By
the requirements for the degree of
Doctor of Philosophy
at the
Defense Committee Members:
Dr. Raluca O. Scarlat, Assistant Professor, Department of Engineering Physics
Dr. Todd Allen, Professor, Department of Engineering Physics
Dr. Kumar Sridharan, Distinguished Research Professor, Department of Engineering Physics
Dr. Oliver Schmitz, Associate Professor, Department of Engineering Physics
Dr. Izabela Szlufarska, Harvey D. Spangler Professor of Engineering, Department of Materials
Science and Engineering
Dr. Cristian I. Contescu, Senior Staff Scientist, Materials Science and Technology Division, Oak
Ridge National Laboratory
i
Abstract
The objective of this research is to experimentally investigate the capability of in-core graphite
components as tritium sink for Fluoride-salt cooled high temperature reactor (FHR) design. FHR uses
eutectic mixture of LiF and BeF2 (33% BeF2+67% LiF) as coolant salt and spherical pebble fuel
elements encapsulating coated fuel particles. Although FHR has many design advantages, it faces many
design challenges, one of which is tritium production and tritium handling. Tritium is produced from
neutron irradiation mainly with 6Li in fluoride salt coolant (Flibe). Tritium has high permeability
through metallic or alloying materials under high the operation temperature and this leads to possible
tritium release to atmosphere. In addition, tritium fluoride (TF) in Flibe is highly corrosive to structural
materials, leading to safety concern in FHR design. Since FHR has large amount of graphite
components in the core (reflectors and fuel pebbles) and graphite is porous material with defects from
manufacturing process and raw materials, which creates surface area and reactive carbon sites that can
physically and chemically adsorb tritium, so this research proposed and evaluated the effectiveness of
using in-core graphite materials (nuclear graphite IG110 and graphite matrix A3) as tritium sink. Their
ability to chemisorb tritium and their chemical stability is dependent upon the density of reactive
carbon sites, which varies among grades of graphite and changes with neutron irradiation.
Microstructural characterization is performed with both IG110 and A3 with standard
characterization technique: Scanned Electron Microscopy (SEM), X-ray diffraction (XRD), Raman
spectroscopy and H2 chemisorption at 700 °C, with the focus of understanding whether empirical
relationships for nuclear graphite could extend their applicability to graphite matrix. Graphite matrix
is different from nuclear graphite in its raw materials and manufacturing method. The two materials
are similar in that they are both high-purity, fine-grain, high-graphitization materials designed for use
in high-temperature nuclear reactors. The parameters observed to be different between A3 and IG110
with a confidence level of greater than 99% were: higher H2 chemisorption, higher anisotropy, higher
rhombohedral phase content, higher degree of graphitization, higher in-plane crystallite size (Lc)
assuming one crystallite group, and higher Raman FWHM(D) for A3. The rest of the interpreted
parameters from XRD were a, La, Lc, and (002) and (004) peak asymmetry; and from Raman La, peak
intensity for D’, G’_2D, G’_3DA, G’_3DB, and peak FWHM for D, G, D’, D+D’.
Porosity analysis show that IG110 is less porous than A3, which corresponds to that the
density of A3 is smaller than that of IG110. Pore size estimation from DFT analysis with N2
adsorption isotherm and mercury porosimetry indicates that the averaged pore size in IG110 is larger
ii
than that in A3. Surface porosity and its depth profile were also characterized with image processing
tool with IG110, A3 and CGB graphite. Depth profiling showed that there might be surface porosity
decrease at graphite near surface, however, no evidence showed that surface porosity changes in the
scale of mm.
The interaction between Flibe and graphite under high temperature is studied. Evidence of
graphite fluorination (C-F formation) and structural change due to salt exposure is observed from
surface characterization (XRD, Raman) and chemical analysis (XPS and GDMS). We propose that
the fluorination of the carbon atoms at edge sites and defect sites can drive the reduction of metal
fluorides with the formation of covalent C-F bonds at those carbon sites that are more reactive than
most of the carbon elements in graphite. Structural change upon salt exposure is confirmed by changes
in lattice parameters and crystallite sizes calculated from graphite XRD and Raman spectra. For
example, XRD r(101) reflection analysis shows a rhombohedral phase content increase from 6.6% to
9.5% upon salt exposure. These are indications that the fluorination process observed here may not
be simply one of occupying existing active sites, but a more complex process by which additional
defects and active sites are introduced in graphite.
To provide information on rate of reaction and rate of species transport within the graphite
and on the equilibrium condition for graphite exposed to Flibe, experiments of varying duration time
are performed. Graphite samples after exposure with liquid Flibe for 12 hours, 24 hours and 2000
hours were analyzed. For samples that were not cleaned before post-experiment characterization
(samples from 12-hour and 240-hour exposure experiment), particles that contains fluorine were
observed and they were distributed on graphite surface as well as in the pores. This fluorine-containing
particle were in various shapes: spheres, cubes, prims. Surface roughness changed significantly upon
liquid Flibe exposure. Elemental analysis with XPS further detected fluorine signals from both cleaned
and uncleaned graphite samples. The analysis also showed that fluorine content increased with
increasing exposure time. Other factors that were in the analyzing matrix are graphite surface
treatment (polished vs. machined), Flibe chemistry (Be-reduced and unreduced), exposure condition
(gas phase exposure vs. liquid phase exposure). The XPS C 1s analysis showed the detected fluorine
signal is dependent on the surface treatment between IG110 and A3. For example, more fluorine
signal was detected on the machined surface of A3 than that of IG110, while less fluorine signal is
detected on the polished surface of A3 than that of IG110. Similar observation was found with sp2/sp3
ratio. The comparative study of graphite exposure to gas phase and liquid phase showed that liquid
Flibe is less reactive than the Flibe gas phase. The result indicates that the fluorine in gas phase is more
iii
reactive than the fluorine in the liquid phase and it could be due to the different in bond types as well
as bond strength in corresponding bond type. Graphite fluorination in reduced Flibe also behaves
quite differently than in unreduced Flibe.
Overall, the use of in-core graphite components as tritium sink as one of method to handle
tritium challenge for FHR design was investigated. The research discusses the relationship between
graphite microstructure and tritium adsorption, implying that tritium adsorption can be increased
through engineering graphite with raw materials and manufacturing process. Graphite chemical
modification (C-F formation) and microstructural change when in contact with Flibe under high
temperature (700 °C) was demonstrated. These results are relevant to FHR safety and design, as
graphite components represent a large volume fraction in the core of FHR.
iv
Acknowledgement
There are many people that have earned my gratitude for their contribution to my time in
graduate school. More specifically, I would like to thank five groups of people, without whom this
thesis would not have been possible: my thesis committee members, my lab mates, my industrial
collaborators, funding agencies, and my family.
First, I am indebted to my thesis advisor, Raluca O. Scarlat. Since my first day in graduate
school, Raluca believed in me like nobody else and gave me endless support. It all started in Fall
2014 when she offered me such a great opportunity to join the heat and mass transport group. On
the academic level, Raluca taught me fundamentals of conducting scientific research. Under her
supervision, I learned how to define a research problem, find a solution to it, and finally publish the
results. On a personal level, Raluca inspired me by his hardworking and passionate attitude. To
summarize, I would give Raluca most of the credit for becoming the kind of scientist I am today.
Besides my advisor, I would like to thank the rest of my thesis committee: Prof. Todd Allen
Prof. Kumar Sridharan, Prof. Oliver Schmitz, Prof. Izabela Szlufarska and Dr. Cristian I. Contescu,
for their insightful comments and encouragement, but also for the hard question which incented me
to widen my research from various perspectives.
I would like to thank my colleagues and all the undergraduate students I have mentored for
their continued support. This dissertation would not have been possible without the intellectual
contribution of Francesco Carotti, who has helped me tremendous with lab work and provided
insights on my research. I am thankful to Allen Chen, Ruifeng Xie, Anthony Wolf, Daniel Cech,
Colin Swee for their great contribution in various part of this project. Moreover, I am grateful to
Michael Young, Louis Chapdelaine, Quentin Desolt, Adrien Hogrel, who have graduated or left the
group for their collaboration and contribution in various projects related to this dissertation. I would
also like to thank my other colleagues that include Alexandra Delmore, Will Derdeyn, Ricardo
Vidrio, Kazi Ahmed, Mohamed Abou Dbai for making my experience in the data management lab
and graduate school exciting and fun.
I would like to thank Dr. Will Windes from Idaho National Laboratory, the researchers at
the MIT Nuclear Reactor Laboratory and at Oak Ridge National Laboratory, who have provided us
with graphite materials and advice on graphite characterization. I’m also grateful to Dr. Tams Gal
who have helped me to get access to the instrument in his lab and gave me great advice analyzing
the data.
v
This work was made possible by U.S. Department of Energy Nuclear Energy University
Program project number 15-8352 and Nuclear Regulatory Commission project number NRC-HQ-
84-15-G-0046. The authors gratefully acknowledge use of facilities and instrumentation supported
by NSF through the University of Wisconsin Materials Research Science and Engineering Center
(DMR-1121288).
