Tritium Transport Analysis in Advanced Sodium-Cooled Fast Reactors

INTRODUCTION 
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INTRODUCTION 
The Sodium-Cooled Fast Reactor (SFR) system is one of six types of plants in Next Generation Nuclear 
Plant (NGNP). The NGNP [1] is in the pre-conceptual design phase with major design selections (e.g., 
reactor core type, core outlet temperature, etc.) still to be carried out. SFR features a fast-spectrum 
reactor and a closed fuel recycle system. The primary mission for SFR is the management of high-level 
wastes and, in particular, the management of plutonium and other actinides [1]. With innovations to 
reduce capital costs, the mission can extend to electricity production, given the proven capability of 
sodium reactors to utilize almost all of the energy in the natural uranium versus the 1% utilized in 
thermal spectrum systems [1].  
One potential problem of using SFR is tritium permeation from the primary coolant through heat 
exchangers and other plant facilities to the environment. In SFR tritium mostly comes from ternary 
fission of the fuel and neutron capturereactions inside boron-containing materials, such as control 
rods and neutron flux shielding blocks, as shown in Chapter 4. Tritium that enters in the primary 
coolant will be circulated or permeated to the secondary coolant through the intermediate heat 
transfer loop. The permeated tritium,successively, enters the product steam/water into steam 
generator through heat exchanger surfaces. The mechanisms of tritium transport are diffusion, bulk 
transport, and permeation (seeFig. 1). 
Transport of different chemical species of tritium in the environment (i.e. HT and HTO) is related to 
physical and chemical processes. Physical processes are bulk transport (tritium moves because of its 
dissolution inside heat transfer fluids) and diffusional transport (tritium motion is driven by 
concentration gradient). Reactions and state changes of chemical species are chemical processes [2]. A 
tritium permeation model, thus is required in order to estimate the total amount of tritium released 
into the environment and circulating inside the plant. In fact the model is applied to the overall SFR 
plant, considering all tritium transport processes inside nuclear plant installations and studying 
systematically tritium transport in each component. The objective in this work is exactly to simulate 
tritium transport behavior in SFR components (according to reference SFR configuration reported in 
Fig. 1) and to predict tritium quantities in different SFR devices by means of solving mass conservation 
laws with computational tools (such as MATLAB packages). The tritium path in a SFR is shown in Fig. 1.  
Recently, other authors developed a tritium permeation analysis code (TPAC) for Very High 
Temperature Gas Reactors  [3] and the current work was deeply inspirited to this permeation code by 
the computational and mathematical structure point of view.
INTRODUCTION 
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Fig. 1 Diffusion, bulk transport and permeation pathways of tritium in a notional heat transport loop [3] 
The programmatic benefits of performing the subject tritium analysis during pre-conceptual design are 
three: 
1) Design Data Needs (DDNs) can be identified during the design phase and generic NGNP 
technology development programs can be modified or supplemented as required; 
2) inadequacies in the current analytical tools to support preliminary and final NGNP design can 
be identified such that these deficiencies can be remedied during the conceptual design 
phase; 
3) significant design issues that need to be addressed during conceptual and preliminary design 
can be identified earlier. 
The DDNs with regard to H
null transport, especially the need for better characterization of 
tritiumpermeabilities adopted in a SFRs and release rates from fuel and control rod materials, were 
confirmed. The expectation that tritium contamination of plant installation in SFRs will be a significant 
design issue was confirmed [4] and design options for resolving the issue were identified. 
Tritium is a radioactive isotope of hydrogen with the half life of 12.32 years. The nucleus of a tritium 
atom consists of a proton and two neutrons. This contrasts with the nucleus of an ordinary hydrogen 
atom and a deuterium atom. Ordinary hydrogen comprises over 99.9%, deuterium comprises 0.02%, 
and tritium comprises about a 10
null null null % of naturally occurring hydrogen (Ref.[3]) The physical and 
chemical properties of tritium are very close to those of hydrogen [2].Typically, tritium exists as a form 
of HT tritium gas (H
null − H
null ) because of isotope exchange reactions between T
null (H
null − H
null ) and H
null [3]. 
