Authors:M. Murali, A. Bhattacharayya, D. Raut, A. Kar, B. Tomar, and V. Manchanda
The high level waste (HLW) generated from the reprocessing of the spent fuel of pressurized heavy water reactor has been characterized
for the minor actinides. The radiation dose of the waste solution was reduced by radiochemical separation of cesium from HLW
by solvent extraction with chlorinated cobalt dicarbollide dissolved in 20% nitrobenzene in xylene. Minor actinides (Np, Pu,
Am, Cm) in the high level waste were assayed by alpha spectrometry following radiochemical separation. The gross alpha activity
determined by liquid scintillation agrees well (within 10%) with the cumulative quantities of actinides determined by alpha
Authors:T. L. White, T. L. White, K. B. Martin, K. B. Martin, L. N. Oji, L. N. Oji, D. P. DiPrete, D. P. DiPrete, and W. R. Wilmarth
One waste remediation process used at the Savannah River Site was the in-tank precipitation of the beta-emitting 137Cs from high-level waste (HLW) using sodium tetraphenylborate (NaTPB) followed by processing the resulting decontaminated
filtrate into grout at the Saltstone Production Facility (SPF). A simple method was developed for the monitoring of tetraphenylborate
(TPB) in high-level waste (HLW) containing up to 0.38 Ci/gal of 137Cs. Separation was achieved by extraction of the high sodium-bearing waste with acetonitrile followed by analysis using reversed-phase
high performance liquid chromatography (HPLC). The sample preparation method allowed for the handling of an organic extraction
layer that had 94% less acitivity than the HLW sample. The subsequent HPLC analysis of the extraction layer determined the
TPB concentration in HLW waste to 0.8 mg/l with a %rsd of 8.
Authors:W. Wilmarth, S. Rosencrance, C. Nash, D. DiPrete, and C. DiPrete
Hanford High Level Waste will require processing to reduce the concentration of various actinide elements prior to encapsulation into low activity glass for disposition. High level waste at Hanford contains elevated actinide concentrations in the supernate because of organic complexants present in the tanks. Traditional removal strategies are not viable processing sequences for the Hanford tanks containing complexants. Reported here is a novel actinide decontamination strategy. This pretreatment sequence consists of addition of calcium nitrate and sodium permanganate. The observed optimum decontamination efficiencies for plutonium and americium are more than 99.5%.
Authors:R. Chitnis, P. Wattal, A. Ramanujam, P. Dhami, V. gopalakrishnan, A. Bauri, and A. Banerji
These studies are an extension of carlier work on the recovery of actinides extracted by Truex solvent from simulated high
level waste solution, with a mixture of weak acid, weak base and complexing agent used as a strippant. The effectiveness of
the proposed strippant, consisting of formic acid, hydrazine hydrate and citric acid, is tested in a counter-current mode
using mixer-settler in the present studies. The studies show that near quantitative recovery of americium and plutonium is
feasible from acid-bearing Truex solvent with no reflux (reextration) of activity. Use of this strippant reduces considerably
the generation of secondary waste.
Authors:R. Chitnis, P. Wattal, A. Ramanujam, P. Dhami, V. Gopalakrishanan, A. Bauri, and A. Banerji
This work deals with the batch studies on stripping of actinides extracted by a mixture octyl(phenyl)-N,N-diisobutyl-carbamoylmethylphosphine
oxide (CMPO) and tri-n-butyl phosphate (TBP) in n-dodecane (Truex solvent) from simulated high level waste (HLW) solution.
The stripping of americium and plutonium from acid-bearing CMPO-TBP mixture is carried out using a mixture of weak acid, weak
base and complexing agent as strippant. A mixture of formic acid, hydrazine hydrate and citric acid appeared to be best suited
for efficient stripping of americium and plutonium. With appropriate modifications in the concentration of individual constitutents,
this strippant can be used for the recovery of actinides from loaded Truex solvent with any acid content.
Authors:Chhavi Agarwal, P. Kalsi, D. Prabhu, P. Pathak, and V. Manchanda
The estimation of low level alpha activity is difficult in waste samples containing large concentration of salts and beta–gamma
activity. In the present study, the feasibility of gross alpha-activity measurement for simulated high level waste (SHLW)
in solution medium by alpha-track registration technique has been attempted. The results showed that it is possible to use
this technique for gross alpha-activity estimation of ~200 Bq/mL in solution medium with a precision and accuracy of ~30%.
The importance of measuring 200 Bq/mL alpha activity in SHLW solutions is that this value corresponds to about 4,000 Bq/g
activity in the solid medium which is the safe disposable limit. The advantage of this method over other methods is that it
is not sensitive to beta–gamma emitters and salts and is very simple and inexpensive.