Last but not least, I would like to express my deepest gratitude to my family and friends. This
dissertation would not have been possible without their warm love, continued patience, and endless
support.
vi
1. Introduction ................................................................................................................................................................ 4
2. Research Background ................................................................................................................................................ 9
3. Literature review of tritium transport behavior in liquid salt and graphite systems ..................................... 24
4. Literature review of trace hydrogen isotope measurement techniques ........................................................... 47
5. Research objectives .................................................................................................................................................. 51
8. Study of hydrogen in graphite ............................................................................................................................. 105
Chapter three: Interaction between graphite and Flibe under high temperature ................................................ 129
9. Graphite fluorination under high temperature ................................................................................................. 133
Chapter Four: Conclusions ........................................................................................................................................... 178
Table of Content
1. Introduction ................................................................................................................................................................ 4
1.2 Overview of Molten Salt Reactors (MSRs) design ...................................................................................... 6
2. Research Background ................................................................................................................................................ 9
2.2 Tritium release regulations .............................................................................................................................. 9
2.2.1. Tritium production in FHR ................................................................................................................... 9
2.2.2. Tritium flow path in FHR system ....................................................................................................... 10
2.2.3. Tritium srouces in different reactor systems ..................................................................................... 11
2.2.3.1.1.1.1. Pressurized Water Reactor (PWR) ...................................................................................... 12
2.2.3.1.1.1.2. Boling Water Reactor (BWR) ............................................................................................... 13
2.2.3.1.1.1.3. Heavy Water Reactor (HWR)............................................................................................... 13
2.2.3.1.1.1.5. Liquid Metal Fast Breeder Reactors (LMFBRs) ................................................................ 13
2.2.3.1.1.1.6. Fluoride-salt High Temperature Reactor (FHR) ............................................................... 14
2.2.4. Tritium handling experiences .............................................................................................................. 14
2.2.5. Uniqueness of tritium handling in FHR ............................................................................................ 16
2.3 Graphite components in FHR ..................................................................................................................... 17
2.3.1. Nuclear graphite IG110 and graphite matrix A3.............................................................................. 17
2.3.3. Graphite microstructure ....................................................................................................................... 21
2.3.3.1.1.1.4. Lattice parameters .................................................................................................................. 22
2.3.3.1.1.1.5. Crystallite sizes ........................................................................................................................ 23
3. Literature review of tritium transport behavior in liquid salt and graphite systems ..................................... 24
3.1 Literature review on tritium transport in salt ............................................................................................. 24
3.2 Literature review on tritium transport in graphite..................................................................................... 29
3.2.1. Study of tritium adsorption in graphite by R.A.Causey et. al ......................................................... 29
3.2.2. Study of tritium adsorption in graphite by H.Atsumi et. al ............................................................ 32
2
3.2.3. Study of tritium adsorption in graphite by E. Hoinkis .................................................................... 39
3.3 Hydrogen isotropic effect ............................................................................................................................. 40
3.4 Literature review on liquid salt and graphite interaction .......................................................................... 41
4. Literature review of trace hydrogen isotope measurement techniques ........................................................... 47
4.1 Nuclear Reaction Analysis (NRA) for Depth Profiling of Tritium ........................................................ 47
4.2 Elastic Recoil Detection (ERD) of 4He+ Ions for Depth Profiling of Tritium Isotope .................... 47
4.3 Cold Neutron Prompt Gamma Activation Analysis (CNPGAA) for Hydrogen Imaging ................. 48
4.4 Tritium Imaging for two-Dimensional Tritium Detection ...................................................................... 48
4.5 Imaging Plate Technique for Tritium Detection ....................................................................................... 48
4.6 Other Possible Techniques for Tritium Detection ................................................................................... 49
5. Research objectives .................................................................................................................................................. 51
Figure 1 Tritium flow path in FHR system [13] ......................................................................................................... 11
Figure 2 Upgrader 2 to remove tritium from D2O designed for Fugen heavy water reactor in Japan _ isotope
exchange unit [22] ............................................................................................................................................................. 15
Figure 3 Tritium removal system WTRF designed by Korea Hydro & Nuclear Power (KHNP) [23] .............. 16
Figure 4 Manufacturing process for fuel compact [27] ............................................................................................ 19
Figure 5 Conventional manufacturing steps for nuclear grade graphite [29] ......................................................... 20
Figure 6 ABA stacking graphite cell [29]...................................................................................................................... 23
Figure 7 Experiment setup for tritium recovery [32] ............................................................................................... 25
Figure 8 Experiment apparatus (left) and Flibe sample container (right) used in the in-core irradiation test [34]
.............................................................................................................................................................................................. 26
Figure 9 Five processes pertaining tritium release from Flibe with presence of H2 [35] ..................................... 26
Figure 10 Schematic diagram of the cross section of the dual-probe system [37] ................................................ 27
Figure 11 Diagram for tritium permeation cell [38] ................................................................................................... 28
Figure 12 Tritium diffusion coefficient of graphite estimated by different scientists [39] .................................. 31
Figure 13 Tritium retention at 1273 K in different graphite and carbon-carbon composites [39] .................... 32
Figure 14 model of hydrogen diffusion in graphite, proposed by Atsumi [48]. .................................................... 33
Figure 15 Relationship between hydrogen solubility and degree of graphitization with and without neutron
irradiation [50] ................................................................................................................................................................... 33
Figure 16 Relationship between hydrogen solubility and lattice constant [54] ...................................................... 34
Figure 17 Relationship between hydrogen solubility and lattice constant with and without neutron irradiation
[51] ....................................................................................................................................................................................... 34
3
Figure 18 Relationship between hydrogen solubility and edge surface area of crystallites [54] ........................... 35
Figure 19 Relationship between hydrogen solubility and crystallite size with and without neutron irradiation
[51] ....................................................................................................................................................................................... 35
Figure 20 Schematic of trap sites model proposed by Kanashenko and modified by Atsumi et.al [55] [56] .... 36
Figure 21 Pressure dependence of bulk hydrogen retention corresponding to Traps 1 and 2 in ISO-880U
graphite exposed to hydrogen gas at 1273 K [57] ....................................................................................................... 36
Figure 22 effect of irradiation on trap site 1 and trap site 2 absorption [58] .......................................................... 37
Figure 23 Pressure change due to hydrogen absorption in graphite pre-annealed at 1273k for 2 hours [58] ... 38
Figure 24 Pressure change due to hydrogen absorption in graphite pre-annealed at 1873k for 2 hours [58] ... 38
Figure 25 Peak identification of TDS spectra of ISO-880U [59]............................................................................. 39
Figure 26 The adsorption sites in graphite [61] .......................................................................................................... 40
Figure 27 Temperature dependence of hydrogen/deuterium absorption in IG-110U at 10kPa [52] ............... 41
Figure 28 Radiography result of CGB graphite rod cross section (from left to right: shorter time exposure to
long time exposure) [62] .................................................................................................................................................. 42
Figure 29 Lithium (left) and Fluorine (right) concentrations in CGB graphite sample, exposed to molten fuel
salt in the MSRE for 102 days (ORNL-4344 Section 11.3, page 148 and page 150) [63] ..................................... 43
Figure 30 Mass concentration ratio, F /Li, vs depth for CGB graphite sample exposed to molten fuel salt in
the MSRE for 188 days. (ORNL-4344 Section 11.3, page 149) [63] ........................................................................ 43
Figure 31 Comparison of fluorine concentrations in CGB graphite samples upon exposure to fuel salt in MSRE
for 102 days (Y-7) and for 188 days (X-13). (ORNL-4344 Section 11.3, page 149) [63] ...................................... 44
List of Tables
Table 1 Summary of features of Gen IV reactor systems ........................................................................................... 6
Table 3 Relative tritium production from different types of reactors [184] ........................................................... 10
Table 4 Summary of Tritium Sources in Reactors [14][15][16] ................................................................................ 12
Table 5. Characteristics of the type of graphite investigated in this study [25] [26] [27] ....................................... 18
Table 6 Summary of diffusivity and solubility of D2 in Flibe [37] ........................................................................... 27
Table 7 Summary of Tritium Diffusivity and Solubility in Flibe, calculated from Equation 8 and Equation
9 ........................................................................................................................................................................................... 28
Table 8 Summary of empirical equation for tritium diffusivity in graphite ............................................................ 30
Table 9 Summary of empirical equation for tritium solubility in graphite ............................................................. 31
Table 10 Summary of Tritium Detection Techniques ............................................................................................... 50
4
1. Introduction
Our world is now facing a dilemma of balancing energy scarcity, economic growth and climate
change. The development of renewable energy technology is regarded as a promising solution to this
dilemma, as it can help to mitigate climate change while at the same time maintain required economic
growth. Many countries have realized the urgency and necessity of developing and implementing
renewable energy technology. For example, the shares of U.S. energy consumption from biofuels,
solar, and wind energy have increased significantly since 1990. In addition, development of nuclear
energy is definitely another way out of this dilemma due to huge amount of energy can be produced
per unit fuel mass compared to other energy forms. Nuclear energy can be taken as a type of clean
energy to some extent except for the fact that development of fuel waste management is yet not
mature. Advocating nuclear energy faces many challenges because the public always consider nuclear
energy as extremely dangerous due to the well-known Three-mile island accident (1979), Chernobyl
accident (1986), and Fukushima accident (2011), regardless of the fact that the possibility of such
severe accidents is extremely low. So, one major endeavor that all nuclear engineers are devoted to is
to design safe and reliable nuclear systems as well as diversify advanced reactor designs, in order to
inherently increase the safety and security of advanced nuclear system as well as the adaptability to
local terrain and environment. The Research and Development (R&D) of Gen IV nuclear reactor
systems is aimed to fulfill this purpose. This thesis is interested in the development of solid fuel type
Molten Salt Reactors (MSRs) design (known as Fluoride-salt cooled high temperature reactor (FHR)),
which is a very promising Gen IV nuclear reactor system. This thesis discusses one of the big
challenges that is encountered in FHR design: tritium production and tritium handling, and then
proposes and investigates the possibility of using in-core graphite components as tritium sink. This
research is valuable both for scientific and engineering field, which could promote the development
of safety and economic design of FHR.