Tritiated water, HTO, is another common form of tritium. In tritiated water, a tritium atom replaces 
one of the hydrogen atoms so the chemical form is HTO rather than H
null O. Tritium in the environment 
has three sources: natural production, release from atmospheric weapon tests, and routine or
INTRODUCTION 
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accidental releases from the nuclear industry [2]. The main sources of tritium released by man are 
linked to the nuclear power cycle: nuclear power stations, nuclear fuel reprocessing plants or tritium 
production plants [2]. Most significant leakages are due to reprocessing plants; releases are essentially 
liquid, and reach several PBq annually [2]. In the near future, some applications could lead to new 
sources of tritium. The future usage of tritium mainly concerns nuclear fusion facilities (like the 
International Thermonuclear Experimental Reactor, ITER). The quantity of tritium used by ITER would 
amount to 1.5 kg/y[2]; the associated release in the atmosphere could be significant.  
Inhalation, ingestion and skin exposure are the three main routes of exposure to tritium for man [2]. 
The radiological impact of tritium results from the combination of the characteristics and the behavior 
of the radionuclide. Some parameters enhance the radiotoxicity of tritium: for example, its radioactive 
half-life relatively long (12.35 years), its uptake by humans is likely to take place and as an isotope of 
hydrogen, it has a high biological importance [2]. The decay of tritium is given by: 
H
null null → H
null null e + 	 β
null null null + ν
null null 	 (1)  
The null (or electron)'s kinetic energy varies, with an average of 5.7 keV[5], and a maximum of 19 keV[6] 
while the remaining energy is carried off by the nearly undetectable electron antineutrino null null null . The 
corresponding maximum track length is 6m in water or biological tissue and the thickness of the 
epidermis and dermis of human skin is 20–100m and 1–3 mm, respectively [2]. As a consequence, 
tritium could inflict damage on humans only when it is present inside the body [2]. 
The radiotoxicity of tritium is relatively low: the dose coefficients per unit of incorporation (the 
effective dose per built-in Becquerel) have been evaluated at 1.8 × 10
null null null 	Sv/Bq for HT (inhalation), 
1.8 × 10
null null null 	Sv/Bq for HTO (ingestion or inhalation) [2]. The occurrence of a cancer due to chronic 
exposure to tritium cannot be excluded, even if epidemiological data are not sufficient to quantify the 
risk yet [2]. As for other radionuclides, the theoretical risk of death by cancer due to tritium 
incorporation has been calculated to be 6.5 × 10
null null Sv
null null incorporated [2]. The World Health 
Organization (WHO) places the limit at an annual effective dose of 0.1mSv, corresponding to the daily 
consumption of tritiated water at a concentration of 10000 	Bq/L during a year. 
All dosimetric information on tritium are reported in Tab. 1
INTRODUCTION 
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Tab. 1Dosimetric information of Tritium [6]
DESCRIPTION OF A SFR POOL TYPE DESIGN 
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Chapter 1 
1 DESCRIPTION OF A SFR POOL TYPE DESIGN 
In this chapter some design issues proposed for SFR (Sodium Cooled Fast Reactors) are briefly 
described. It is also aimed to show the technical and engineering characteristics of this type of plant, in 
order to make the reader possible to understand the critical aspects of sodium fast reactors regarding 
the tritium transport mechanisms inside this nuclear power plant typology, well illustrated in Chapter 
4. This descriptive sectionis mainly focused on pool-type design of sodium fast reactors, which is one 
of the two most important configurations adopted for this technology. 
The descriptions are intentionally concise in order to give a general horizontal view of the spectrum of 
possible milestones regarding a tritium permeation analysis. 
