Removal of Carbon-14 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, PhD

Idaho State University:  1776 Science Center Dr., Idaho Falls, ID, 83401

smittar4@isu.edu, mldg@isu.edu

 

INTRODUCTION

Public concerns regarding availability of energy and environmental health are driving the growth of nuclear energy.  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 [5].  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 [6].  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 [1].  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 [2]. Fachinger et al. demonstrated the concept of thermal treatment of irradiated nuclear graphite in the presence of steam or oxygen [3].

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.  While oxygen is naturally adsorbed onto the graphite, for some experiments argon is mixed with small quantities of oxygen or carbon dioxide gas to increase the oxidation rate of surface species.  The oxygen combines with the surface to form carbon-oxygen bonds that are expected mostly in the form of carbon monoxide (CO) [4].

Graphite oxidation kinetics are highly temperature dependent.  Below 700°C the oxidation is limited by the rate at which graphite chemically reacts with oxygen [4].  By approximately 800°C oxidation is limited by the diffusion of oxygen into, and product gases out of, the graphite pore structure.  Above ~900°C, the rate limiting phenomenon is oxygen diffusion through the surface-boundary layer of gases [4]. 

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 has resulted in a complete experiment design, assembly/testing of apparatus and initiation of unirradiated graphite treatment.  Figure 1 shows the experimental apparatus design.

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

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 will be taken periodically during the experiment.  Solution samples from unirradiated graphite treatment will be analyzed for total carbon via titration and those from irradiated graphite will be analyzed for carbon-14 via liquid scintillation counting.

NOMENCLATURE

 

CO = Carbon Monoxide

CO₂ = Carbon Dioxide

NaOH = Sodium Hydroxide

CuO = Cupper Oxide

GIF = Generation IV International Forum

USDoE = United States Department of Energy

HTGR = High Temperature Gas Reactor

REFERENCES

 

1.     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)

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

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

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

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

6.    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