1.1 Gen IV nuclear reactor systems
Six concepts of nuclear reactor systems were selected out of about one hundred concepts by
the Generation IV International Forum (GIF) in 2002 and were known as Gen IV nuclear reactor
systems, as listed in Table 1. The six nuclear reactor concepts were chosen on a basis of the meeting
the following objectives: economic competitiveness, enhanced safety and reliability, minimal
radioactive waste generation, and weapon proliferation resistance [1]. Gen IV reactor systems have
many advanced design concepts in terms of selection of neutron spectrum (fast or thermal), coolant,
5
and type of fuel cycles. Gen IV nuclear reactor systems all feature high designed coolant outlet
temperature and high electrical efficiency of 40% to 50% and can be used in non-electrical application,
e.g. hydrogen production[2].
Among the six nuclear reactor system concepts, Very High Temperature Reactors (VHTRs)
and Molten Salt Reactors (MSRs) are quite popular due to the previous experimental operation
experiences or prototype construction. VHTRs is characterized by its unique fuel pebble design,
which consists of coated fuel particles embedded in a graphite matrix with a spherical shape (known
as fuel pebbles). The fuel pebbles are located in a graphite core and cooled by flowing helium through
the reactor core. This design allows the temperature of helium coolant to be as high as 900 °C to
1000 °C, resulting in high electrical efficiency and capability of supplying process heat for non-
electrical applications. MSRs is characterized by the variability of the fuel form: solid fuel form (same
as VHTR fuel design) or liquid fuel form where fuel is dissolved in liquid salt and the fuel salt performs
as both fuel and coolants. More descriptions of MSRs design features are given in Section 1.2.
6
Table 1 Summary of features of Gen IV reactor systems
System Neutron
Molten Salt
Reactor Thermal
Electricity &
hydrogen
Sodium-
cooled
hydrogen
Super-
Critical
Water-
cooled
Reactor
Thermal/
hydrogen
Molten Salt Reactors (MSRs) and Fluoride-salt cooled High Temperature Reactors (FHRs)
designs are two types of advanced nuclear reactor systems. Thermal MSRs and FHRs both employ
liquid salts as coolants and graphite as moderator. MSRs design has several inherent safety design
features. First, the reactor operates at high temperature and low pressure because the coolant/fuel
salts have high melting temperatures and high boiling temperatures and no two-phase instability or
coolant boiling would occur even in accident scenarios. Second, MSRs can remove excess heat from
reactor core by natural circulation when coolant/fuel salts circulate through reactor core under gravity
7
and density change in accident conditions. Third, in most MSR designs, the density of salts decrease
with increasing temperature, which leads to negative reactivity feedback and in turns mitigate reactivity
increase in accident conditions [3]. Because of these design advantages, MSRs design is favored by
many countries (USA, Japan, China, and Russian) and by several companies in various countries
(Terrestrial Energy, ThonCon Power, Seaborg Technologies, etc.) and has been developed rapidly in
recently few years. The design concept proposed by each company or country is different in terms of
fuel type, coolant type, neutron spectrum, designed power, application and so on [4].
MSRs are salt-fueled reactors, in which the fuel is dissolved within the salt. FHRs are salt-
cooled reactors, in which the core contains a solid fuel and liquid salt coolant. There are many
advantages in the design of salt-fueld MSRs: no need for fuel fabrication, thus reducing fuel cycle cost;
dissolved fuel materials allows homogenous composition, thus resulting in no formation of hot spot
and great flexibility in the fuel cycle; intrinsic safety due to negative temperature and thermal fuel
coolant reactivity and the possibility of draining the fuel coolant from reactor core under accident
conditions; the option of online refueling and online fuel reprocessing. The concept of salt cooled
MSRs is favored by less companies. Among those companies, both Kairos energy in USA and
Shanghai Institute of Applied Physics (SINAP) in China are developing pebble bed fuel type MSRs.
The fuel pebble design is similar to that in VHTR design with minor changes in dimensions and pebble
compositions. Advantages of fuel pebble design include: inherent safety with low possibility of fuel
damage or core melting down due to the quite high melting point of graphite materials; increased fuel
burnup due to the option of online fuel pebble processing or online fuel pebble refueling.
Even though liquid salts have thermal and neutronic stability at high temperature and they do
not react with air or water, the use of liquid salts as coolant or fuel coolant brings design challenges in
the meantime. First, the chemistry and thermal-dynamic behavior of liquid salts under neutron
irradiation is still unclear. Second, fission products like lanthanides, noble gases or noble metals will
dissolve in liquid salts and will change their thermal and neutronic performances, thus they need to be
removed continuously. Third, liquid salts are corrosive to both structural materials, especially under
high temperature and neutron irradiation, due to the fact that many alloying elements (e.g. Chromium)
dissolve in liquid salts and form fluorides or chlorides [5]. The impurities from both liquid salts and
structural materials as well as the oxidative fission products make liquid salts even more corrosive. In
addition, for salt-cooled MSRs using solid fuel, the compatibility of liquid salt with graphite matrix is
yet to be investigated.
8
In this thesis, the focus is on FHR technology. This technology is being commercially
developed by Kairos Power. FHR uses eutectic mixture of LiF and BeF2 (33% BeF2+67% LiF) as
coolant salt and spherical pebble fuel elements encapsulating coated fuel particles. FHR uses nuclear
air-Brayton combined cycle (NACC) with a modified off-shelf gas turbine from General Electric and
can provide both base-load electricity when operated with only nuclear heat and peak mode when
combined with other energy sources [6]. Besides materials challenges that are also faced by other MSRs
designs, FHR also faces the challenges of tritium handling. Tritium is produced from neutron
irradiation with the fluoride salt coolant (we call it Flibe hereafter) and large amount of tritium is
produced due to the presence of 6Li. Tritium has high permeability through metallic or alloying
materials that are proposed to be used in the heat exchanger in FHR especially under high the
operation temperature and this leads to possible tritium release to atmosphere through heat exchanger
in the secondary loop. In addition, another tritium species in Flibe, tritium fluoride, is highly corrosive
to structural materials and this leads to safety concern in FHR design. So, it is necessary to look for
techniques and methods that can be minimize tritium release to air and corrosion to structural
materials. In this thesis, the possibility of using large amount of graphite components as tritium sink
is investigated. Section 2 gives full background of this study and full literature reviews. Section 2.1
introduces details of tritium production and tritium flow path in FHR and literature reviews on tritium
production and tritium handling experiences in various types of reactors. Section 2.3 introduces the
two types of the graphite materials used in FHR and graphite microstructure parameters that exhibit
close relationships with tritium absorption. Section 1 introduces previous studies on tritium transport
in liquid salt, tritium transport in graphite and the interaction between salt and graphite. Section 1
gives literature review on trace hydrogen isotope measurement techniques and proposes the
techniques can be applied in this study. Section 1 lists the research questions that this thesis attempts
to answer.
2. Research Background
Tritium is an isotope of hydrogen. Unlike hydrogen and deuterium, which are stable isotopes
with natural abundance of 99.98% and 0.02% separately. Tritium is a radioactive nuclide which only
naturally exists in a trace amount. Among all the radioactive nuclides transmuted in nuclear reactor,
tritium is one of the least hazardous one due to its relatively short half-life (12.23 years) and low beta
energy (maximum energy of 18 eV and average energy of 5.6 eV). However, tritium gas permeates
through structural materials especially under high temperature, which increases of the possibility of
tritium release to the air from the pipeline, heat exchanger, etc. Tritium dissolves in water and produce
titrated water through isotope exchange, which could contaminate the underground water system in
the neighbor of the nuclear power plant (NPP). Tritium could also be released to the air through
evaporation of titrated water or spent fuel pool. These characteristic of tritium makes tritium handling
very essential and challenging for all types of reactors.
