Removal of 14C From Irradiated Graphite for Waste Volume Reduction and Bulk Graphite Recycle: Thermal Treatment

Removal of ¹⁴C from Irradiated Graphite for Waste Volume Reduction and Bulk Graphite Recycle: 

Thermal Treatment

Tara E. Smith and Mary Lou Dunzik-Gougar

Idaho State University:  1776 Science Center Dr., Idaho Falls, ID, 83402

smittar4@isu.edu  mldg@isu.edu

INTRODUCTION

Nuclear power contributes 20% of the electricity used around the world today.  To improve the efficiency of future nuclear power production, the U.S. and other countries are developing advanced nuclear systems coined “Generation IV concepts” as established by the Generation IV International Forum, which institutes protocol for individual countries to lead the development of a concept in which they have particular interest [1].  The U.S. Department of Energy commissioned the Idaho National Laboratory to develop the High Temperature Gas-Cooled Reactor (HTGR).  The HTGR uses graphite coated fuel particles, graphite reflector and core structural components, and helium coolant [2].  As graphite is bombarded with neutrons, carbon-14 (¹⁴C) is produced through three distinct reactions with ¹³C, ¹⁴N, and ¹⁷O.  In irradiated reactor graphite, the majority of ¹⁴C has been found on the surface [3].  This occurrence indicates the most significant of the production reactions is neutron activation of ¹⁴N, because ¹⁴N is present on the surface and in surface pores as N₂ gas deposited from air.  Carbon-14 is relatively long-lived (half life = 5730 years) and has significant mobility in groundwater and atmospheric systems.  For these reasons, disposal of large irradiated graphite components from an HTGR would likely be costly.  Further, the value of such pure nuclear–grade graphite is expected to increase and recycle may be an option.  Thus, a means of removing the majority surface ¹⁴C from the bulk graphite ¹²C is being investigated.  Pyrolysis and oxidation in a steam atmosphere have been suggested as ¹⁴C decontamination methods [4]. Fachinger et al. demonstrated the concept of thermal treatment of irradiated nuclear graphite in the presence of steam or oxygen [5].

DESCRIPTION OF experimental WORK

Work discussed here is part of a larger project with the objective to determine the chemical nature of the ¹⁴C in irradiated graphite and to use this information to determine an optimal method for ¹⁴C removal.  This paper summarizes thermal treatment development work. 

Graphite samples are heated to a temperature in the range of 800-1500°C in the presence of inert argon gas, which carries any gaseous products released during treatment.  In the presence of argon alone, graphite surface species will react with naturally adsorbed oxygen.  To increase the oxidation rate of irradiated graphite surface species, small quantities of oxygen or carbon dioxide gas are added to the argon carrier gas.  Oxygen combines with the surface to form carbon-oxygen bonds that are expected to result mostly in the formation of carbon monoxide (CO) [6].

Experimental parameters including treatment temperature, gas composition and gas flow rate will be optimized and are summarized in Table 1.

Table I. Key Thermal Treatment Experimental Parameters

Results

Work to date includes thermal treatment of un-irradiated POCOFoam® and nuclear grade NBG-18 graphite in the presence of argon doped with oxygen. During thermal treatment, the graphite is heated in a tube furnace, through which high purity argon and oxygen flows.  After the gas leaves the thermal treatment furnace, it passes through a gas analyzer for chemical species identification.  Before final collection, carbon species in the gas are fully oxidized to CO₂ via reaction with copper oxide at 800°C.   CO₂ gas readily dissolves into sodium hydroxide (NaOH) solution, from which samples are taken periodically during the experiment. Solution samples from irradiated graphite treatment will be analyzed for carbon-14 via liquid scintillation counting. Figure 1 shows the experimental apparatus design.

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Fig.1. Experimental apparatus for thermal treatment of nuclear graphite to remove ¹⁴C

Experiments were performed at 700°C and 1400°C for two levels of dopant, 1.5 sccm and 2.6 sccm of O₂, to evaluate the effects of temperature and quantity of available oxidizing agent (and associated oxidation mechanism) on the oxidation of un-irradiated graphite.  Thermal treatment of irradiated graphite is in progress. The normalized selective removal rate of ¹⁴C will be evaluated at the same conditions described above.  The most recent results from the experiments will be presented.

NOMENCLATURE

HTGR = High Temperature Gas Reactor

ATR = Advanced Test Reactor at Idaho National Lab

CO = Carbon Monoxide

CO₂ = Carbon Dioxide

NaOH = Sodium Hydroxide

MURR = University of Missouri Research Reactor

NBG = Designation for nuclear grade graphites produced by SGL Group

REFERENCES

1.     NUCLEAR ENERGY AGENCY. “Generation IV International Forum.” Technical Secretariat. 2009

2.    Gen-IV International Forum.  “The Very-High-Temperature Reactor (VHTR) is a graphite-moderated, helium-cooled reactor with a thermal neutron spectrum.” http://www.gen-4.org/Technology/systems/vhtr.htm

3.     ELECTRIC POWER RESEARCH INSTITUTE, “Graphite Decommissioning: Options for Graphite Treatment, Recycling, or Disposal, including a Discussion of Safety-Related Issues,” EPRI Technical Report 1013091(March 2006)

4.     J. FACHINGER, L.V.WERNER, “Decontamination of Nuclear Graphite,“ 3^rd International Topical Meeting on High Temperature Reactor Technology. (October 2006)

5.     FACHINGER, J., PODRUHZINA, T., VON LENSA, W., “Decontamination of Nuclear Graphite by Thermal Treatment,” Solutions for Graphite Waste, Manchester, UK (March 2007)

6.     U.S. Department of Energy Environmental Management Spent Fuel Management Office. “Graphite Oxidation Thermodynamics/Reactions,” DOE/SNF/REP-018, DOE (1998)