1.1 General Description 
A fast neutron reactor or simply a fast reactor is a category of nuclear reactor in which the fission 
chain reaction is sustained by fast neutrons. Such a reactor needs no neutron moderator, but must use 
fuel that is relatively rich in fissile material when compared to that required for a thermal reactor  
The Sodium-Cooled Fast Reactor (SFR) system features a fast-spectrum reactor and closed fuel recycle 
system. The majority of natural uranium is the isotope U
null null null making up about 99.3 %. The remaining 
0.7 % is U
null null null the isotope required for thermal fission in modern light water reactors. The fast 
neutrons are used to breed plutonium from U
null null null and these plutonium isotopes then undergo fission 
to produce heat. Therefore fast reactors can utilize uranium much more efficiently than thermal 
reactors. Since water acts as a moderator and will slow neutrons out the fast spectrum, liquid metals, 
such as sodium, are used as coolants in these fast reactors transferring the heat to a power conversion 
system used to produce electricity. 
Beside the advantage of more efficient use of natural uranium, SFRs can also be used to breed fuel; in 
fact they can be designed to produce more fuel than they consume, by use of an external ring of 
U
null null null (the blanket) where the plutonium is bred and subsequently recycled.
DESCRIPTION OF A SFR POOL TYPE DESIGN 
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The primary mission for the SFR is management of high-level wastes and, in particular, management 
of plutonium and other actinides [1]. With innovations to reduce capital cost, the mission can extend 
to electricity production, given the proven capability of sodium reactors to utilize almost all of the 
energy in the natural uranium versus the 1% utilized in thermal spectrum systems. Although it is 
currently (2010) uneconomic [1], a fast neutron reactor can reduce the total radiotoxicity of nuclear 
waste, and dramatically reduce the waste's lifetime.They can also use all or almost all of the fuel in the 
waste. Fast neutrons have an advantage in the transmutation of nuclear waste. With fast neutrons, 
the ratio between splitting and the capture of neutrons of plutonium or minor actinide is often larger 
than when the neutrons are slower, at thermal or near-thermal "epithermal" speeds. The transmuted 
odd-numbered actinides (e.g. from Pu-240 to Pu-241) split more easily. After they split, the actinides 
become a pair of "fission products." These elements have less total radiotoxicity. Since their fission 
products have a maximum half life of 27 years,the result is to reduce nuclear waste lifetimes from tens 
of millennia (from transuranic isotopes) to a few centuries. The processes are not perfect, but the 
remaining transuranics are reduced from a significant problem to a tiny percentage of the total waste, 
because any larger amounts can be used as fuel [7]. 
Two fuel options exist for the SFR: MOX and mixed uranium-plutonium-zirconium metal alloy (metal) 
[1]. The experience with MOX fuel is considerably more extensive than with metal [2]. SFRs require a 
closed fuel cycle to enable their advantageous actinide management and fuel utilization features. Both 
are highly developed as a result of many years of work in several national reactor development 
programs. Burnups in the range of 150–200 GWd/Mton have been experimentally demonstrated for 
both [8]. Nevertheless, the databases for oxide fuels are considerably more extensive than those for 
metal fuels [8]. 
There are two primary fuel cycle technology options: an advanced aqueous process, and  the 
pyroprocess, which derives from the term pyrometallurgical process. Both processes have similar 
objectives which are the recovery and recycle of 99.9% of the actinides and the achievement of low 
decontamination factor of the product.The technology base for the advanced aqueous process comes 
from the long and successful experience in several countries with PUREX process technology 
These fuel cycle technologies must be adaptable to thermal spectrum fuels in addition to serving the 
needs of the SFR. This is needed for two reasons: first, the startup fuel for the fast reactors must come 
ultimately from spent thermal reactor fuel; second, for the waste management advantages of the 
advanced fuel cycles to be realized (namely, a reduction in the number of future repositories required 
and a reduction in their technical performance requirements), fuel from thermal spectrum plants will 
need to be processed with the same recovery factors [1]. Thus, the reactor technology and the fuel
DESCRIPTION OF A SFR POOL TYPE DESIGN 
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cycle technology are strongly linked. Consequently, much of the research recommended for the SFR is 
relevant to crosscutting fuel cycle issues. 