2.1 Tritium challenges in FHR design
2.2 Tritium release regulations
NRC and EPA establish limits for tritium concentration in water effluents to groundwater, for
tritium concentration in air effluent, and maximum worker dose. The EPA drinking water standard is
0.74 Bq/mL (4 mrem/year intake, Radionuclides Rule 66 FR 76708), and the NRC effluent limit is 37
Bq/mL in effluent water, and 3.7 mBq/mL in effluent air (25 mrem/yr, 10 CFR 20 Appendix B, Table
2). Besides FHRs, tritium is also produced in current commercial reactor-PWR, BWR, CANDU, and
advanced Gen III reactor design – HTGR. In this section, tritium production from different reactor
types are discussed. Tritium source, tritium production amount and typical tritium handling technique
for each reactor type is included in the discussion.
2.2.1. Tritium production in FHR
Tritium is produced in FHR from neutron irradiation with the constituents in Flibe, Equation
1 to Equation 4 gives the main nuclear reactions that lead to tritium production as well as the reaction
cross sections [7]. The main contributor to tritium production is 6Li, whose natural abundance is in
the range of 3.75% to 7.59%. Even though 99.995% 7Li is proposed to use to produce eutectic Flibe
in order to minimize tritium production, tritium production in FHR is still high compared to
commercial PWR and BWR reactor [8].
0 1 + 3
4 ; σ=148.026 barns Equation 1
10
1 + 1 3 ; σ=1.0e-3 barns Equation 2
4 9 + 0
6 ; σ=3.63e-3 barns Equation 3
2 6 → 3
6 + − + Equation 4
A comparison of tritium production per GWe power per day from different types of reactor
designs is summarized in Table 2. The table shows that FHR produces about 5500 Ci/GWe day
tritium, which is about 100 times more than that from PWR and two times more that from CANDU
heavy water reactor. The same as CANDU reactor, tritium management systems are necessary to add
in FHR to control tritium air emission. Tritium handling experience in CANDU systems and its
suitability in FHR systems are discussed in Section 2.2.3.
Table 2 Relative tritium production from different types of reactors [184]
Reactor Power (GWe) Relative Tritium Production (Ci/GWe day)
FHR 0.410 5500
MSBR 1.000 2420
PWR 1.000 63
BWR 1.000 58
HWR(CANDU) 3.512 3800
HTGR 0.300 7.8776
LMFBR 0.6 0.39
2.2.2. Tritium flow path in FHR system
The diagram in Figure 1 shows tritium flow path in FHR system: tritium produced in liquid
Flibe in the core of FHR exists in Flibe mainly in two species: T2/TF; some of the tritium species
diffuse to the surface of the fuel elements and then diffuse to the bulk of graphite materials and being
trapped within the fuel elements; some of the dissolved tritium species flow with liquid Flibe to the
heat exchanger in the secondary loop, where T2 easily permeates through the metallic or alloying walls
in the heat exchanger resulting in tritium leakage to air and TF corrode the structural materials by
oxidizing some elements to its fluorides, e.g. CrF2. The ratio of T2 and TF can be controlled by the
chemistry of liquid Flibe, also known as redox control. There are plenty of literatures and discussions
about Flibe redox control but this is not the scope of this specific study [9][10][11][12].
11
2.2.3. Tritium srouces in different reactor systems
Tritium can be produced from the following sources, as summarized in Table 2 [14][15]: 1)
ternary fission of fuel material (U-235, U-233, U-238); 2) neutron activation of control rod material
– boron carbide; 3) neutron activation with boron or boric acid (burnable poison for reactivity
control) and lithium cation (PH control); 4) neutron activation of deuterium in water coolant; 5)
neutron activation with impurities from structural materials; 6) neutron activation of nitrogen in the
air from the containment atmosphere; 7) neutron activation with Be from the secondary neutron
source – Sb-Be. The amount of tritium produced from each source is highly dependent on the reactor
types, structural materials, etc. This section will briefly discuss tritium sources in each reactor type.
12
Table 3 Summary of Tritium Sources in Reactors [14][15][16]
Element
(barns)
Neutron
B 2 + → 2 + + 0.324 1
3 2 4
LMFBRs,MSRs 3 + → + + 2 4
3 7
1 2 0.00055 HWR(CANDU)
2 3 2280 0-2.38ev
3 408.026 0-2.38ev
4 3 7
+ + 1 3
3 7 + 1
9 19 + 1
7 + 1 3
4 6
3 ~ HTGR, MSRs
2.2.3.1. Pressurized Water Reactor (PWR)
Tritium production in PWR could be from all of the mentioned sources. The majority of the
tritium produced from the ternary fission remains inside the fuel element (especially with zircaloy as
the fuel cladding material). In the reactor cooling system (RCS), most of the tritium is produced from
B-10 from the burnable poison - boric acid. Around 80% of the tritium in RCS is from reaction 3 and
reaction 7 or reaction 8, as shown in [17].
13
2.2.3.2. Boling Water Reactor (BWR)
Tritium production in BWR is primarily from ternary fission (reaction 1), burnable poisons
(reaction 2&3), and deuterium (reaction 4) in the coolant. Tritium production from BWR is about
10% of that from PWR because there is no boric acid added to the coolant in BWR.
2.2.3.3. Heavy Water Reactor (HWR)
HWR use D2O as coolant and moderator, and Deuterium is also a contributor for tritium
under neutron activation. So, reaction 4 is the main source for tritium in HWR despite for a very low
cross section ~ 0.00316e-12 m2. Tritium produced from ternary fission as well as neutron activation
with Li and Be can be neglected in HWR (less than 0.5%)[14]. The CANDU reactor (CANada
Deuterium Uranium) is a heavy water design, which uses deuterium as coolant and moderator. 97%
of tritium is produced from neutron activation with deuterium atoms and dissolved in water in the
form of HTO. The average production rate of tritium of a CANDU power reactor is about 2.4e16
Bq/year of tritium of deuterium per reactor unit. A representative total release from a CANDU power
reactor is about 3.3 X 1014 Bq/yr through gaseous and liquid pathways which represents about 1.5%
of the total tritium produced in a year [18].
2.2.3.4. High Temperature Gas-cooled Reactors (HTGR)
Tritium is produced in HTGR from ternary fission of U235 and U233, neutron activation with
impurities from graphite moderator and matrix graphite of fuel pebble (mainly Li and B), helium
coolant (He-3), and boron control element. Ternary fissions of U-233, U-235 and U-238 are the main
contributor for tritium in HTGR, which was estimated to be about two times of the rest of the sources
[14]. Helium obtained from natural gas wells, which is currently being used to cool HTGRs, has 3He
compositions of 0.5 x l0-5 to 3 x l0-5 %. Helium attained from air has higher 3He contents (1 .3 x 10-
4%). Tritium production by activation of lithium is approximately equivalent to that from activation
of 3He [19][20]. Lithium is primarily present in HTGRs as impurities in graphite with abundance in
ppm level. Boron is used as a neutron absorber and is present as a burnable poison in fuel elements
and absorber rods. It is estimated for HTR-10 that the tritium production at the end of 20 years
operation is about 2116 Ci in the reactor core.
2.2.3.5. Liquid Metal Fast Breeder Reactors (LMFBRs)
Ternary fission and by neutron activation reactions with boron in B4C control rods are the
main two tritium sources in LMFBR. Lithium impurities in fuel also contributes small portion of total
14
tritium production. The hard neutron spectrum in LMFBR increases the total tritium production
compared to thermal reactors, because both the cross section for neutron activation reactions and
ternary fission yield increase with neutron energy. Also, Pu-239, which is the fuel material for breeder
reactors, has higher tritium yield than U-235.
2.2.3.6. Fluoride-salt High Temperature Reactor (FHR)
FHR belongs to the type of Molten Salt Reactors (MSRs). As introduced above, it uses high-
temperature liquid salt (2LiF+BeF2) as coolant, graphite as moderator and reflector, and fuel pebble
design which uses matrix graphite. Besides ternary fission with fuel material and burnable poison in
control rod, the big contribution of tritium is from neutron irradiation with liquid salt which has Li
and Be. Preliminary estimation of tritium production in Mark I design gives that a 410MWe PB-FHR
produces about 2508 Ci of tritium per day at equilibrium and potentially several times as much during
the first few years of operation.
2.2.4. Tritium handling experiences
Tritium could be removed from reactor systems in gaseous form or in liquid form. HWR,
HTGR, LMFBR and FHR are the four reactor types with larger tritium production. In this section,
the techniques used or designed for tritium handling in these reactors are summarized.
