Removal of ruthenium from high-level radioactive liquid waste generated during reprocessing of spent fuel

2013 ◽  
Vol 52 (1-3) ◽  
pp. 514-525 ◽  
Author(s):  
Khushboo Singh ◽  
N.L. Sonar ◽  
T.P. Valsala ◽  
Y. Kulkarni ◽  
Tessy Vincent ◽  
...  
Keyword(s):  
Liquid Waste ◽  
Spent Fuel ◽  
Radioactive Liquid Waste ◽  
High Level

Radiation Effects on Hollandite Ceramics developed for Radioactive Cesium Immobilization

2003 ◽  
Vol 792 ◽  
Author(s):  
V. Aubin ◽  
D. Caurant ◽  
D. Gourier ◽  
N. Baffier ◽  
S. Esnouf ◽  
...  
Keyword(s):  
Radiation Effects ◽  
Liquid Waste ◽  
Fission Products ◽  
Radioactive Cesium ◽  
Radioactive Liquid Waste ◽  
Epr Spectrum ◽  
Paramagnetic Resonance ◽  
Γ Radiation ◽  
The Stability ◽  
High Level

ABSTRACTProgress on separating the long-lived fission products from the high level radioactive liquid waste (HLW) has led to the development of specific host matrices, notably for the immobilization of cesium. Hollandite (nominally BaAl2Ti6O16), one of the main phases constituting Synroc, receives renewed interest as specific Cs-host wasteform. The radioactive cesium isotopes consist of short-lived Cs and Cs of high activities and Cs with long lifetime, all decaying according to Cs+→Ba2++e- (β) + γ. Therefore, Cs-host forms must be both heat and (β,γ)-radiation resistant. The purpose of this study is to estimate the stability of single phase hollandite under external β and γ radiation, simulating the decay of Cs. A hollandite ceramic of simple composition (Ba1.16Al2.32Ti5.68O16) was essentially irradiated by 1 and 2.5 MeV electrons with different fluences to simulate the β particles emitted by cesium. The generation of point defects was then followed by Electron Paramagnetic Resonance (EPR). All these electron irradiations generated defects of the same nature (oxygen centers and Ti3+ ions) but in different proportions varying with electron energy and fluence. The annealing of irradiated samples lead to the disappearance of the latter defects but gave rise to two other types of defects (aggregates of light elements and titanyl ions). It is necessary to heat at relatively high temperature (T=800°C) to recover an EPR spectrum similar to that of the pristine material. The stability of hollandite phase under radioactive cesium irradiation during the waste storage is discussed.

Studies on hydrolysis and radiolysis of N,N,N′,N′-tetraoctyl-3-oxapentane-1,5-diamide

2002 ◽  
Vol 90 (3) ◽  
Author(s):  
Y. Sugo ◽  
Y. Sasaki ◽  
S. Tachimori
Keyword(s):  
Gamma Irradiation ◽  
Room Temperature ◽  
Degradation Products ◽  
Liquid Waste ◽  
Amide Bond ◽  
Radioactive Liquid Waste ◽  
Ether Bond ◽  
Nuclear Fuel Reprocessing ◽  
High Level ◽  
Fuel Reprocessing

SummaryHydrolytic and radiolytic stabilities of a promising extractant, N,N,N′,N′-tetraoctyl-3-oxapentane-1,5-diamide (TODGA), for actinides in high-level radioactive liquid waste from nuclear fuel reprocessing were investigated in air at room temperature. Hydrolysis by nitric acid was not observed, whereas radiolysis by gamma irradiation was notably observed. The radiolysis study showed that an amide-bond, an ether-bond, and a bond adjacent to the ether-bond tended to be broken by gamma irradiation, and dioctylamine and various N,N-dioctylmonoamides were identified as the main degradation products by GC/MS and NMR analyses. The

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.

scholarly journals The extraction of SR2+ with dicyclohexano-18-crown-6 in conventional organic solvent and ionic liquid diluents

2020 ◽  
Vol 85 (7) ◽  
pp. 909-922 ◽  
Author(s):  
Zheng Wei ◽  
Yang Gao ◽  
Yu Zhou ◽  
Caishan Jiao ◽  
Meng Zhang ◽  
...  
Keyword(s):  
Ionic Liquid ◽  
Nitric Acid ◽  
Distribution Ratio ◽  
Nitric Acid Concentration ◽  
Liquid Waste ◽  
Fission Products ◽  
Carbon Chain ◽  
Radioactive Liquid Waste ◽  
Solvent Systems ◽  
High Level

90Sr (t1/2 = 28.8 a), one of the most significant fission products in high-level radioactive liquid waste (HLLW), contributes to a large part of the heat load and radiation. Removal of 90Sr from the HLLW is beneficial for the final treatment of nuclear waste. In this paper, the extraction of Sr2+ was carried out using dicyclohexano-18-crown-6 (DCH18C6) in a variety of diluents including conventional organic solvents and novel ionic liquid solvents. The effect of several factors, such as nitric acid concentration, crown ether concentration and initial strontium concentration on the extraction of Sr2+ have been studied comprehensively. The higher distribution ratio and the stripping efficiency of Sr2+ were obtained with the binary diluents consisted of n-octanol and acetylene tetrachloride, which were compared with that using pure n-octanol as diluent. As for the CnmimNTf2 (n = 2, 4, 6) ionic liquid solvent systems, the distribution ratio of Sr2+ was much higher in the nitric acid medium with low concentration than in the traditional solvent systems. The results showed that DC2mimNTf2 > DC4mimNTf2 > DC6mimNTf2, which indicated that shorter carbon chain benefits the extraction of Sr2+.

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.

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