Fission product partitioning in aerosol release from simulated spent nuclear fuel

The long-lived fission product79Se is tightly bound to the UO2lattice in spent nuclear fuel; it will thus be released only very slowly from a geological repository for radioactive waste.

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79Se: Geochemical and Crystallo-Chemical Retardation Mechanisms

MRS Proceedings ◽

10.1557/proc-556-1115 ◽

1999 ◽

Vol 556 ◽

Cited By ~ 1

Author(s):

Fanrong Chen ◽

Peter C. Burns ◽

Rodney C. Ewing

Keyword(s):

Fission Product ◽

Nuclear Fuel ◽

Spent Nuclear Fuel ◽

Structural Connectivity ◽

Structural Similarity ◽

High Solubility ◽

Alkaline Conditions ◽

Ph Conditions ◽

Selenium Analysis ◽

Aqueous Species

Abstract79Se is a long-lived (1.1×106 years) fission product which is chemically and radiologically toxic. Under Eh-pH conditions typical of oxidative alteration of spent nuclear fuel, selenite or selenate are the dominant aqueous species of selenium. Because of the high solubility of metalselenites and metal-selenates and the low adsorption of selenite and selenate aqueous species under alkaline conditions, selenium may be highly mobile. However, 79Se released from altered fuel may be immobilized by incorporation into secondary uranyl phases as low concentration impurities, and this may significantly reduce the mobility of selenium. Analysis and comparison of the known structures of uranyl phases indicate that (SeO3) may substitute for (SiO3OH) in structures with the uranophane anion-topology (α.-uranophane, sklodowskite, boltwoodite) which are expected to be the dominant alteration phases of UO2 in Si-rich groundwater. The structural similarity of guillemninite, Ba[(UO2)3 (SeO3)2O2](H2O) 3 to phurcalite, [(UO2)3(PO4)2O2](H2O)7, suggests that the substitution (SeO3)↔ (PO4) may occur in phurcalite. The close similarity between the sheets in the structures of rutherfordine and [(UO2)(SeO3)] implies that the substitution (SeO3) ↔ (CO3) can occur in rutherfordine. However, the substitutions: (SeO3) ↔ (SiO3OH) in soddyite and (SeO3) ↔ (PO4) in phosphuranylite may disrupt their structural connectivity and are unlikely to occur.

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On the effects of fission product noble metal inclusions on the kinetics of radiation induced dissolution of spent nuclear fuel

Journal of Nuclear Materials ◽

10.1016/j.jnucmat.2008.04.018 ◽

2008 ◽

Vol 378 (1) ◽

pp. 55-59 ◽

Cited By ~ 46

Author(s):

Martin Trummer ◽

Sara Nilsson ◽

Mats Jonsson

Keyword(s):

Fission Product ◽

Nuclear Fuel ◽

Noble Metal ◽

Spent Nuclear Fuel ◽

Radiation Induced ◽

Kinetics Of

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Advances in Development of the Fission Product Extraction Process for the Separation of Cesium and Strontium From Spent Nuclear Fuel

11th International Conference on Environmental Remediation and Radioactive Waste Management, Parts A and B ◽

10.1115/icem2007-7077 ◽

2007 ◽

Cited By ~ 1

Author(s):

Jack Law ◽

Dean Peterman ◽

Cathy Riddle ◽

David Meikrantz ◽

Terry Todd

Keyword(s):

Fission Product ◽

Nuclear Fuel ◽

Spent Nuclear Fuel ◽

Fuel Cycle ◽

Feed Solution ◽

The United States ◽

Department Of Energy ◽

Hydrodynamic Performance ◽

United States Department ◽

Geological Repository

The Fission Product Extraction (FPEX) Process is being developed as part of the United States Department of Energy Advanced Fuel Cycle Initiative for the simultaneous separation of cesium (Cs) and strontium (Sr) from spent light water reactor (LWR) fuel. Separation of the Cs and Sr will reduce the short-term heat load in a geological repository, and when combined with the separation of americium (Am) and curium (Cm), could increase the capacity of the geological repository by a factor of approximately 100. The FPEX process is based on two highly specific extractants: 4,4',(5')-Di-(t-butyldicyclo-hexano)-18-crown-6 (DtBuCH18C6) and Calix[4]arene-bis-(tertoctylbenzo-crown-6) (BOBCalixC6). The DtBuCH18C6 extractant is selective for strontium and the BOBCalixC6 extractant is selective for cesium. Results of flowsheet testing of the FPEX process with a simulated feed solution in 3.3-cm centrifugal contactors are detailed. Removal efficiencies, distribution coefficient data, coextraction of metals, and process hydrodynamic performance are discussed along with recommendations for future flowsheet testing with actual spent nuclear fuel.