Heavy water recycle is completed in Fugen using two heavy water upgraders. The upgrader-2
(a CECE process) recovers tritium from the degraded heavy water and reduces its effluent tritium
concentration lower than 3700 Bq/cm3. Rn, Tn separation IC was developed for continuous
atmospheric tritium monitoring [21].
15
Figure 2 Upgrader 2 to remove tritium from D2O designed for Fugen heavy water reactor in Japan _
isotope exchange unit [22]
In 1999, Korea Hydro & Nuclear Power (KHNP) made decision to construct the WTRF to
reduce tritium levels at WNPG. The construction and commis- sioning for the WTRF have been
successfully completed and then KHNP declared in-service on June 2007. The WTRF has been
designed to process 100 kg/h of heavy water with the overall tritium extraction efficiency of 97% per
single pass and to meet the requirement of the target tritium level of 0.37TBq/kg in the moderator
(Miller et al.2006). The upper limit of tritium concentration in the heavy water feed is 2.22 TBq/kg.
The recovered tritium is immobilized as a metal hydride to secure its long term storage at equal to or
greater than 99% tritium. The process design has been based on the heavy water de-tritiation and
tritium handling technologies developed and proven by AECL, Kinectrics and KEPRI. The tritium
separation process is to produce a stream sufficiently enriched in tritium to be economical for storage
and a stream sufficiently depleted in tritium that can be recycled to the reactor systems. The de-
tritiation process of the WTRF is consisted of three parts; a LPCE, a CD and a tritium storage. As
shown in Figure 3, in the front-end process, tritium is transferred from tritiated water into deuterium
gas and the tritium-depleted heavy water is returned for reuse at the reactor. Tritium from the front-
end process is separated from protium and deuterium and enriched to a high specific activity in the
back-end process. Then, the tritium-depleted deuterium returns to the front-end process, and tritium
is stored as a titanium metal tritide in a separated vessel[23].
16
Figure 3 Tritium removal system WTRF designed by Korea Hydro & Nuclear Power (KHNP) [23]
Tritium management remains a big challenge for FHR design, and researchers are working on
different in-core and out-of-core methods can be applied in tritium control and tritium management
in FHR system. One common approach to avoid tritium release to the atmosphere through the heat
exchangers is to use coating to form a tritium barrier. The normally used coating material is aluminum-
oxide, and the mass diffusivity of tritium in Al2O3 is measured to be very small, about 10-19 m2/s [24].
A preliminary calculation showed that the tritium release can be reduced by a factor of 100 with the
Al2O3 coating [8]. Other tritium handling proposals include: In-core tritium sink to take advantage of
the large amount of the graphite components in the reactor core; out-of-core tritium sink to use
graphite or other high porosity materials to absorb and store tritium for future processing.
2.2.5. Uniqueness of tritium handling in FHR
The uniqueness of tritium handling in the FHR systems is as follows. First, as discussed in
Section 2.2.2, tritium that is produced in Flibe can exist in two forms: T2/TF and the ratio of the two
species is controlled by Flibe chemistry, as indicated in Equation 5 and Equation 6, where M is any
metal element. The F-/F2 ratio dissolved in liquid Flibe is in dynamic equilibrium at high temperature,
which is affected by corrosive with structural materials and cover gas composition. Equation 5 shows
T2/TF that T2/TF is highly dependent on the F-/F2 ratio.
½ T2 + ½ F2 <--> TF Equation 5
2/n M + F2 <-->2/n MFn Equation 6
17
The mass transport behaviors of T2 and TF are different in the system, and it is not easy to
study the transport phenomenon of T2 and TF separately. Second, there are three transport barriers
for tritium species being absorbed in graphite eventually: tritium diffusion and solubility in Flibe,
tritium diffusion through Flibe-graphite interface and tritium diffusion and solubility in graphite. None
of the three processes or interaction mechanisms are fully understood or extensively being studied.
Besides, the physical and chemical interaction between the liquid Flibe and graphite materials is unclear
and thus its effect on tritium interaction is unknown. Since T2 and TF behave differently in the system
and Flibe redox potential change will affect T2/TF ratio in liquid salt, it is necessary to add this
difference into the matrix.
2.3 Graphite components in FHR
There are mainly two types of graphite materials in the core of FHR: one type is nuclear grade
graphite which is commonly used as moderators and reflectors in both commercial reactors and
advanced reactor designs; the other type is graphite matrix A3, which is the matrix materials used to
encapsulate TRISO fuel particles and form the spherical shape. A3 is used specifically for making the
fuel elements for the VHTR and FHR in Gen IV nuclear reactor systems.
2.3.1. Nuclear graphite IG110 and graphite matrix A3
The various characteristics of the two grades of graphite under study are summarized in Table
4 below. The difference in raw materials and manufacturing processes lead to the differences in
graphite microstructures. The graphite microstructures that are relevant and studied in this research is
described in Section 2.3.3.
18
Table 4. Characteristics of the type of graphite investigated in this study [25] [26] [27]
Item A3 IG-110
Formation Promess Press @ 1mm/s Cold Isostatic Pressing
Heat Treatment
under 600 psi ~2100 °C
Grain Size (µm) ~100 20
IG-110 graphite batch was provided by David Carpenter from MIT and the material was
originally procured from Toyo Tonso Co. Japan [25]. IG-110 is manufactured from mixture of
petroleum coke and tar pitch binder, following by the cold isostatic pressing to reach the near-isotropic
or isotropic materials. Graphitization typically occurs at temperatures ~2100 °C[26].
Manufacturing process of A3 is slightly different from that of IG-110, because matrix graphite
encapsulates fuel particles and forms fuel pebbles. The temperature for graphitization is much lower
than IG-110, because uranium easily diffuses out of the matrix at temperature higher than 1900 °C[27].
Matrix graphite uses more graphitized raw material to compensate the adverse effect of low
graphitization temperature. The matrix precursor powder for compact lot A3-H08 was a jet milled
blend of 64 wt% natural graphite (Asbury 3482), 16 wt% synthetic graphite (Graftech GTI-D), and
20 wt% high purity novolac resin (Hexion D_SD-1708 with 7.5% hexamethylenetetramine added).
ORNL developed a novel fuel pebble manufacturing process originated from the techniques used for
peach bottom reactor and dragon project. A detailed description can be obtained from the paper
published by Pappano et al. [27].
19
2.3.2. Graphite manufacturing process
Nuclear grade graphite materials are produced by modification of conventional manufacturing
process [28] [29]. Purification is taken as a key step to remove purities that have significant neutron-
absorption cross sections. Figure 5shows the conventional nuclear-grade graphite manufacturing
process. Various purification methods could be applied to achieve graphite impurities of 99.999% or
even higher.
In general, graphite manufacturing started with carbonaceous filler material and carbonized
binder. The mostly used filler material are petroleum coke, metallurgical coke and binder material
usually is coal-tar pitch. Petroleum coke can be easily graphitized, and the carbon can achieve a high
degree of crystallinity with a temperature of 2800 °C to 3000 °C. The binder material, coal-tar pitch,
is the heavy residue from the distillation of coal tar from by-product coke ovens. Coal-tar pitch is a
thermoplastic material, allowing thorough mixing of the filler particles with binder particles, faster
formation of filler-binder mixture. Another favorable feature of coal-tar pitch is the high carbon
content (93%). With some feasible process, a large portion of the carbon forms bond with the carbon
20
from the filler particles. Coal-tar pitch also has high specific gravity, which helps spreading the carbon
atoms in and around the filler particles. For the manufacturing of graphite matrix A3, resin is selected
as the binder because Resin has low specific gravity and result in formation of low-density graphite
matrix that enables the fuel pebbles float in liquid Flibe.
Figure 5 Conventional manufacturing steps for nuclear grade graphite [29]
As shown in Figure 5, the filler material is calcined at temperatures no high than 1300 °C to
remove gaseous hydrocarbons and shrink the filler material before [28]. About 25 % of weight loss
will be observed after the calcination due to loss of hydrocarbons, moisture and even dust. The filler
material (coke) is then crashed and milled into particles with required size. Particle shape and density
are different for filler materials from different sources. The higher the crystallinity, the higher the
density for the filler material. Coal-tar pitch is also distilled through a series of process to remove
water, light oils, medium oils and heavy creosote. The filler particles are mixed with distilled binder
material with addition of additives to increase the bulk density of graphite. The mixture is extruded
and molded to form the shape of a block or cylinder. During the extrusion operation, the filler particles
are aligned with their long dimensions parallel to the direction of extrusion. This anisotropy originates
from the predominate orientation of the crystallite layer planes parallel to the long particle dimension.