A range of plant size options are available for the SFR, ranging from modular systems of a few hundred 
null null null null null to large monolithic reactors of 1500–1700 null null null null null . [1] Sodium core outlet temperatures are 
typically 530–550ºC [1]. The primary coolant system can either be arranged in a pool layout 
schematized in Figure 1-1(a common approach, where all primary system components are housed in a 
single vessel), or in a compact loop layout (see Figure 1-2), favored in Japan. 
 
Figure 1-1 Pool Type SFR Configuration [1] 
In the pool type design the reactor core, primary pumps, IHXs (Intermediate Heat Exchangers) and 
DHXs (Decay Heat Exchanger) are all immersed in a pool of sodium coolant within the reactor vessel, 
so the primary radioactive sodium does not leave the reactor vessel during normal operation. In this 
way LOCAs (Loss Of Coolant Accidents) affecting the primary sodium are extremely unlikely. The 
primary pump is totally submerged in the vessel, so the risk of having large sodium pipes under the 
core level (which could lead the core to become uncovered after a large LOCA) is eliminated. A 
disadvantage of this configuration is that a larger reactor vessel is required to host the heat
DESCRIPTION OF A SFR POOL TYPE DESIGN 
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exchangers and a larger quantity of primary sodium is needed. Moreover, the reactor vessel requires 
more complex internal structures [9]. 
In the loop type SFR design the primary coolant is circulated through pumps and IHXs in pipes which 
are external to the reactor tank. The main advantages of this design are: compactness, easier in-
service inspection and maintenance, smaller amount of sodium needed. The big disadvantage of this 
configuration is the possibility of a sodium leakage, which is much higher than in a pool-type SFR. 
 
Figure 1-2 Loop Type SFR Configuration [10] 
For both options,there is a relatively large thermal inertia of the primarycoolant[1]. A large margin to 
coolant boiling is achievedby design, and is an important safety feature of thesesystems. Another 
major safety feature is that the primarysystem operates at essentially atmospheric 
pressure,pressurized only to the extent needed to movefluid. Sodium reacts chemically with air, and 
withwater (see par. 2.4.4), and thus the design must limit the potential forsuch reactions and their 
consequences. To improvesafety, a secondary sodium system acts as a bufferbetween the radioactive 
sodium in the primary systemand the steam or water that is contained in the conventionalRankine-
cycle power plant. If a sodium-waterreaction occurs, it does not involve a radioactive release. 
1.2 Technology base for SFR 
Sodium-cooled liquid metal reactors are the most technologically developed of the six Generation IV 
systems [8]. SFRs have been built and operated in France, Japan, Germany, the United Kingdom, 
Russia and the United States. Demonstration plants ranged from 1.1 MW
null null (at EBR-I in 1951) to 1200 
MW
null null null (at Super-Phoenix in 1985), and sodium-cooled reactors are operating today in Japan, France,
DESCRIPTION OF A SFR POOL TYPE DESIGN 
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and Russia [8]. As a benefit of these previous investments in technology, the majority of the R&D 
needs presented for the SFR in this roadmap are performance-related [8]. With the exception of 
passive safety assurance, there are few viability issues with regard to the reactor systems.  
There is an extensive technology base in nuclear safety that establishes the passive safety 
characteristics of the SFR and their ability to accommodate all of the classical anticipated transients 
without scram events and without fuel damage. Landmark tests of two of these events were done in 
RAPSODIE (France) in 1983 and in EBR-II (United States) in 1986. Still, there is important viability work 
to be done in safety. Key needs are to confirm reliability of passive feedback from heat-up of reactor 
structures and to establish the long-term coolability of oxide or metal fuel debris after a bounding case 
accident. 
