(738a) 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.

 SHAPE  \* MERGEFORMAT

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)