Alpha radiolytic degradation of N,N-dihexyl octanamide P. N. Pathak • D. R. Prabhu • M. Bindu • S. C. Tripathi • V. K. Manchanda Received: 7 October 2010 / Published online: 9 November 2010 Ó Akade ´miai Kiado ´, Budapest, Hungary 2010 Abstract The primary purpose of this study was to understand the alpha radiolytic degradation behavior of N,N-dihexyl octanamide (DHOA) vis a vis tributyl phos- phate (TBP) solutions in n-dodecane under plutonium loading conditions. These studies were carried out as a function of dose on different Pu loaded samples (contain- ing 0.002-10 g/L Pu) from 4 M HNO 3 medium. These Pu loaded solutions were evaluated for stripping behavior by contacting with 0.5 M NH 2 OH at 0.5 M HNO 3 solutions. Organic phase analysis was carried out by gas chroma- tography (GC) and by visible spectrophotometry. These studies clearly indicated that Pu stripping becomes difficult with increased dose in the case of TBP system. On the other hand, no such problem was observed in DHOA system during stripping of plutonium, thereby indicating that DHOA is a promising candidate for the reprocessing of high burn up Pu rich spent fuels. Keywords Tributyl phosphate Á Dihexyloctanamide Á Plutonium Á Degradation Introduction The challenging task of recovery and purification of Pu from irradiated U and of 233 U from irradiated Th is accomplished conventionally by the well known tri-n-butyl phosphate (TBP) based PUREX and THOREX processes, respectively [1]. Whereas TBP is the workhorse of tech- nologists engaged in spent fuel reprocessing, extensive experience with this extractant has revealed that there are few drawbacks associated with the use of TBP, which are of concern to the separation scientists and technologists. Prominent amongst them are: high aqueous solubility, poor chemical/radiation stability and interference of degradation products during stripping of Pu/U, poor decontamination factor (DF) values of Pu/U with respect to fission products, low limiting organic concentration (LOC) of Pu(IV) and Th(IV), and generation of large volume of secondary (phosphate) waste. These problems are of particular con- cern during the reprocessing of short-cooled thermal reactor fuels as well as Pu rich fast reactor fuels. Holland et al., carried out a systematic gamma radiolytic degrada- tion study on TBP/n-dodecane [2]. The irradiated samples were analyzed for gaseous and acidic degradation products. The gaseous products observed were hydrogen and meth- ane and acidic products were dibutyl phosphate (DBP) and monobutyl phosphate (MBP). May et al. reported that distribution of Pu(IV) between 0.5–5 M HNO 3 and 1.1 M TBP/decane solution in the presence of DBP (0–0.05 M) and U(VI) (0–0.42 M) was significantly influenced by the direction (forward/back- ward), and time of equilibration [3]. As the contact time was increased, increasing amount of Pu(IV) was stripped back into the aqueous phase. There was a marked decrease in extractability of Pu(IV) from the aqueous phase con- taining DBP. This effect was more pronounced at lower initial aqueous HNO 3 concentrations and in several cases equilibrium could not be attained even after 1 h contact. This poor kinetics was attributed to the formation of inextractable Pu(IV)- DBP complexes along with poly- meric hydrolytic species of Pu(IV) [4, 5]. P. N. Pathak Á D. R. Prabhu Á V. K. Manchanda (&) Radiochemistry Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085, India e-mail: vkm@barc.gov.in M. Bindu Á S. C. Tripathi Fuel Reprocessing Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085, India 123 J Radioanal Nucl Chem (2011) 288:137–142 DOI 10.1007/s10967-010-0890-0