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Fission product interactions with nitrogen donor ligands used for spent nuclear fuel treatment

Polyhedron ◽

10.1016/j.poly.2012.10.030 ◽

2013 ◽

Vol 50 (1) ◽

pp. 154-163 ◽

Cited By ~ 21

Author(s):

Emma Aneheim ◽

Bohumir Grüner ◽

Christian Ekberg ◽

Mark R.StJ. Foreman ◽

Zuzana Hájková ◽

...

Keyword(s):

Fission Product ◽

Nuclear Fuel ◽

Spent Nuclear Fuel ◽

Donor Ligands ◽

Nitrogen Donor Ligands ◽

Fuel Treatment ◽

Nitrogen Donor

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Corrosion Study of SIMFUEL in Aerated Carbonate Solution Containing Calcium and Silicate

MRS Proceedings ◽

10.1557/opl.2013.77 ◽

2013 ◽

Vol 1518 ◽

pp. 139-144

Author(s):

Hundal Jung ◽

Tae Ahn ◽

Roberto Pabalan ◽

David Pickett

Keyword(s):

Fission Product ◽

Dissolution Rate ◽

Nuclear Fuel ◽

Polarization Resistance ◽

Spent Nuclear Fuel ◽

Electrochemical Impedance ◽

Major Effect ◽

Solution Chemistry ◽

High Rate ◽

Dissolution Rates

ABSTRACTThe corrosion behavior of simulated spent nuclear fuel (SIMFUEL) was investigated using electrochemical impedance spectroscopy and solution chemistry analyses. The SIMFUEL was exposed to aerated solutions of NaCl+NaHCO3 with and without calcium (Ca) and silicate. Two SIMFUEL compositions were studied, representing spent nuclear fuel (SNF) corresponding to 3 or 6 at % burnup in terms of fission product equivalents of surrogate elements. For all tested cases, the polarization resistance increased with increased immersion time, indicating possible blocking effects due to accumulation of corrosion products on the SIMFUEL surface. The potential-pH diagram suggests formation of schoepite that may cause the increase in the polarization resistance. The addition of Ca and silicate produced no measureable change in the polarization resistance measured at the corrosion potential. The dissolution rate ranged from 1 to 3 mg/m2-day, which is similar to the range of dissolution rates for SIMFUEL and SNF reported in the literature for comparable conditions. SIMFUEL burnup did not have a major effect on the dissolution rate. Analysis of the solution chemistry shows that uranium is the dominant element dissolved in the posttest solutions, and the dissolution rates calculated from uranium (U) concentrations are consistent with the dissolution rates obtained from impedance measurements. Simulated-fission product elements (i.e., barium, molybdenum, strontium, and zirconium) dissolved from the SIMFUEL electrode at a relatively high rate. Sorption test results indicated significant sorption of U onto the oxide formed on stainless steel. Electrochemical methods were found to be effective for measuring the uranium dissolution rate in real time.

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Determination of actinide and fission-product isotopes in very-high-burnup spent nuclear fuel

Journal of Radioanalytical and Nuclear Chemistry ◽

10.1007/s10967-008-0709-4 ◽

2008 ◽

Vol 277 (1) ◽

pp. 59-64 ◽

Cited By ~ 6

Author(s):

V. S. Sullivan ◽

D. L. Bowers ◽

M. A. Clark ◽

D. G. Graczyk ◽

Y. Tsai ◽

...

Keyword(s):

Fission Product ◽

Nuclear Fuel ◽

Spent Nuclear Fuel ◽

High Burnup ◽

Very High

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Treatment of Spent Nuclear Fuel for Separation, Immobilization and Disposal of Heat Generating High Level Fission Products, Caesium and Strontium

Journal of Chemical Engineering ◽

10.3329/jce.v30i1.34796 ◽

2017 ◽

Vol 30 (1) ◽

pp. 37-42

Author(s):

Md Akhlak Bin Aziz ◽

Afrin Ahsan ◽

Md Monsurul Islam Khan ◽

Zahid Hasan Mahmood

Keyword(s):

Fission Product ◽

Nuclear Fuel ◽

Nuclear Waste ◽

Chemical Engineering ◽

Spent Nuclear Fuel ◽

Fission Products ◽

Extraction Process ◽

Safe Disposal ◽

Synthetic Rock ◽

High Level

Separation of heat generating, high level fission product caesium and strontium from spent nuclear fuel boosts the capacity of waste repositories by as much as fifty times. For efficient use of already scarce repositories, separation of such fission products is mandatory. Separations of caesium and strontium using Chlorinated Cobalt Dicarbollide (CCD), PEG (Polyethylene Glycol), UNEX process and by Calixarenes or Fission Product Extraction Process (FPEX) were discussed. The UNEX method was then proposed as the most feasible method option. Following separation, nuclear waste immobilization techniques for such high-level fission product were discussed. The techniques included usage of concrete, glass and synthetic rock. Among them synthetic rock was identified as the most suitable one for immobilization of high-level nuclear waste. Finally, a safe disposal system with necessary required geology was shown for safe disposal of the waste.Journal of Chemical Engineering, Vol. 30, No. 1, 2017: 37-42

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