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

1989 ◽  
Vol 120 (1) ◽  
pp. 21-23 ◽  
Author(s):  
Jiri Strnad ◽  
Klaus Heckmann ◽  
Klaus Gompper
Keyword(s):  
Liquid Waste ◽  
Intermediate Level

Progress at ANSTO on a Synroc Plant for Intermediate-Level Waste from Reactor Production of 99Mo

2014 ◽  
Vol 94 ◽  
pp. 111-114 ◽  
Author(s):  
Eric R. Vance ◽  
S.A. Moricca ◽  
M.W.A. Stewart
Keyword(s):  
Room Temperature ◽  
Hot Isostatic Pressing ◽  
Liquid Waste ◽  
Fission Products ◽  
Intermediate Level ◽  
Waste Form ◽  
Production Plant ◽  
Reducing Atmosphere ◽  
Nuclear Waste Forms ◽  
Reactor Production

Intermediate level waste from ANSTO’s expanded 99Mo production plant will consist of ~5000L/year of 6M NaOH + 1.4 NaAlO2 + fission products. Detailed engineering is being carried out on a synroc plant to immobilise this waste in a glass-ceramic, with completion scheduled for 2016. The liquid waste will be mixed with precursors and dried before being calcined in a reducing atmosphere to control fission product volatility. The calcine will be transferred to 30L metal cans which will be hot isostatically pressed at 1000°C/30MPa for 2h, then cooled to room temperature and stored preparatory to final disposal. Laboratory scale waste form material will pass 90°C PCT tests. In addition, legacy intermediate level uranyl nitrate-based liquid waste from 99Mo production at ANSTO between the 1980s and 2005 via irradiation of UO2 targets will also be immobilised by the same process to form a Synroc-type waste form. Some examples illustrating the wide applicability of hot isostatic pressing to consolidate nuclear waste forms will be given showing the advantages for particular wastes, notably high waste loadings and the absence of off-gas in the high temperature consolidation step. The immobilisation of a variety of low-level liquid and solid wastes from 99Mo production will also be discussed.

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

1999 ◽  
Vol 242 (2) ◽  
pp. 341-348 ◽  
Author(s):  
P. K. Sinha ◽  
P. M. Satyasai ◽  
R. Shankar ◽  
R. Muthiah ◽  
S. Bera ◽  
...  
Keyword(s):  
Ion Exchange ◽  
Liquid Waste ◽  
Intermediate Level

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

MRS Advances ◽  
2018 ◽  
Vol 3 (20) ◽  
pp. 1123-1131 ◽  
Author(s):  
Supphatuch Ukritnukun ◽  
Charles Christopher Sorrell ◽  
Daniel Gregg ◽  
Eric R. Vance ◽  
Pramod Koshy
Keyword(s):  
Nuclear Waste ◽  
Raw Materials ◽  
Liquid Waste ◽  
High Energy ◽  
Intermediate Level ◽  
Curing Temperature ◽  
Mix Design ◽  
Nuclear Waste Storage ◽  
Waste Storage ◽  
Main Component

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.

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

1988 ◽  
Vol 127 ◽  
Author(s):  
Bert-G. Brodda ◽  
Xu Mingxia
Keyword(s):  
Ph Value ◽  
Liquid Waste ◽  
Acrylic Resin ◽  
Intermediate Level ◽  
Cement Matrix ◽  
Cement Slurry ◽  
Cement Ratio ◽  
Diffusion Controlled ◽  
Leaching Behaviour ◽  
Sulfide Ions

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.

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