With larger filler particle size, the anisotropy is reduced. The material is then baked at temperature of
21
800 °C to 1000 °C, allowing polymerization and cross-linking to occur within the binder and between
the binder and filler materials with release of large quantities of hydrogen. The material ends up being
hard and brittle after the baking process. Impregnation with binder materials is also needed to tune
the density of the graphite. Final step is graphitization at temperature around 3000C to facilitate the
crystal growth and perfect the internal order. At last, purification is applied to remove unwanted
impurities from the graphite material, boron as we mentioned above. Thermal purification, diffusion
of impurities out of graphite under high temperature, and chemical purification, reaction of impurities
with halides, are the two main techniques for graphite purification.
2.3.3. Graphite microstructure
As described in Section 2.3.2, the microstructure of graphite materials are highly dependent
on the selection and characteristic of raw materials, on the selection of powder pressing and sintering
technique and also on the purification process. Even for the graphite materials with the same raw
material recipe and manufacturing process, the microstructure of the produced graphite is still not the
same because the microstructure and impurity contents of the raw materials are different from batch
to batch. Furthermore, graphite materials has very complex microstructures: pore structures are
irregular and hardly to define; definition of graphite grain is not unclear which makes it hard to identify
grain size and grain shape accurately; the degree of isotropy can be varied from one type to another.
Despite of those complexity, graphite materials still contains lattice structure and crystallite structure
which are consisted of lattices. Traditional characterization techniques can be applied to estimate
graphite lattice parameter and crystallite sizes. In addition, density, porosity, specific surface areas
which are often used to characterize graphite materials for their chemical and physical performances
are also introduced in this section.
2.3.3.1. Density
The density of ideal graphite with ABAB stacking order and no defect, is calculated to be 2.265
g/cm3. In reality, the density of both natural graphite and artificial graphite are less than 2.265 g/cm3,
and even less than 2 g/cm3 due to the defects, pores in the microstructure. Density can reflect the
level of perfection of one type of graphite. Density measurement is not easy for graphite, because the
volume of the graphite sample is difficult to be measured with low uncertainty. Graphite volume plays
an important role in density measurement with different kinds of techniques.
22
2.3.3.2. Porosity
Total porosity of the graphite can be defined by the ratio of apparent density to theoretical
density (2.265 g/cm3). In graphite, there are two types of porosity: open porosity and closed porosity.
Open porosity is the fraction of the pores that are accessible by fluids or gases. Close porosity is the
fraction of the pores that are accessible by fluids or gases. There are two different understanding
regarding to closed porosity. The first interpretation is the pore volume that are too small for the
liquid or gas to access under the pressure as high as there are no structural damage. The other
interpretation is that the pore is closed due to the resistance to alignment of individual crystallites,
which are inaccessible regardless of the pore size. Measurements of open and closed porosity are very
challenging considering the above discussed reasons.
2.3.3.3. Specific surface area
Specific surface area of graphite can be simply defined as the surface area of the pore
structures per unit sample mass. Graphite specific area can be estimated from pore size distribution
and the assumption of pore structure. By analyzing the adsorption isotherm curves obtained from
gas adsorption method, surface area can be estimated, as discussed in Section 6.3.1.
2.3.3.4. Lattice parameters
Ideal graphite is consisted of infinite networks of carbon hexagons in a parallel order. in real
graphite structure, the graphene sheet is finite, and always with defects, resulting in unsaturated
carbon bond with free, active electrons/weak spin-pairing bonds or buckling of the graphene sheet
with introduction of foreign groups, for example, -OH, -O-, =O, etc. Figure 6 showed a perfect
graphite cell with ABA stacking order. Conventionally, the unit graphite is simplified to tetragonal
crystal, as indicated in Figure 6. Lattice parameter “a” and “c”, basal plane, prism face is also shown
in the figure. In this case, you can use miller index for hexagonal crystal or that for cubic cell. For a
perfect graphite crystal, lattice parameter “a” is 2.46A, interlayer spacing is 6.71A [29] and density
2.26 g/cm3 (from POCO Graphite).
23
2.3.3.5. Crystallite sizes
Crystallite is defined as the maximum successive unit in both basal plane and c direction.
The size of crystallite is dependent on the type of the raw material, the particle size of the raw
material, and graphitization temperature.
Several techniques can be used for characterizing graphite microstructures. None of the
techniques is perfect due to the complexity and diversity in graphite microstructure. In the following
section. The techniques that are used to characterize A3 and IG-110 in our lab are described. And
the comparison results are discussed in section detail Section 1 in Chapter Two.
24
3. Literature review of tritium transport behavior in liquid salt and graphite systems
As mentioned in Section 2.2.3, there are three tritium transport barriers to consider in the
liquid salt and graphite system in order to study whether or not the in-core graphite components can
be used as tritium sink. In this section, literature review of tritium transport in liquid salt and in graphite
is summarized separately. Tritium transport mechanisms and parameters relevant to mass transport
(diffusivity, solubility) are introduced and discussed. In addition, literature studies on the interaction
between liquid salt and carbon materials is discussed and its effect on tritium absorption in graphite is
proposed.
3.1 Literature review on tritium transport in salt
The study of tritium behavior as well as tritium speciation in salt was started from 1980s. One
motivation of the study is the observation of large tritium production from Molten Salt Reactor
Experiment which was operated from 1956-1959. Another Motivation is the idea of using liquid LiF-
BeF2 as the blanket for fusion reactors.
As mentioned in the previous chapter, tritium is produced in Flibe mainly from neutron
activation with Li-6. Tritium is produced in the form of T+ and remains in the liquid salt predominately
as TF with some amount of HT. The equation that produces HT is shown in Equation 7. The H2 is
from the trace residue from Flibe purification process [30]. Trace amount of tritium also exist in the
form of T-, which bonds to Li and forms LiT. T- is very reactive and can easily be oxidized to T+ or
T2.
+ + → + + Equation 7
In 1980s, Japanese researchers Jun. OISHI et al designed an experiment apparatus to recover
tritium from molten Flibe [31][32]. The theory of their experiment is based on the hypothesis that
tritium species in molten salt are in two forms: condensable TF and non-condensable HT/T2. Gas
chromatography and proportional counter were used to measure tritium concentration. The diagram
of the experiment is shown in Figure 7.
25
Figure 7 Experiment setup for tritium recovery [32]
A.Suzuki took a further step to study the tritium speciation in salt by studying the effect of H2
partial pressure in the sweeping gas. The study was intended to understand the production and
behavior of TF and with the intention to benefit the application of Flibe blanket in fusion reactors
and also in breeding reactors. Figure 8 shows the in-core tritium production apparatus and the design
of Flibe container [33]. A series of controlled experiments were done by varying the fraction of H2 in
the He+H2 purging gas from 0% to 100%. Ionization chambers were used to measure tritium
concentration. In the experiment, the combination of an aluminum reduction bed and a molecular
sieve bed separated the tritium species: TF and HT. A. Suzuki described the combination use of the
two beds in detail [34] [35].
26
Figure 8 Experiment apparatus (left) and Flibe sample container (right) used in the in-core
irradiation test [34]
A.Suzuki and his co-workers gradually developed and proposed five processes that could
happen in Flibe with the presence of H2, as shown in Figure 9. He concluded that the tritium species
released from Flibe is highly dependent on the amount of H2 in the system and F potential in the salt:
1) with large amount of H2 in the purge gas, HT is the main species released from salt; 2) with small
amount of H2 and low F potential, more TF are released compared to HT; 3) with small amount of
H2 but high F potential, more HT are released.
Figure 9 Five processes pertaining tritium release from Flibe with presence of H2 [35]
Diffusion coefficient measurement of hydrogen isotopes were started with the Japan-US joint
research project (JUPITER-II) [36][37]. A dual-probe system was designed, as shown in Figure 10.
The probes were made from Ni, which is very permeable to hydrogen isotopes. The hydrogen
27
concentration was measured by the quadrupole mass spectrometer (QMS). As shown in the diagram,
the experiment was designed to track and account for all the hydrogen isotopes introduced to the
system by purging Ar gas through all the head space above salt as well as barrier volumes (marked in
Figure 10). The experiment was performed with Deuterium, and the diffusion coefficient and
solubility for deuterium in Flibe were calculated and summarized in Table 5.
Figure 10 Schematic diagram of the cross section of the dual-probe system [37]
Table 5 Summary of diffusivity and solubility of D2 in Flibe [37]
Deuterium Diffusivity (m2/s) Solubility (mol/m3 Pa)
600 oC 8.0e-10 3.0e-04
650 oC 3.0e-09 1.0e-04
Within the scope of JUPITER-II, INL started to perform some experiments with tritium with
the Safety and Tritium Applied Research (STAR) facility[38] . The experiment setup is shown in Figure
11. Argon gas was also used to carry the tritium release to the open space above the salt. The diagnostic
system was designed to measure the concentration of tritium in a range from 0.1ppm to 10 vol% with
the combination of quadrupole mass spectrometer, gas chromatography, and proportional counter
[38]. The experiment was performed in the temperature range from 550 oC to 700 oC. The experiment
28
data were then fitted to the Arrhenius equation and the correlation of diffusion coefficient (Dflibe.T)
and solubility (Kflibe.T) were given, as shown in Equation 8 and Equation 9,
, = . × − (− ×
) ; ∈ ( , ) Equation 8
, = . × − (− ×
) ; ∈ ( , ) Equation 9
where T is the experiment temperature [38]. They concluded that tritium diffuses in Flibe in the
atomic form: bond to BeF4 2- ions or as HT, or both.
