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Title: Dissolution of used nuclear fuel using recycled nitric acid

Abstract

An evaluation was performed on the feasibility of using HB-Line anion exchange column waste streams from Alternate Feedstock 2 (AFS-2) processing for the dissolver solution for used nuclear fuel (UNF) processing. The targeted UNF for dissolution using recycled solution are fuels similar to the University of Missouri Research Reactor (MURR) fuel. Furthermore, the objectives of this experimental program were to validate the feasibility of using impure dissolver solutions with the MURR dissolution flowsheet to verify they would not significantly affect dissolution of the UNF in a detrimental manner.

Authors:
 [1];  [1];  [1]
  1. Savannah River National Lab., Aiken, SC (United States)
Publication Date:
Research Org.:
Savannah River Site (SRS), Aiken, SC (United States)
Sponsoring Org.:
USDOE Office of Environmental Management (EM)
OSTI Identifier:
1368683
Report Number(s):
SRNL-STI-2014-00448-M
Journal ID: ISSN 0149-6395
Grant/Contract Number:
AC09-08SR22470
Resource Type:
Journal Article: Accepted Manuscript
Journal Name:
Separation Science and Technology
Additional Journal Information:
Journal Name: Separation Science and Technology; Journal ID: ISSN 0149-6395
Publisher:
Taylor & Francis
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; dissolver; H-Canyon; UNF

Citation Formats

Almond, Philip M., Daniel, Jr., William E., and Rudisill, Tracy S.. Dissolution of used nuclear fuel using recycled nitric acid. United States: N. p., 2017. Web. doi:10.1080/01496395.2017.1296869.
Almond, Philip M., Daniel, Jr., William E., & Rudisill, Tracy S.. Dissolution of used nuclear fuel using recycled nitric acid. United States. doi:10.1080/01496395.2017.1296869.
Almond, Philip M., Daniel, Jr., William E., and Rudisill, Tracy S.. Mon . "Dissolution of used nuclear fuel using recycled nitric acid". United States. doi:10.1080/01496395.2017.1296869. https://www.osti.gov/servlets/purl/1368683.
@article{osti_1368683,
title = {Dissolution of used nuclear fuel using recycled nitric acid},
author = {Almond, Philip M. and Daniel, Jr., William E. and Rudisill, Tracy S.},
abstractNote = {An evaluation was performed on the feasibility of using HB-Line anion exchange column waste streams from Alternate Feedstock 2 (AFS-2) processing for the dissolver solution for used nuclear fuel (UNF) processing. The targeted UNF for dissolution using recycled solution are fuels similar to the University of Missouri Research Reactor (MURR) fuel. Furthermore, the objectives of this experimental program were to validate the feasibility of using impure dissolver solutions with the MURR dissolution flowsheet to verify they would not significantly affect dissolution of the UNF in a detrimental manner.},
doi = {10.1080/01496395.2017.1296869},
journal = {Separation Science and Technology},
number = ,
volume = ,
place = {United States},
year = {Mon Mar 20 00:00:00 EDT 2017},
month = {Mon Mar 20 00:00:00 EDT 2017}
}

