboratory experiments, supported by modeling. The experiments vary from traditional leach tests, to more specific tests for the determination of particular parameters, and highly realistic experiments. This results in a description of the phenomena that are expected upon disposal of the waste forms, and in quantitative data that allow a conservative long-term prediction of the in situ life time of the waste form. The predictions are validated by in situ experiments in the underground research laboratory HADES. The final objective of these studies, is to estimate the contribution of the waste form to the overall safety of the disposal system, as part of the Safety and Feasibility Case, planned by the national agency ONDRAF/NIRAS. The recent change of the Belgian disposal concept from an engineered barrier system based on the use of bentonite clay to a system based on a concrete buffer has caused a reorientation of the research programme. The expertise in the area of clay-waste interaction will however be maintained, to develop experimental methodologies in collaboration with other countries, and as a potential support to the decision making in those countries where a clay based near field is still the reference. The paper explains the current R&D approach, and highlights some recent experimental set-ups available at SCK•CEN for this purpose, with some illustrating results.

The IAEA Coordinated Research Programme on the Performance of High-Level Waste Forms and Packages Under Repository Conditions

1992 ◽  
Vol 294 ◽  
Author(s):  
Vladimir S. Tsyplenkov
Keyword(s):  
Near Field ◽  
Research Programme ◽  
High Level Waste ◽  
Future Research ◽  
Spent Fuel ◽  
Waste Forms ◽  
Safety Assessments ◽  
High Level ◽  
Coordination Meeting

ABSTRACTThe IAEA initiated, in 1991, a Coordinated Research Programme (CRP), with the aim of promoting the exchange of information on the results obtained by different countries in the performance of high-level waste forms and waste packages under conditions relevant to final repository. These studies are being undertaken to obtain reliable data as input to safety assessments and environmental impact analyses, for final disposal purposes. The CRP includes studies on waste forms that are presently of interest worldwide: borosilicate glass, Synroc and spent fuel.Ten laboratories leading in investigation of high-level waste form performance have already joined the programme. The results of their studies and plans for future research were presented at the first Research Coordination Meeting, held in Karlsruhe, Germany, in November 1991. The technical contributions concentrated on effecting an understanding of dissolution mechanisms of waste forms under simulated repository conditions. A quantitative interpretation of the chemical processes in the near field is considered a prerequisite for long-term predictions and for the formulation of a "source term" for performance assessment studies.

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).

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