(249b) 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