Structural Materials for Fusion Reactors

Structural materials for fusion reactors represent one of the most challenging materials engineering problems facing humanity. The 14.1 MeV high-energy neutrons produced by D-T fusion reactions subject materials to an irradiation environment far more severe than any existing nuclear reactor. To simultaneously meet the conflicting requirements of improving power generation efficiency, enhancing maintainability, and reducing radioactive waste, development of low-activation materials to replace conventional stainless steels is being pursued on a global scale.

This chapter systematically covers the requirements for fusion reactor structural materials, the main candidate materials currently under development, mechanisms of material damage from neutron irradiation, materials testing facilities, and the databases and standards necessary for design.

Required Properties for Structural Materials

Structural materials for fusion reactors must simultaneously satisfy a wide range of performance requirements that are qualitatively different from those for conventional nuclear reactors or thermal power plants. Understanding these requirements is essential for grasping the direction and challenges of materials development.

Radiation Resistance

The first wall and blanket structural materials of a fusion reactor are exposed to enormous quantities of high-energy neutrons throughout the operational period. Neutron irradiation causes atomic-level defects to accumulate in the material, resulting in significant degradation of mechanical properties.

In commercial fusion reactors, structural materials are expected to receive irradiation doses of approximately 100-200 dpa (displacements per atom) during their operational lifetime. This far exceeds the irradiation doses experienced by materials in current fission reactors (several tens of dpa). Furthermore, the neutron energy spectrum differs significantly from fission reactors, and changes in material composition due to transmutation reactions become pronounced.

Structural materials must be able to withstand such high fluence (accumulated irradiation) and maintain structural integrity over extended periods. Specifically, the following properties are required:

• Suppression of irradiation hardening
• Suppression of ductile-to-brittle transition temperature (DBTT) increase
• Suppression of swelling (irradiation-induced volume expansion)
• Suppression of irradiation creep
• Suppression of helium embrittlement

High-Temperature Strength

Increasing power generation efficiency requires higher coolant temperatures. From the perspective of Carnot efficiency, the efficiency $\eta$ of a heat engine is limited by the high-temperature side temperature $T_H$ and the low-temperature side temperature $T_L$ according to the following equation:

$$\eta_{\text{Carnot}} = 1 - \frac{T_L}{T_H}$$

In actual power generation systems, higher coolant outlet temperatures allow higher steam turbine inlet temperatures, improving power generation efficiency. Current light water reactor power generation efficiency is approximately 33%, but fusion power plants aim for 40% or higher, and eventually 50% or higher.

Therefore, structural materials are required to have sufficient creep strength and tensile strength at operating temperatures. Creep is a phenomenon where materials deform over time under constant stress, becoming more pronounced at higher temperatures.

The strain rate $\dot{\varepsilon}$ in the steady state of creep deformation is expressed as a function of stress $\sigma$ and absolute temperature $T$ as follows:

$$\dot{\varepsilon} = A \sigma^n \exp\left(-\frac{Q}{RT}\right)$$

Here, $A$ is a material constant, $n$ is the stress exponent (typically 3-8), $Q$ is the activation energy, and $R$ is the gas constant. This equation shows that the creep rate increases exponentially with temperature.

Low-Activation Properties

Low-activation properties are an important requirement unique to fusion reactor structural materials. Materials subjected to neutron irradiation become activated and remain radioactive for extended periods after reactor shutdown. The magnitude and duration of this induced radioactivity vary greatly depending on the types of elements composing the material.

The advantages of low-activation materials are as follows:

1. Improved maintainability: After reactor shutdown, decay heat and radiation levels decrease relatively quickly, facilitating remote maintenance operations
2. Simplified waste management: If generation of long-lived nuclides is minimized, the disposal classification of radioactive waste can be reduced, alleviating the burden on final disposal sites
3. Recycling possibility: If materials can be reused (recycled) after approximately 100 years of cooling, effective use of resources and reduction of waste volume become possible

Elements suitable for low activation include Si, V, Cr, Ti, Fe, C, and W. Conversely, Ni, Mo, Nb, and Co are elements to be avoided as they generate long-lived radioactive nuclides.

As an indicator of low activation, the time variation of specific activity (radioactivity per unit mass) is used. Goals for fusion reactor materials include achieving levels that allow hands-on maintenance approximately 100 years after shutdown, or meeting criteria for shallow land burial.

Thermal Properties

The thermal properties of structural materials significantly affect reactor design and safety.

Thermal conductivity: Materials with high thermal conductivity can efficiently transfer heat from the plasma to the coolant. This suppresses temperature rise at the first wall surface and reduces material degradation. Heat flux $q$ by thermal conduction is expressed by Fourier’s law as:

$$q = -k \nabla T$$

Here, $k$ is the thermal conductivity and $\nabla T$ is the temperature gradient.

Coefficient of thermal expansion: Since fusion reactors undergo repeated startup-shutdown cycles, structural materials experience large temperature changes. At joints between dissimilar materials, thermal stresses arise due to differences in thermal expansion coefficients. Materials with small thermal expansion coefficients and small differences from joined materials are desirable.

Specific heat: Specific heat affects the rate of temperature change during transient conditions. Adequate specific heat is important for mitigating temperature rise during abnormal events (such as plasma disruptions).

Chemical Properties

Since structural materials are in direct contact with coolants and tritium breeding materials, chemical compatibility is important.