Authors:A. Ramanujam, P. Achuthan, P. Dhami, R. Kannan, V. Gopalakrishnan, V. Kansra, R. Iyer, and K. Balu
A generator system has been developed for the preparation of carrier-free 90Y from 90Sr present in the high level waste (HLW) of the Purex process by employing a supported liquid membrane (SLM) using 2-ethylhexyl-2-ethylhexyl phosphonic acid (KSM-17 equivalent to PC 88A) supported on a polytetrafluoro ethylene (PTFE) membrane. When uranium depleted Purex HLW at appropriate acidity is passed sequentially through octyl (phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) sorbed on chromosorb-102 (abbreviated as CAC) and Zeolite AR1 (synthetic mordenite) columns, all the trivalent, tetravalent and hexavalent metal ions and monovalent 137Cs ions are sorbed. After adjusting to pH 2 with NaOH the resulting effluent is used as feed in a single stage membrane cell partitioned with a PTFE membrane impregnated with KSM-17 and having a feed and receiver compartment with 5.0 ml capacity each. The receiver compartment was filled with a 0.5M HNO3 or 0.5M HCl stripping solution. 90Y alone is preferentially transported across the membrane leaving behind all the impurities viz. 90Sr, 125Sb, 106Ru, 106Rh, etc. in the feed compartment. This technique can yield 90Y in mCi levels in a pure and carrier-free form for medical applications. The feed can be reused repeatedly after allowing for 90Y buildup.
Authors:V. Gopalakrishnan, P. Dhami, A. Ramanujam, M. Krishna, M. Murali, J. Mathur, R. Iyer, A. Bauri, and A. Banerji
Bench-Scale studies on the partitioning and recovery of minoractinides from the actual and synthetic sulphate-bearing high level waste (SBHLW) solutions have been carried out by giving two contacts with 30% TBP to deplete uranium content followed by four contacts with 0.2M CMPO+1.2M TBP in dodecane. The acidity of the SBHLW solutions was about 0.3M. In the case of actual SBHLW, the final raffinate contained about 0.4% -activity originally present in the HLW, whereas with synthetic SBHLW the -activity was reduced to the background level.144Ce is extracted almost quantitative in the CMPO phase,106Ru about 12% and137Cs is practically not extracted at all. The extraction chromatographic column studies with synthetic SBHLW (aftertwo TBP contacts) has shown that large volume of waste solutions could be passed through the column without break-through of actinide metal ions. Using 0.04M HNO3>99% Am(III) and rare earths could be eluted/stripped. Similarly >99% Pu(IV) and U(VI) could be eluted.stripped using 0.01M oxalic acid and 0.25M sodium carbonate, respectively. In the presence of 0.16M SO
(in the SBHLW) the complex ions AmSO
, UO2SO4, PuSO
and Pu(SO4)2 were formed in the aqueous phase but the species extracted into the organic phase (CMPO+TBP) were only the nitrato complexes Am(NO3)3·3CMPO, UO2(NO3)2·2CMPO and Pu(NO3)4·2CMPO. A scheme for the recovery of minor actinides from SBHLW solution with two contacts of 30% TBP followed by either solvent extraction or extraction chromatographic techniques has been proposed.
Authors:M. Murali, D. Raut, D. Prabhu, P. Mohapatra, B. Tomar, and V. Manchanda
Efficacy of chlorinated cobaltdicarbollide in a modified diluent, 20% nitrobenzene in xylene was tested for the extraction
and recovery of Cs from simulated high-level waste (HLW) solutions generated from PHWR-fuel reprocessing. Concentration of
the reagent, composition of the diluent, numbers of contacts, the nature of stripping agents are some of the parameters optimized
for the complete removal of Cs from such waste solutions. The above solvent extraction procedure can be applied to genuine
HLW solutions for effective reduction of the dose due to Cs so that HLW can be handled in fume hoods for its characterization.
We describe tests of EiChrom Industries' Ni-selective ion exchange resin for use in analysis of63Ni in Savannah River Site high level waste. We report measurement of63Ni content in two sets of Savannah River Site glass product from the Defense Waste Processing Facility. The63Ni β-decay activity was chemically separated in quintuplicate from fission product and plutonium α-β activities of up to 103 times the observed63Ni content. The separation used a Ni-dimethlyglyoxime precipitation followed by radiochemical purification using the Ni-selective
extraction chromatography resin. Further removal of interfering activity was based on diagnosis of observed radioactivity
in each successive product phase. We analyzed eleven plant glass product samples using seven spiked standard addition duplicates
to measure63Ni recovery in the separations. Selected liquid scintillation β-decay spectra are shown to validate the method. Interpretation
of accuracy is based upon three distinct comparisions to predicted63Ni content.