Journal Article:
Free Publicly Available Full Text
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  • The deinventory and deactivation of the Savannah River Site's (SRS's) FB-Line facility required the disposition of approximately 2000 items from the facility's vaults. Plutonium (Pu) scraps and residues which do not meet criteria for conversion to a mixed oxide fuel will be dissolved and the solution stored for subsequent disposition. Some of the items scheduled for dissolution are composite materials containing Pu and tantalum (Ta) metals. The preferred approach for handling this material is to dissolve the Pu metal, rinse the Ta metal with water to remove residual acid, and burn the Ta metal. The use of a 4 Mmore » nitric acid (HNO{sub 3}) solution containing 0.2 M potassium fluoride (KF) was initially recommended for the dissolution of approximately 500 g of Pu metal. However, prior to the use of the flowsheet in the SRS facility, a new processing plan was proposed in which the feed to the dissolver could contain up to 1250 g of Pu metal. To evaluate the use of a larger batch size and subsequent issues associated with the precipitation of plutonium-containing solids from the dissolving solution, scaled experiments were performed using Pu metal and samples of the composite material. In the initial experiment, incomplete dissolution of a Pu metal sample demonstrated that a 1250 g batch size was not feasible in the HB-Line dissolver. Approximately 45% of the Pu was solubilized in 4 h. The remaining Pu metal was converted to plutonium oxide (PuO{sub 2}). Based on this work, the dissolution of 500 g of Pu metal using a 4-6 h cycle time was recommended for the HB-Line facility. Three dissolution experiments were subsequently performed using samples of the Pu/Ta composite material to demonstrate conditions which reduced the risk of precipitating a double fluoride salt containing Pu and K from the dissolving solution. In these experiments, the KF concentration was reduced from 0.2 M to either 0.15 or 0.175 M. With the use of 4 M HNO{sub 3} and a reduction in the KF concentration to 0.175 M, the dissolution of 300 g of Pu metal is expected to be essentially complete in 6 h. The dissolution of larger batch sizes would result in the formation of PuO{sub 2} solids. Incomplete dissolution of the PuO{sub 2} formed from the metal is not a solubility limitation, but can be attributed to a combination of reduced acidity and complexation of fluoride which slows the dissolution kinetics and effectively limits the mass of Pu dissolved.« less
  • In this work, the authors report the results of a study of plutonium dioxide oxidation with cerium(IV) in 1-4 M solutions of nitric acid, containing 2 x 10{sup {minus}3}-4 x 10{sup {minus}2} M of Ce(IV) and 1 x 10{sup {minus}3}-2x10{sup {minus}2} mol of PuO{sub 2} per liter of the solvent. The initial rate of dissolution of plutonium dioxide follows the equation (-d(PuO{sub 2})/d{tau}){sub tau}{ge}o = K{sub eff}(PuO{sub 2})(Ce(IV)){sub 0} = K{sup {prime}}{sub eff}S{sub PuO{sub 2}}(Ce(IV)){sub O}. The thermodynamic activation parameters of this process are evaluated as {delta}H{sub 298} = 54.4 {plus_minus}1.6 kJ mol{sup {minus}1}, {delta}G{sup No.}{sub 298} = 94.6{plus_minus} 0.8more » kJ mol{sup {minus}1}, and {delta}S{sup No.}{sub 298} = -134{plus_minus}8 J mol{sup {minus}1} K{sup {minus}1}. A probable mechanism of the reaction of Ce(IV) with PuO{sub 2} is discussed.« less
  • A kinetic study is made of the oxidative dissolution of plutonium dioxide calcined at 600-800 and 1600{degrees}C in nitric acid in the presence of ozone. Interphase reaction PuO{sub 2} + Ce(IV) is the limiting stage of the dissolution of low-calcined PuO{sub 2} at (PuO{sub 2}){sub O} {le} 5x10{sup {minus}3} mol 1{sup {minus}1} solvent. In the case of systems containing sulfuric acid, the same is true for (PuO{sub 2}){sub O} {le} 1x10{sup {minus}4} mol 1{sup {minus}1} solvent. With (PuO{sub 2}){sub O} = 0.08-0.85 mol 1{sup {minus}1} standard solvent (4 M HNO{sub 3} + (1-3) x 10{sup {minus}2}M Ce + 1.5 xmore » 10{sup {minus}2}M H{sub 2}SO{sub 4}) and at t = 70-92{degrees}C, the limiting stage of the process is interphase oxidation of cerium (III) with ozone and/or ozone transfer through the gas-solution interface. For high-calcined PuO{sub 2}, linear dependences on plutonium dioxide weight per liter of the solvent in the system are established for both the apparent rate constant of the process and cerium(IV) concentration: k= (1.19 {plus_minus}0.09) x 10{sup {minus}2}(PuO{sub 2}){sub O}, mol{sup {minus}1} 1{sup {minus}1} s{sup {minus}1}; (Ce(IV)) = (1.4{plus_minus}0.1) x 10 {sup {minus}2}-(9.21{plus_minus}1.4) x 10{sup {minus}2}(PuO{sub 2}), M.« less