Compatibility with coolants: Candidate coolants under consideration include water, helium gas, liquid lithium, and liquid lead-lithium alloy (LiPb). Material selection must consider reactivity and corrosion characteristics with each coolant.

Interaction with hydrogen isotopes: Tritium dissolves, diffuses, and permeates through structural materials. From the perspective of tritium confinement and recovery, properties such as hydrogen solubility, diffusion coefficient, and permeation coefficient are important.

Mechanical Workability

From an industrial materials perspective, the following processing characteristics are necessary:

• Weldability: Welding is essential for fabricating large structures. Ensuring the soundness and reliability of weld zones is important
• Formability: Plastic workability for fabricating components with complex shapes
• Machinability: Mechanical workability for achieving precise dimensional accuracy
• Industrial-scale manufacturability: Ability to manufacture the large quantities of materials needed for fusion reactors while maintaining quality

Reduced Activation Ferritic/Martensitic Steel (RAFM Steel)

Reduced Activation Ferritic/Martensitic Steel (RAFM steel) is the most advanced in development among fusion reactor structural materials and is the primary candidate for demonstration power reactors.

Development History

RAFM steel was developed based on conventional high-chromium ferritic/martensitic steels (e.g., HT-9, Modified 9Cr-1Mo steel). These conventional steel grades had been demonstrated to exhibit excellent radiation resistance as fuel cladding and duct materials for fast breeder reactors.

However, conventional steel grades contained elements such as Mo, Nb, and Ni that generate long-lived radioactive nuclides. RAFM steel was developed by replacing these elements with low-activation elements such as W, V, and Ta. This replacement successfully achieved low-activation properties while maintaining high-temperature strength and radiation resistance.

Major RAFM Steel Grades

The major RAFM steel grades developed worldwide and their chemical compositions are shown below.

F82H (Japan)

F82H is a RAFM steel developed by the Japan Atomic Energy Research Institute (now Japan Atomic Energy Agency, JAEA) and was the first low-activation steel manufactured on an industrial scale in the world. Trial production at a 5-ton scale began in 1991, followed by 15-ton scale production.

Standard chemical composition (mass%):

Element	Composition Range	Target Value
C	0.08-0.12	0.10
Cr	7.5-8.5	8.0
W	1.8-2.2	2.0
V	0.15-0.25	0.20
Ta	0.02-0.08	0.04
Mn	0.3-0.6	0.5
Si	0.05-0.15	0.10
N	≤0.02	-
Fe	Bal.	-

The main characteristics of F82H are as follows:

• Complete replacement of Mo with W achieves low activation
• Minor Ta addition achieves grain refinement and carbide stabilization
• Homogeneous and stable mechanical properties
• Good weldability

Mechanical properties (after normalizing and tempering):

Property	Room Temperature	400°C	550°C
0.2% Yield Strength (MPa)	520	420	350
Tensile Strength (MPa)	650	510	400
Elongation (%)	22	18	20
Reduction of Area (%)	75	75	80

EUROFER97 (Europe)

EUROFER97 is a RAFM steel developed under the European Union’s fusion research and development program. Through joint development by research institutions and steel companies across European countries, the first prototype was manufactured in 1997.

Standard chemical composition (mass%):

Element	Composition Range	Target Value
C	0.09-0.12	0.11
Cr	8.5-9.5	9.0
W	1.0-1.2	1.1
V	0.15-0.25	0.20
Ta	0.10-0.14	0.12
Mn	0.4-0.6	0.5
Si	≤0.05	-
N	0.015-0.045	0.03
Fe	Bal.	-

EUROFER97 is characterized by slightly higher Cr content and higher Ta content compared to F82H. The increased Ta achieves austenite phase stabilization and improved creep strength.

Mechanical properties (after normalizing and tempering):

Property	Room Temperature	400°C	550°C
0.2% Yield Strength (MPa)	530	430	360
Tensile Strength (MPa)	670	530	420
Elongation (%)	20	17	18
Reduction of Area (%)	73	73	78

JLF-1 (Japan)

JLF-1 (Japanese Low-activation Ferritic steel No. 1) is a RAFM steel developed primarily by a consortium of Japanese universities (Kyoto University, Tohoku University, etc.).

Standard chemical composition (mass%):

Element	Target Value
C	0.10
Cr	9.0
W	2.0
V	0.20
Ta	0.07
Mn	0.45
N	0.05
Fe	Bal.

JLF-1 has an intermediate composition with W content similar to F82H and Cr content similar to EUROFER97. Its relatively high N content is expected to provide dispersion strengthening effects through VN precipitates.

CLAM (China)

CLAM (China Low Activation Martensitic) steel is a RAFM steel developed primarily by the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP).

Standard chemical composition (mass%):

Element	Target Value
C	0.10
Cr	9.0
W	1.5
V	0.20
Ta	0.15
Mn	0.5
Si	0.05
Fe	Bal.

CLAM steel is developed using Chinese indigenous manufacturing technology while referencing the composition of EUROFER97. Application to China’s fusion program CFETR (China Fusion Engineering Test Reactor) is planned.

Microstructure and Strengthening Mechanisms

The excellent mechanical properties of RAFM steel derive from their characteristic microstructure and combined strengthening mechanisms.

Martensitic Transformation

RAFM steel forms a martensitic structure when cooled by air or oil from the austenitizing temperature (typically 950-1050°C). This martensitic transformation is a diffusionless transformation characteristic of Fe-Cr-C alloys, where the parent phase (austenite) lattice shears to form the body-centered cubic (bcc) or body-centered tetragonal (bct) martensite phase.