Figure 11 Diagram for tritium permeation cell [38]
Table 6 Summary of Tritium Diffusivity and Solubility in Flibe, calculated from Equation 8 and
Equation 9
550 oC 9.2355E-11 3.7467E-05
600 oC 1.9909E-10 7.0899E-05
650 oC 3.8134E-10 1.2163E-04
700 oC 6.6565E-10 1.9316E-04
3.2 Literature review on tritium transport in graphite
Tritium in graphite has been extensively studied in the design effort of fusion devices. In
fusion reactors, tritium is the fuel material and tritium can permeate through the graphite confinement,
which has adverse effect on fusion reactor design. Understanding tritium transport mechanism in
graphite is necessary for the fusion community in order to find a way to minimize tritium retention
graphite liners. This purpose is exactly opposite to that for FHR reactor design, yet these experimental
and computational studies are still quite valuable for our study due to the systematic experiment matrix
and the corresponding modeling tests.
This section will introduce the research work that is relevant to the study of hydrogen transport
behavior in FHR operation condition from mainly two scientists: H. Atsumi and R. A. Causey. Both
research on hydrogen isotopes in graphite is to investigate the performance of graphite materials as
plasma facing materials in fusion devices, while Atsumi et. al mainly used hydrogen as a surrogate and
Causey et.al mainly used deuterium or tritium generated from in-core neutron irradiation with lithium
carbonate powder containing 96% enriched 6Li.
3.2.1. Study of tritium adsorption in graphite by R.A.Causey et. al
R.A.Causey summarized four ways that tritium can be retained in graphite: “1) saturation of
the implant area; 2) co-deposition with carbon on the surfaces ; 3) adsorption on internal porosity; 4)
trans-granular diffusion with trapping”[39]. Since tritium produced in liquid Flibe in the core of FHR
is mainly in two forms: TF and T2 under about 700 °C, so process 1) and process 2) could not occur
in FHR because there are no energetic tritium ions in FHR while process 3) and process 4) are
discussed in this section.
Causey et. al discovered tritium adsorption in graphite internal porosity because of the
observation of the amount of tritium measured in the graphite near-surface (µm) and that measured
in the entire graphite sample [40]. Later he measured that tritium penetrated into POCOAXF-5Q
graphite to a depth of 2 mm at about 500 °C with deuterium-tritium plasma [41], and proposed that
the depth penetration is due to graphite internal pores which opens to the surface. This hypothesis is
confirmed by Strehlow et al’s experimental results, where he observed the direct relation between
amount of tritium measured in the mechanically removed graphite layer and the porosity of
corresponding layer [42]. Causey also mentioned that although it is observed for measurement taken
under all temperatures, the adsorption on internal porosity is more significant for low temperature
tritium exposure, which indicated that it is physi-sorption rather than chemisorption. Causey then
30
proposed that “hydrogen absorbs on carbon surfaces at different types of sites with multiple or
perhaps continuous activation energies” in the range of 0.22 to 2.2 eV/mol [39]. In this case, tritium
diffuses in graphite by jumping from one activation site to another along with possible dissociation
and recombination rather than diffuses as a molecule. He also pointed out that ion irradiation could
increase graphite porosity and increase tritium absorption by about one order of magnitude. In
addition, Causey proposed that tritium diffusion along the basal planes are faster that along the prism
planes from the observation that the activation energy of tritium diffusion in highly orientated graphite
(2.7 eV) is higher than that in isotropic graphite (1.1 eV).
By reviewing the research of other scientists, Causey et al summarized the activation energy
for tritium diffusion in graphite, heat of solution for tritium dissolution in graphite and the averaged
energy of trapping sites in graphite along with the best empirical equations for diffusivity and solubility
[39]. A full table summarizing the empirical equations for diffusivity and solubility can be found in
Table 7 and Table 8. Figure 12 plotted the tritium diffusion coefficient of graphite with the empirical
equations listed in Table 7 in the temperature range 560 – 1400 °C.
Table 7 Summary of empirical equation for tritium diffusivity in graphite
Literatures Diffusivity (cm2/s) Temperature range
(°C)
K. Rohrig et al.
HTR operating
Saeki [46] 0.00038 exp (-2.22 eV/kT) > 800
Causey et al [39] 0.93 exp (-2.8 eV/kT) > 800
31
* unpublished work at General Atomic
Table 8 Summary of empirical equation for tritium solubility in graphite
Solubility (atom
Atsumi et al [47] 6.44e-5 exp (0.2 eV/kT) > 800
Figure 12 Tritium diffusion coefficient of graphite estimated by different scientists [39]
As Causey indicated in his research, the tritium trapping sites could be from the prism planes
along the crystallite, indicating that tritium adsorption in graphite is highly dependent on graphite
microstructure [39], as shown in Figure 13, where tritium retention in different types of graphite and
carbon-carbon composite are presented.
32
Figure 13 Tritium retention at 1273 K in different graphite and carbon-carbon composites [39]
3.2.2. Study of tritium adsorption in graphite by H.Atsumi et. al
Atsumi et.al proposed the hydrogen transport mechanism after studying the thermal
desorption curve for various types of graphite, as shown in Figure 14. Unlike other related studies in
the fusion community, Atsumi used hydrogen gas as a media and aimed at determining the hydrogen
transport mechanisms.
Hydrogen molecules will arrive at the surfaces of filler grains through open pores with keeping
the state of molecules. Then the hydrogen molecules dissociate into atoms at the grain surfaces, for
the amount of hydrogen dissolves into graphite is proportional to the square root of hydrogen
pressure[48]. The hydrogen atoms migrate into a filler grain will be trapped at thermally stable defects
such as the dangling carbon bonds at edge surfaces of crystallite.
33
Figure 14 model of hydrogen diffusion in graphite, proposed by Atsumi [48].
Besides the hydrogen transport mechanism, Atsumi and his colleagues also studied the effect
of graphite microstructure on hydrogen absorption as well as the effect of irradiation. The
microstructures he focused in his study are lattice constant c0, crystallite size (La and Lc), surface area,
grain size, degree of graphitization. Neutron irradiation will also change the graphite microstructures,
and Atsumi combined the studies on effect of microstructure and irradiation in his later research
[49][50][51][52].
solubility and degree of graphitization with and
without neutron irradiation [50]
graphite with lower degree of graphitization.
And neutron irradiation decreases the degree
of graphitizaition, resulting in lower
hydrogen retention in graphite. The
relationship between lattice spacing and
degree of graphitization applied by Atsumi
is:
Equation 10
and lattice constant [54]
relationship with hydrogen retention in
graphite: graphite with larger c0 is
expected to absorb more hydrogen under
a specific condition.
non-irradiated graphite material. The
increase in c0 corresponds with increase
in hydrogen retention in graphite.
Figure 17 Relationship between hydrogen solubility
and lattice constant with and without neutron
irradiation [51]
solubility and edge surface area of crystallites
[54]
crystallite edge from the XRD diffraction
pattern. Then they plotted the hydrogen
retention vs. the estimated surface area, and the
plot showed that the larger the surface area,
more hydrogen absorbed by the graphite
material. The hypothesis is the dangling carbon
bonds are the unsaturated carbon at the edge of
the crystallite. The larger the surface area, the
more dangling bonds will exist.
They also calculated the crystallite size before
and after neutron irradiation. The relationship
between crystallite size and hydrogen retention
was plotted. The curve showed that hydrogen
retention is higher for graphite with smaller
crystallite size. This observation is in consistent
with the relationship between the surface area
and the retention, because smaller crystallite
means larger surface area with more dangling
carbon bonds.
solubility and crystallite size with and without
neutron irradiation [51]
During the experiment, Atsumi et al observed that the pressure decrease curve wouldn’t fit
into the simple diffusion model after about 10m in the case of high initial hydrogen pressure. This
observation motivated him to separate the hydrogen absorption in two trapping sites, which was
proposed first by Kanashenko et al. [18] (literature from atsumi-2003), as shown in Figure 20.
Trap sites 1 are the carbon dandling bonds
located at the interstitial cluster between the
graphene layers. The energy of trap site 2 is
estimated as 4.4ev, which is the difference
between the bond strength of C-H in
hydrocarbon (4.45ev*2) with the dissociation
energy of H2 (4.5ev)
proposed by Kanashenko and modified by
Atsumi et.al [55] [56]
Trap sites 2 are the dandling carbon bonds
located at the surface of crystallite. The energy
of trap site 1 is estimated as 2.6ev, which is the
energy of trap site 1 (4.4ev) minus the relaxation
energy (~1ev/C)
retention corresponding to Traps 1 and 2 in ISO-
880U graphite exposed to hydrogen gas at 1273 K
[57]
dominating process in graphite for hydrogen
gas absorption, and this process is pressure
dependent, which suggests an equilibrium
state will be reached under different initial
pressure. Hydrogen absorption in trap site 1
is not dependent on the pressure, which
suggests no equilibrium state exists and the
absorption at trap site 1 is a one-sided
process.
trap site) located at the surface of the
crystallite are the barrier for hydrogen to
reach the interstitial site inside the crystallite.