The technology base for fabrication of oxide fuel assemblies is substantial, yet further extension is 
needed to make the process remotely operable and maintainable. The high-level waste form from 
advanced aqueous processing is vitrified glass, for which the technology is well established [8]. The 
pyroprocess has been under development since the inception of the Integral Fast Reactor program in 
the United States in 1984 [8]. When the program was cancelled in 1994, pyroprocess development 
continued in order to treat EBR-II spent fuel for disposal [8]. In this latter application, plutonium and 
minor actinides were not recovered and pyroprocess experience with these materials remains at 
laboratory scale. However,batch size for uranium recovery is at the tens-of-kilogram scale, about that 
needed for deployment. Remote fabrication of metal fuel was demonstrated in the 1960s. Significant 
work has gone into repository certification of the two high-level waste forms from the pyroprocess, a 
glass-bonded mineral (ceramic) and a zirconium stainless steel alloy [8]. 
1.3 Description of SFR components 
Since the objective of this work is to develop a tritium permeation code which needs different 
technical, nuclear and thermal-physical data of SFR pool type design reference configuration, in this 
section a brief qualitative description is performed for all components in which tritium transport is 
analyzed.In Figure 4-32is reported a flow diagram of an overall SFR plant, called Prototype Fast 
Breeder Reactor (PFBR) 
1.3.1 Core and main vessel. 
Figure 1-3shows an example of a core configuration. A homogenous core concept with different fissile 
enrichment zonesis adopted for power flattening. The active core(inner core and outer core indicated 
in yellow and in red respectively in Figure 1-3),where most of the nuclear heat is generated, consists of 
a certain numberof fuel subassemblies. Each fuel subassembly contains also a discrete quantityof 
bonded pins with a given outer diameter. Each pin has a column of annular MOX fuel pellets and
DESCRIPTION OF A SFR POOL TYPE DESIGN 
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anupper and lower blanket columns (Outer core in red indicated in Figure 1-3). Thelinear power in the 
fuel pin is of order of 450 W/cm[11].  
 
Figure 1-3 SFR Core Configuration [12] 
Blanket assemblies are special fertile elements disposed to enhance breeding. In a common 
configuration the active fuel assemblies are located in the middle of the core and the blanket 
assemblies are all around. The blanket containsfertile material (U-238 and/or Th-232) which is 
converted into fissile materialthrough neutron capture.The use of a blanket was present in many 
designs of the past. In GEN-IVprojects, for proliferation resistance reasons, the presence of assemblies 
with high 239Pu content is not desired anymore. It is preferable to adopt a homogeneous core, where 
the same fuel assemblies contain both fissile and fertile material, performing both burning and 
breeding, with a conversion ratio around 1[9]. 
Inside core subassembly we find also a certain number of absorber rods, composed bya portion of 
controland safety rods (CSR) and diverse safety rods (DSR),oftenarranged in rings ([11], [12] as 
reported in Figure 1-3). Others independent and diverse shut downsystems are provided for ensuring 
safe shut down of the reactoreven when one system is not available. Both the systems aredesigned to 
shutdown the reactor as fastest as possible. In addition,axial shielding is provided within the 
subassemblies and radialshielding subassemblies are provided within the core. These areoptimized in 
order to have the required flux at in-vessel neutrondetector locations. They also serve the purpose of 
limiting theactivation of the secondary sodium and radiation damage of gridplate. In SFR subassembly, 
neutron flux shielding are in general characterized by boron carbide B
null C shields (blue elements in 
Figure 1-3), because of its large neutron captures cross sections (see Table 4-1).  
The use of Reflector materials (light blue elements of Figure 1-3) around the active region decreases 
the leakage of neutrons from the core. This has a positive balance in neutron economy. Another good