The formed martensite contains a high density of dislocations and supersaturated carbon in solid solution. In this state, the material is hard but brittle, so tempering treatment (typically at 720-780°C) is applied. Tempering causes supersaturated carbon to precipitate as carbides and allows dislocation recovery and rearrangement, optimizing the balance between strength and toughness.

Strengthening Mechanisms

The strength of RAFM steel is achieved through the following combined strengthening mechanisms:

1. Solid solution strengthening: Alloying elements such as Cr and W dissolve in the Fe lattice, creating lattice strain that impedes dislocation motion

2. Precipitation strengthening: Carbides (mainly M₂₃C₆ type) precipitated during tempering provide dislocation pinning effects. Fine MX-type carbonitrides such as VN and TaC also contribute to precipitation strengthening

3. Dislocation strengthening: Dislocations introduced by martensitic transformation and processing interact with each other, strengthening the material

4. Grain refinement strengthening: Fine grains strengthen the material because grain boundaries function as barriers to dislocation motion. The Hall-Petch relationship expresses the relationship between yield stress $\sigma_y$ and grain size $d$ as:

$$\sigma_y = \sigma_0 + k_y d^{-1/2}$$

Here, $\sigma_0$ is the lattice friction stress and $k_y$ is the Hall-Petch coefficient.

Heat Treatment Conditions and Properties

The properties of RAFM steel vary greatly depending on heat treatment conditions. Standard heat treatment conditions and their effects are shown below.

Normalizing

Normalizing is a treatment that heats the material above the austenitizing temperature and air cools it.

Standard conditions:

• F82H: 1040°C × 40 min → air cooling
• EUROFER97: 980°C × 30 min → air cooling

This treatment forms a uniform martensitic structure.

Tempering

Tempering is applied to the martensitic structure after normalizing.

Standard conditions:

• F82H: 750°C × 1 hour → air cooling
• EUROFER97: 760°C × 90 min → air cooling

Tempering temperature and time significantly affect the final properties of the material.

Tempering Condition	Hardness	Toughness	Creep Strength
Low temperature/short time	High	Low	High
High temperature/long time	Low	High	Somewhat lower

Operating Temperature Range and Limits

The operating temperature range of RAFM steel is limited by material properties at both lower and upper limits.

Lower Limit Temperature (approximately 350-400°C)

The lower limit temperature is primarily determined by embrittlement due to neutron irradiation. Upon irradiation, the ductile-to-brittle transition temperature (DBTT) increases, raising the risk of brittle fracture at low temperatures.

The DBTT shift $\Delta\text{DBTT}$ increases with irradiation dose (dpa) and tends to saturate eventually. For F82H, DBTT shifts of 200-250°C after approximately 10 dpa irradiation have been reported.

Brittle fracture is prevented by setting the operating temperature sufficiently above the DBTT. Considering safety margins, the lower limit temperature is set at approximately 350-400°C.

Upper Limit Temperature (approximately 550°C)

The upper limit temperature is primarily determined by creep strength and softening due to thermal aging.

The creep rupture time $t_r$ of RAFM steel can be organized as a function of temperature and stress using the Larson-Miller parameter $P_{LM}$:

$$P_{LM} = T (\log t_r + C)$$

Here, $T$ is the absolute temperature (K), $t_r$ is the rupture time (hours), and $C$ is a material constant (approximately 20 for RAFM steel).

At temperatures above 550°C, the decrease in creep strength becomes pronounced, significantly limiting design stresses. Additionally, prolonged high-temperature exposure causes coarsening of M₂₃C₆ carbides, reducing strength.

Welding Technology

Welding of RAFM steel is extremely important from the perspective of structural fabrication.

Welding Methods

The following welding methods have been applied to RAFM steel:

1. TIG welding (GTAW): The most commonly used method. Good weld quality is obtained, but welding speed is slow

2. Electron beam welding (EBW): A method using high energy density electron beams. Deep penetration is possible with low heat input

3. Laser welding: A method using high energy density laser beams. High-speed welding is possible

4. Hot isostatic pressing (HIP) bonding: A method that performs solid-state bonding under high temperature and pressure. Low risk of weld defects since no melting is involved

Post-Weld Heat Treatment (PWHT)

Weld zones in RAFM steel become as-quenched martensite, which is hard and brittle. Therefore, post-weld heat treatment (PWHT) is essential.

PWHT conditions:

• F82H: 720-750°C × 1-2 hours
• EUROFER97: 750-760°C × 2 hours

PWHT tempers the martensite in the weld zone, restoring toughness. However, repeated PWHT causes carbide coarsening, so the number of welds (repairs) is limited.

Oxide Dispersion Strengthened (ODS) Steel

Oxide Dispersion Strengthened (ODS) steel is an advanced material based on RAFM steel with dramatically improved high-temperature strength through dispersion of nanometer-sized oxide particles.

Concept and Manufacturing Method

The concept of ODS steel is to disperse highly stable oxide particles (mainly Y₂O₃ or Y₂Ti₂O₇) finely in steel to impede dislocation motion. These oxide particles are thermodynamically extremely stable and resist coarsening even during long-term high-temperature exposure.

ODS steel cannot be manufactured by conventional melting methods. Mechanical alloying (MA) is used to uniformly disperse oxide particles.