After all trap sites 2 are occupied, hydrogen
could reach trap site 1 and get absorbed by
higher energy sites.
Figure 22 effect of irradiation on trap site 1 and
trap site 2 absorption [58]
Neutron irradiation will increase the
concentrations of both trap site 1 and trap site
2, which corresponding to the hydrogen
retention by trap site 1 and trap site 2. As
shown in Fig.1 in [atsumi-2009], it took about
1500s to saturate trap site 2 before irradiation,
while it took about 5000s after irradiation,
which also indicated the trap site 2
concentration increases with irradiation.
absorption in graphite pre-annealed at 1273k for 2
hours [58]
Atsumi’s research. The results showed that
graphite sample after 1273K, 2hrs annealing
showed higher concentration for trap sites
2(from the time it took to reach saturation)
compared to that for 1873K, 2hrs annealing.
Due to the decrease in trap site 2
concentration after high temperature
1 becomes prominent. [58]
absorption in graphite pre-annealed at 1873k for 2
hours [58]
[59]
“diffusion-controlled desorption
Peak 2 was attributed to
“recombination-controlled process”
crystallite edge site” with the
activation energy of 2.6 ev, and “it was
desorbed by the diffusion-controlled
was attributed to trap site 1 located at
an intercalate cluster loop with high
activation energy 4.4ev
3.2.3. Study of tritium adsorption in graphite by E. Hoinkis
E. Hoinkis did more theoretical research to determine the preferred trapping location for
hydrogen atoms. He calculated the absorption energy to be 2.5 ev/H2, which is very close to the
energy of trap site 2 (~2.6ev) calculated by Atsumi. With the information of H2 dissociation energy
of 4.52 ev/H2, the bond energy calculated from Hoinkis’s experiment was 3.5ev/H. He attributed this
bond energy to the average bond energy of armchair and Zig-Zag sites. This results was in consistent
with the bond energy calculated by Chen and Yang [60]. In Chen and Yang’s theory, the C-H bond
energy at armchair site is 3.1ev/H and that at the zig-zag site is 3.86 ev/H.
40
Carbon atoms in a perfect basal plane
are chemical inert. The carbon atoms
that are chemically active to hydrogen
in graphite are mainly those with
unused valence orbitals or with
unstable electron orbitals.
atoms can be located at the edge of a
plane/ crystallite or at the defect in
the basal plane.
face. V is a single vacancy in the basal
plane. S is a single C atom
3.3 Hydrogen isotropic effect
The isotope effect on graphite absorption showed that H2 diffuses faster than D2 under the
same condition, as shown in Figure 27. The activation energies of hydrogen and deuterium
diffuison were estemated from the slope of the plot, which is 152 Kj/mol and 158 kj/mol
respectively. Depite the discrenpencies of the dependece of diffusion coefficient on temperature,
the slop indicated that it is resonable to use hydrogen as a sorragate in terms of cost effectiveness
and safety for comparasion test and verification test.
41
Figure 27 Temperature dependence of hydrogen/deuterium absorption in IG-110U at 10kPa [52]
3.4 Literature review on liquid salt and graphite interaction
No literatures have been found to study the hydrogen transport at the interface, but there are
a few studies on the chemical interaction and physical interaction between fluoride salts and carbon
materials.
The Molten Salt Reactor Experiment (MSRE), which was successfully operated in 1960s, also
used Flibe based fuel salt and nuclear grade graphite (CGB) in the core. Although CGB graphite was
engineered so that fuel salt infiltration would be minimal, the salt impregnation experiment, where
CGB graphite was exposed to fuel salt at 700 °C under 150 psig pressure, showed an average of less
than 0.2% of the bulk volume of graphite specimens were filled with salt, which was lower than the
design limit (0.5%) [62]. The radiography results of the graphite rod cross section showed that liquid
salt penetrated deeper in graphite with longer exposure time, as shown in Figure 28.
42
Figure 28 Radiography result of CGB graphite rod cross section (from left to right: shorter time
exposure to long time exposure) [62]
MSRE further did in-core testing with the CGB graphite samples, and the results showed a
decline in concentration of lithium and fluorine with depth (Figure 29) and the lithium-to-fluorine
ratios scattered around that for fuel salt, not that for LiF (Figure 30). It was proposed that slight
amounts of fuel migrated into graphite probably through open pores on the surface and flow the path
of micro-cracks which is a cluster of porosity in bulk graphite. The comparison of 19F concentration
detected from sample Y-7 (102 days fuel salt exposure) and X-13 (188 days exposure) indicated that
19F penetration depth didn’t change with depth, but 19F concentration is higher at the same distance
from the surface, as shown in Figure 31.
43
Figure 29 Lithium (left) and Fluorine (right) concentrations in CGB graphite sample, exposed to
molten fuel salt in the MSRE for 102 days (ORNL-4344 Section 11.3, page 148 and page 150) [63]
Figure 30 Mass concentration ratio, F /Li, vs depth for CGB graphite sample exposed to molten
fuel salt in the MSRE for 188 days. (ORNL-4344 Section 11.3, page 149) [63]
44
Figure 31 Comparison of fluorine concentrations in CGB graphite samples upon exposure to fuel
salt in MSRE for 102 days (Y-7) and for 188 days (X-13). (ORNL-4344 Section 11.3, page 149) [63]
In most recent studies, Young et al studied the chemical interaction between nuclear graphite
IG-110 and LiF-NaF-KF (FLiNaK) at 500 ºC for 16 hours at high pressure. At the exposure pressure,
salt intrudes in the graphite pores and micro-cracks. X-ray diffraction (XRD) and K-edge X-ray
absorption near-edge structure (XANES) of the salt-intruded graphite showed partial replacement of
C-H bonds in the graphite by C-F bonds [64]. He et al. studied the microstructure evolution of
FLiNaK-intruded nuclear graphite with pressure varied from 1 to 10atm at 650 ºC for 20 hours. X-
ray Diffraction (XRD) and Raman Spectroscopy showed an improvement of graphite stacking order,
reduction in d-spacing fluctuation and enlargement of the crystallite size along the c-axis; He et al.
attributes this microstructure evolution to the closure of Mrozowski cracks in graphite, under the
mechanical stresses induced by the solidification of the intruded molten salt [65].
In addition to the research on the interaction between liquid salt and nuclear grade graphite,
there are intensive studies on modifying chemical, electrochemical, and physical properties of graphite
materials with fluorination, where molten fluoride salts are used as solvents, reagents, and electrolytes
for graphite fluorination [66]–[74].
45
Graphite fluorination has been studied in battery design to increase the rechargeability of
carbon anode, optoelectronics, in photonics to change the electrical conductivity of the graphite, and
in electrolytic metal and gas production to control the performance of the graphite anode. Molten
fluoride salts are also used as coolants and fuel solvents in advanced nuclear reactors that contain
high-purity graphite components [75]–[77]. In lithium ion secondary batteries, carbon materials are
used as anode and fluorination modifies the surface structure and chemical species and improves
anode performances [66]. In Electric Double Layer Capacitors (EDLCs), fluorination of porous
carbon electrodes can improve the interactions at the electrolyte interface, increasing the specific
capacitance of the EDLCs [67]. In electrolytic production of F2 gas production from molten KF-
2HF, the formation of a C-F film on the carbon anode decreases the electroactive surface and changes
the roughness and surface heterogeneity of the carbon anode; these modifications change the shape
of the electrolytic F2 bubbles and improve the electrochemical system performance. X-ray
photoelectron spectrometer (XPS) and Nuclear magnetic resonance (NMR) showed the formation of
both covalent and semi-ionic C-F bonds upon graphite exposure to KF-2HF [68]. In the Hall-Heroult
process for the electrochemical production of aluminum from molten NaF-AlF3-CaF2 (cryolite),
fluorine evolution occurs at the carbon anode. XPS and Scanned Electron Microscopy (SEM) shows
evidence of surface fluorination of the graphite anode, which then leads to reduced wettability, causing
operational instability [69].
Fluorination of carbon nanostructures, such as graphene and single-walled carbon nanotubes
(SWNT), has also been studied to understand their chemistry and modify their electronic properties.
Fluorination of graphene-based materials by XeF2 results in the formation of insulating materials by
creating covalent C-F bonds and by modifying sp2 hybridized bonds to sp3 hybridized bonds [70].
Fluorination of a graphene film with XeF2 at 30 ºC resulted in the formation of C-F toget