Manufacturing Process

1. Raw powder preparation: Matrix steel powder and Y₂O₃ powder are blended in specified ratios

2. Mechanical alloying: Powders are mixed and milled for extended periods (tens of hours) in a high-energy ball mill. During this process, Y₂O₃ dissolves and decomposes into the steel powder

3. Consolidation: MA powder is densified by hot isostatic pressing (HIP) or hot extrusion

4. Hot working: Processing to final shape by extrusion, rolling, forging, etc.

5. Heat treatment: Microstructure adjustment through normalizing and tempering

During heat treatment, dissolved Y and O recombine and precipitate as nanometer-sized oxide particles (Y₂Ti₂O₇, Y₂O₃, etc.).

Major ODS Steel Grades

9Cr-ODS (Japan)

9Cr-ODS steel, developed primarily by the Japan Atomic Energy Agency (JAEA) and Kyoto University, is being developed as an advanced cladding material for fusion reactors and fast breeder reactors.

Standard chemical composition (mass%):

Element	Composition
Cr	9.0
W	2.0
Ti	0.20
Y₂O₃	0.35
C	<0.02
Fe	Bal.

12Cr-ODS (Japan)

12Cr-ODS steel has higher Cr content than 9Cr-ODS steel and exhibits superior oxidation and corrosion resistance.

Standard chemical composition (mass%):

Element	Composition
Cr	12.0
W	2.0
Ti	0.30
Y₂O₃	0.25
C	<0.02
Fe	Bal.

14Cr-ODS (Europe)

14Cr-ODS steel developed in Europe has a single-phase ferritic structure and does not undergo martensitic transformation.

Standard chemical composition (mass%):

Element	Composition
Cr	14.0
W	1.0
Ti	0.30
Y₂O₃	0.30
Fe	Bal.

Microstructure and Strengthening Mechanism

The outstanding high-temperature strength of ODS steel originates from the dispersion strengthening mechanism by oxide particles.

Oxide Particle Characteristics

The characteristics of representative oxide particles are shown below.

Oxide	Crystal Structure	Density (g/cm³)	Melting Point (°C)	Average Particle Size (nm)
Y₂O₃	Cubic	5.0	2410	5-30
Y₂Ti₂O₇	Pyrochlore	5.3	1825	2-10
Y₄Al₂O₉	Monoclinic	4.4	1940	5-20

Y₂Ti₂O₇ is known to disperse more finely than Y₂O₃ and has superior high-temperature stability.

Orowan Mechanism

Strengthening by dispersed particles is explained by the Orowan mechanism. Additional energy is required for dislocations to bypass (circumvent) particles, which leads to material strengthening.

The Orowan stress $\tau_{\text{Orowan}}$ is expressed as:

$$\tau_{\text{Orowan}} = \frac{0.4 G b}{\pi \sqrt{1-\nu}} \cdot \frac{\ln(d_p / b)}{\lambda}$$

Here, $G$ is the shear modulus, $b$ is the magnitude of the Burgers vector, $\nu$ is Poisson’s ratio, $d_p$ is the particle diameter, and $\lambda$ is the interparticle spacing.

This equation shows that finer particles dispersed at higher density (smaller $\lambda$) provide greater strengthening effect.

High-Temperature Properties

The greatest advantage of ODS steel is its high-temperature strength surpassing RAFM steel.

Creep Strength Comparison

A comparison of 10,000-hour creep rupture strength at 700°C is shown below.

Material	Creep Rupture Strength (MPa)
F82H (RAFM)	~30
9Cr-ODS	~120
12Cr-ODS	~100

ODS steel exhibits approximately 3-4 times the creep strength of RAFM steel. This makes it possible to extend the upper operating temperature limit to 700°C or higher.

Operating Temperature Range

The operating temperature range of ODS steel is estimated as follows:

Material	Lower Limit Temperature	Upper Limit Temperature
9Cr-ODS	400°C	700°C
12Cr-ODS	400°C	750°C
14Cr-ODS	450°C	800°C

Challenges and Prospects

ODS steel, while having excellent high-temperature properties, faces the following challenges:

1. Manufacturing cost: Mechanical alloying is a time- and energy-intensive process, resulting in high manufacturing costs

2. Large-scale material production: Currently, manufacturing large monolithic components is difficult, requiring development of joining technologies

3. Anisotropy: Materials produced by hot extrusion exhibit anisotropy in mechanical properties due to grain morphology anisotropy

4. Joining technology: Since oxide particles coarsen and disappear upon melting, conventional fusion welding is not applicable. Development of solid-state joining technologies such as diffusion bonding and friction stir welding is underway

Vanadium Alloys

Vanadium alloys are candidate materials with higher operating temperatures and superior low-activation properties compared to RAFM steel. Combined with liquid lithium-cooled self-cooled blankets, they offer the possibility of achieving high power generation efficiency and high tritium breeding ratios.

Development of V-Cr-Ti Alloys

Development of vanadium alloys began in the 1980s at Argonne National Laboratory in the United States. After examining various alloy compositions, V-4Cr-4Ti (4 mass% Cr, 4 mass% Ti, balance V) was selected as the most promising composition.

History of Composition Optimization

In the composition optimization of vanadium alloys, the following factors were considered:

Role of chromium (Cr):

• Improved high-temperature strength through solid solution strengthening
• Reduced affinity for oxygen (suppression of oxygen embrittlement)
• Reduced workability and embrittlement above 5%

Role of titanium (Ti):

• Fixing of interstitial impurities such as oxygen, nitrogen, and carbon (getter effect)
• Grain refinement
• Coarsening of Ti-based precipitates above 4%

Through these studies, V-4Cr-4Ti was established as the optimal composition.

Major V-4Cr-4Ti Alloys

NIFS-HEAT (Japan)

NIFS-HEAT (National Institute for Fusion Science - Heat Exchanger And Tritium) is a vanadium alloy developed and manufactured primarily by the National Institute for Fusion Science.

Chemical composition (mass%):

Element	Composition
Cr	4.0
Ti	4.0
O	<200 ppm
N	<100 ppm
C	<100 ppm
V	Bal.

NIFS-HEAT is manufactured from approximately 200 kg ingots by vacuum arc remelting (VAR), and plates, bars, and tubes have been fabricated by rolling.

Mechanical properties (after annealing):

Property	Room Temperature	400°C	700°C
0.2% Yield Strength (MPa)	350	280	200
Tensile Strength (MPa)	420	350	250
Elongation (%)	30	25	25

US-DOE Alloys (USA)

Vanadium alloys manufactured with support from the U.S. Department of Energy (DOE) are managed by identification numbers such as Heat 832665. Irradiation testing and property evaluation are being conducted primarily at Argonne National Laboratory.

Physical and Chemical Properties

The physical properties of vanadium alloys are shown below.

Physical constants:

Property	Value
Density	6.1 g/cm³
Melting Point	1910°C
Thermal Conductivity (20°C)	31 W/(m·K)
Coefficient of Thermal Expansion (20-700°C)	10 × 10⁻⁶ /K
Elastic Modulus (20°C)	128 GPa

Vanadium alloys have the following excellent properties:

1. High-temperature strength: High-temperature strength surpassing RAFM steel allows extension of the upper operating temperature to 700°C

2. Compatibility with liquid lithium: Excellent chemical compatibility with liquid lithium makes them suitable for Li-cooled self-cooled blankets

3. Low neutron absorption cross-section: Small thermal neutron absorption cross-section provides excellent neutron economy

4. Excellent low-activation properties: Vanadium itself is a low-activation element, with relatively short induced radioactivity levels and half-lives

Effect of Impurities

The mechanical properties of vanadium alloys are significantly affected by interstitial impurities such as oxygen, nitrogen, and carbon.

Effect of Oxygen

Oxygen dissolves interstitially in vanadium, hardening and embrittling the material. The relationship between oxygen concentration and ductility is summarized as follows:

Oxygen Concentration (ppm)	Room Temperature Elongation (%)	DBTT (°C)
<200	>25	<-100
300-500	15-25	-50 to 0
>800	<10	>50

Controlling oxygen concentration below 200 ppm is important for maintaining good mechanical properties.

Impurity Control

The following measures are necessary to maintain impurities at low levels:

1. Use of high-purity raw materials: Increase purity of vanadium raw materials
2. Control of melting/manufacturing atmosphere: Melting and heat treatment in vacuum or high-purity argon atmosphere
3. Heat treatment environment management: Oxygen removal using getter materials (such as Ti foil)

Compatibility with Liquid Lithium Cooling

Vanadium alloys are the optimal structural material for self-cooled blanket concepts using liquid lithium as both coolant and tritium breeding material.

Advantages of Self-Cooled Blankets

1. High tritium breeding ratio: Tritium breeding ratios above 1.4 are achievable using pure lithium
2. High thermal efficiency: Efficient heat removal is possible due to excellent heat transfer characteristics of liquid metals
3. Simple structure: Structure is simplified because the coolant and breeding material are the same substance

MHD Pressure Drop Challenge

When liquid metal flows through a magnetic field, eddy currents are generated by electromagnetic induction, which interact with the magnetic field to produce Lorentz forces. These forces act against the flow, causing large pressure drops (MHD pressure drops).

The MHD pressure drop $\Delta p_{\text{MHD}}$ is roughly estimated as:

$$\Delta p_{\text{MHD}} \propto \sigma_L v B^2 L$$

Here, $\sigma_L$ is the electrical conductivity of the liquid metal, $v$ is the flow velocity, $B$ is the magnetic field strength, and $L$ is the channel length.

In the strong magnetic field environment of fusion reactors (several T), MHD pressure drops can reach tens of MPa, making development of insulating coatings (e.g., Er₂O₃, Y₂O₃, AlN) an important challenge.

Operating Temperature Range

The operating temperature range of vanadium alloys is estimated as follows:

Limiting Factor	Temperature
Lower limit temperature (irradiation embrittlement)	~400°C
Upper limit temperature (creep, impurity reactions)	~700°C

The upper limit temperature is approximately 150°C higher than RAFM steel, and improvement in power generation efficiency can be expected.

SiC/SiC Composite Materials

SiC/SiC composite materials (silicon carbide fiber-reinforced silicon carbide composites) are candidate materials enabling the highest operating temperatures. Operation above 1000°C is possible, and they are being developed as advanced materials for future fusion reactors aiming for power generation efficiency above 50%.

Properties of SiC Ceramics

Silicon carbide (SiC) is a covalently bonded ceramic with excellent high-temperature properties.

Basic properties of SiC:

Property	Value
Theoretical Density	3.21 g/cm³
Melting Point (decomposition point)	~2700°C
Thermal Conductivity (room temperature)	120-270 W/(m·K)
Coefficient of Thermal Expansion (room temperature-1000°C)	4-5 × 10⁻⁶ /K
Elastic Modulus	380-430 GPa
Flexural Strength	300-600 MPa

SiC is composed only of the low-activation elements Si and C at the atomic level, exhibiting excellent low-activation properties. Radioactivity decays to levels allowing hands-on maintenance approximately 100 years after irradiation.

Need for SiC/SiC Composites

Bulk SiC ceramics, despite their excellent high-temperature properties, have the following critical drawbacks as structural materials:

1. Brittleness: Ceramics have no plastic deformation capability and fail suddenly due to stress concentration
2. Low fracture toughness: Fracture toughness $K_{IC}$ is 2-3 MPa·m^0.5, approximately 1/10 that of metallic materials
3. Low reliability: Large scatter in strength makes structural design difficult

To solve these problems, SiC/SiC composites combining SiC fibers and SiC matrix have been developed.

Composite Structure

SiC/SiC composites consist of the following elements:

1. SiC fibers: Function as reinforcement and support loads
2. Interface layer: Exists between fibers and matrix, providing appropriate interfacial bonding
3. SiC matrix: Holds fibers and transfers loads

SiC Fibers

Third-generation SiC fibers with high crystallinity and stoichiometric composition are used for fusion reactor SiC/SiC composites.

Representative SiC fibers:

Fiber	Manufacturer	Diameter (μm)	Tensile Strength (GPa)	Elastic Modulus (GPa)
Hi-Nicalon Type S	Nippon Carbon	12	2.6	420
Tyranno SA3	Ube Industries	7.5	2.8	380
Sylramic-iBN	COI Ceramics	10	3.2	400

These fibers have near-stoichiometric composition and superior high-temperature stability and radiation resistance compared to conventional SiC fibers (Nicalon, Tyranno, etc.).

Interface Layer

The interface layer is a crucial element controlling the fracture behavior of SiC/SiC composites. With an appropriate interface layer, cracks propagate around fibers, enabling energy absorption through fiber pull-out.

The following interface layers are used for fusion reactor applications:

• Pyrolytic carbon (PyC): The most common interface layer material. However, there are challenges regarding stability under neutron irradiation
• Multilayer SiC/PyC: Interface layer with improved radiation resistance

Matrix Formation Methods

Various methods are used to form the SiC matrix.

Chemical vapor infiltration (CVI): A method that supplies gases such as methyltrichlorosilane (CH₃SiCl₃) to fiber preforms and deposits SiC through high-temperature decomposition. High-purity, high-crystallinity SiC matrix is obtained, but densification takes time and residual porosity tends to remain.

Polymer impregnation and pyrolysis (PIP): A method that impregnates SiC precursor polymers such as polycarbosilane and converts them to SiC through pyrolysis. Densification proceeds through repeated impregnation and pyrolysis cycles.

NITE method (Nano-Infiltration Transient Eutectic-phase method): A method developed by Kyoto University. A slurry containing SiC fine powder and sintering aids is impregnated into fiber preforms and densified by hot pressing. Higher density matrices are obtained compared to CVI.

Mechanical Properties

Typical mechanical properties of fusion reactor SiC/SiC composites manufactured by the NITE method are shown below.

Property	Room Temperature	1000°C
Flexural Strength (MPa)	300-400	280-350
Tensile Strength (MPa)	200-300	180-260
Fracture Toughness (MPa·m^0.5)	15-25	15-25
Elastic Modulus (GPa)	250-300	240-280

SiC/SiC composites exhibit significantly higher fracture toughness than monolithic SiC ceramics. This is due to energy absorption mechanisms such as fiber pull-out, crack deflection, and crack bridging.

Operating Temperature Range

The operating temperature range of SiC/SiC composites is the widest among candidate materials.

Limiting Factor	Temperature
Lower limit temperature (thermal conductivity decrease)	~800°C
Upper limit temperature (strength decrease, oxidation)	~1000°C or higher

However, significant decrease in thermal conductivity under neutron irradiation is a challenge, and the lower limit temperature may increase in the irradiation environment.

Hermeticity and Joining Technology

High hermeticity is required for structural materials to contain tritium breeding materials and coolants in fusion reactors. SiC/SiC composites are inherently porous, making hermeticity an important challenge.

Hermetic Layer Formation

Hermetic layers are formed by the following methods:

1. CVD-SiC layer: Dense SiC layer is formed on the composite surface by chemical vapor deposition (CVD)
2. Glass sealing: Surface pores are sealed with low-melting glass

Joining Technology

The following methods are being studied for joining SiC/SiC composites:

1. NITE joining: Solid-state joining using powder similar to the NITE method as bonding material
2. Reaction sintering joining: Joining utilizing SiC formation through reaction of Si and C
3. Brazing: Joining using high-temperature brazing materials (though this introduces dissimilar materials)

Combination with Helium Gas Cooling

SiC/SiC composites achieve maximum performance when combined with helium gas-cooled blankets.

Helium gas has the following advantages:

1. Chemical inertness: No reaction with SiC
2. High-temperature usability: No melting point, no phase change at high temperatures
3. Low induced radioactivity: Almost no activation by neutron irradiation

With helium gas outlet temperatures of 900°C or higher, power generation efficiency above 50% is achievable with gas turbine combined cycle systems. Furthermore, high-temperature heat can be used for process heat such as hydrogen production (thermochemical methods).

Material Damage from Neutron Irradiation

Understanding material damage from neutron irradiation is the most important issue in fusion reactor structural materials development. The 14.1 MeV neutrons produced by D-T fusion reactions have much higher energy than fission reactor neutrons (average energy approximately 2 MeV) and cause unprecedentedly severe irradiation damage to materials.

Displacement Damage

When neutrons collide with atoms in materials, atoms are displaced from their lattice positions and move to interstitial positions. This process is called “displacement.”

PKA and Defect Cascades

The atom first displaced by collision with a neutron is called the “Primary Knock-on Atom (PKA).” In elastic collisions with 14.1 MeV neutrons, PKAs can acquire energies of several hundred keV.

Such high-energy PKAs secondarily displace numerous surrounding atoms. This chain displacement process forms a region where numerous vacancies and interstitial atoms are locally generated. This is called a “displacement cascade.”

Definition of dpa

Dpa (displacements per atom) is widely used as a measure of irradiation dose.

$$\text{dpa} = \int_0^t \sigma_d \phi(E, t') dt'$$

Here, $\sigma_d$ is the displacement cross-section and $\phi(E, t')$ is the neutron flux.

1 dpa statistically means that all atoms in the material have been displaced once each. In reality, many atoms are displaced multiple times while some are not displaced at all. Also, many of the displaced vacancies and interstitial atoms immediately recombine and annihilate.

NRT Model

The Norgett-Robinson-Torrens (NRT) model is standardly used for estimating the number of displaced atoms.

$$\nu_{\text{NRT}}(E) = \frac{0.8 E_d}{2 E_{\text{th}}}$$

Here, $E_d$ is the damage energy (PKA energy minus electronic excitation contribution), and $E_{\text{th}}$ is the displacement threshold energy (typically around 40 eV).

Irradiation Dose in Fusion Reactor Environment

The irradiation dose that structural materials of commercial fusion reactors will receive during their operational lifetime is estimated as follows:

Component	Neutron Wall Load (MW/m²)	Annual dpa	dpa after 5 Years Operation
First Wall	1-2	10-20	50-100
Blanket Front	0.5-1	5-10	25-50
Blanket Rear	0.1-0.3	1-3	5-15

Transmutation Products

In addition to displacement damage, 14.1 MeV high-energy neutrons cause various nuclear reactions, changing the chemical composition of materials.

Helium Generation

Helium (He) is generated in materials through (n, α) reactions. Generated helium is insoluble in the lattice and aggregates to form bubbles.

Helium generation rates in major structural materials are shown below.

Material	He Generation Rate (appm/dpa)
RAFM Steel	10-12
Vanadium Alloys	3-5
SiC/SiC	70-100

Here, appm is atomic parts per million (number per million atoms).

In SiC/SiC composites, the helium generation rate is remarkably high because the nuclear reaction cross-sections of light elements (Si, C) are large.

Hydrogen Generation

Hydrogen (H) is generated through (n, p) reactions. Hydrogen can dissolve in the lattice but precipitates when supersaturated, causing hydrogen embrittlement.

Material	H Generation Rate (appm/dpa)
RAFM Steel	40-50
Vanadium Alloys	4-6
SiC/SiC	80-120

Solid Transmutation Products

As a result of (n, p) and (n, α) reactions, parent atoms are converted to different elements. In steel materials, conversions mainly to Mn, Cr, etc. occur.

Also, rhenium (Re) and osmium (Os) are generated from tungsten (W).

$$^{184}\text{W} + n \rightarrow ^{185}\text{W} \rightarrow ^{185}\text{Re}$$ $$^{186}\text{W} + n \rightarrow ^{187}\text{W} \rightarrow ^{187}\text{Re}$$ $$^{187}\text{Re} + n \rightarrow ^{188}\text{Re} \rightarrow ^{188}\text{Os}$$

Re and Os dissolve in W and may form brittle intermetallic compounds such as the σ phase.

Irradiation Hardening

Materials harden significantly due to neutron irradiation.

Mechanism

Irradiation hardening occurs because defect clusters generated by irradiation (dislocation loops, voids, precipitates, etc.) impede dislocation motion.

The increase in yield stress $\Delta\sigma_y$ due to irradiation is expressed by the obstacle strength model as:

$$\Delta\sigma_y = M \alpha G b \sqrt{N d}$$

Here, $M$ is the Taylor factor (approximately 3 for polycrystals), $\alpha$ is the obstacle strength factor (0.1-1), $G$ is the shear modulus, $b$ is the Burgers vector, $N$ is the obstacle density, and $d$ is the obstacle size.

Saturation of Irradiation Hardening

Irradiation hardening increases with irradiation dose and tends to saturate eventually. The saturation dose and saturation hardening amount depend on irradiation temperature and material.

For F82H, hardening saturates at approximately 10 dpa under irradiation at 300°C, with yield stress increasing by approximately 300 MPa.

Ductile-to-Brittle Transition Temperature (DBTT) Shift

Body-centered cubic metals (RAFM steel, vanadium alloys, etc.) have the property of transitioning from ductile to brittle below a certain temperature. This transition temperature is called the Ductile-to-Brittle Transition Temperature (DBTT).

Mechanism of DBTT Shift

Irradiation causes the DBTT to increase. This is because while irradiation hardening increases the yield stress, the cleavage fracture stress remains almost unchanged.

The DBTT can be defined as the temperature at which yield stress equals cleavage fracture stress. When yield stress increases, the temperature at which both become equal shifts to higher temperatures, raising the DBTT.

DBTT Shift Amount

The DBTT shift of RAFM steel is a function of irradiation temperature and dose.

Irradiation Temperature	Irradiation Dose (dpa)	DBTT Shift (°C)
300°C	5	150-200
300°C	10	200-250
400°C	10	100-150
500°C	10	50-100

Higher irradiation temperatures promote defect recovery, suppressing DBTT shift.

Swelling (Irradiation-Induced Volume Expansion)

The phenomenon of material volume expansion due to neutron irradiation is called swelling.

Mechanism

Swelling occurs when vacancies generated by irradiation aggregate to form voids. The thermodynamic stability of vacancies and voids depends on temperature, and swelling reaches maximum at a certain temperature range (swelling peak temperature).

The swelling peak temperature is approximately 0.3-0.5 times the melting point (homologous temperature). For RAFM steel (melting point approximately 1500°C), it is around 400-550°C, and for vanadium alloys (melting point approximately 1910°C), around 500-650°C.

Swelling Rate

The swelling rate $S$ is defined as the volume change rate.

$$S = \frac{\Delta V}{V_0} \times 100 \quad (\%)$$

RAFM steel exhibits superior swelling resistance compared to austenitic stainless steel. For F82H, swelling remains below 1% even after irradiation at 500°C to 100 dpa.

Helium Embrittlement

Helium generated by transmutation significantly degrades the high-temperature properties of materials.

Mechanism

Helium atoms are insoluble in the metal lattice and accumulate at grain boundaries and precipitate interfaces to form bubbles. These helium bubbles reduce high-temperature creep rupture life and cause grain boundary embrittlement.

Helium embrittlement becomes pronounced above 550°C. This is because helium diffusion becomes active at high temperatures, promoting helium accumulation at grain boundaries.

Helium Embrittlement Temperature

The relationship between helium concentration and embrittlement onset temperature is roughly summarized as follows:

He Concentration (appm)	Embrittlement Onset Temperature (°C)
<10	>600
10-100	550-600
100-500	500-550
>500	<500

In SiC/SiC composites, despite the high helium generation rate, helium bubbles disperse finely and significant embrittlement has not been observed.

Irradiation Creep

Creep deformation under irradiation (irradiation creep) proceeds by a mechanism different from thermal creep.

Mechanism

Irradiation creep occurs when point defects (vacancies, interstitial atoms) continuously generated by neutron irradiation move anisotropically under stress.

The irradiation creep strain rate $\dot{\varepsilon}_{\text{irr}}$ is proportional to stress and irradiation rate.

$$\dot{\varepsilon}_{\text{irr}} = B \sigma \dot{\phi}$$

Here, $B$ is the irradiation creep coefficient, $\sigma$ is the stress, and $\dot{\phi}$ is the irradiation rate (dpa/s).

Irradiation Creep Coefficient

Irradiation creep coefficients of major structural materials are shown below.

Material	Irradiation Creep Coefficient B (MPa⁻¹·dpa⁻¹)
RAFM Steel	(1-3) × 10⁻⁶
Vanadium Alloys	(2-5) × 10⁻⁶
SiC/SiC	(0.5-2) × 10⁻⁶

Unlike thermal creep, irradiation creep has small temperature dependence. Therefore, consideration is necessary even for deformation at low temperatures.

Irradiation Test Facilities

Development of fusion reactor materials requires irradiation testing that simulates the fusion environment. However, since no facility exists that steadily generates D-T fusion reactions, alternative irradiation facilities are used.

Irradiation in Fission Reactors

Currently, most material irradiation testing is performed in fission reactors.

Major Irradiation Facilities

Facility	Country	Reactor Type	Maximum Neutron Flux (n/cm²/s)	Maximum Dose Rate (dpa/year)
HFIR	USA	Test Reactor	2 × 10¹⁵	~20
ATR	USA	Test Reactor	1 × 10¹⁵	~10
BOR-60	Russia	Fast Reactor	3 × 10¹⁵	~20
JOYO	Japan	Fast Reactor	4 × 10¹⁵	~25

Limitations of Fission Reactor Irradiation

The neutron spectrum of fission reactors differs significantly from fusion reactors. There are particularly limitations in the following points:

1. Neutron energy: The average neutron energy of fission reactors is approximately 2 MeV, significantly lower than fusion neutrons (14.1 MeV)

2. He/dpa ratio: Helium generation rates in fission reactors are orders of magnitude lower than in fusion reactors. For RAFM steel, it is approximately 0.3 appm/dpa in fission reactors versus approximately 10 appm/dpa in fusion reactors

3. Damage structure: The structure of defect cascades from high-energy neutrons differs from that of low-energy neutrons

To bring the He/dpa ratio closer to fusion conditions, specimens with added ⁵⁹Ni or ¹⁰B may be used.

$$^{59}\text{Ni} + n \rightarrow ^{56}\text{Fe} + \alpha$$ $$^{10}\text{B} + n \rightarrow ^{7}\text{Li} + \alpha$$

These reactions generate additional helium, enabling simulation of He/dpa ratios close to fusion conditions.

Ion Beam Irradiation

Methods irradiating materials with charged particle beams generated by accelerators are also utilized.

Advantages

1. High irradiation doses can be achieved in short times (dpa rate is 100-1000 times that of fission reactors)
2. Irradiation conditions (temperature, damage rate, helium injection rate) can be precisely controlled
3. Almost no residual radioactivity is generated, making post-irradiation handling easy

Limitations

1. Damage region is limited to near the surface (several μm to tens of μm)
2. Damage rate is significantly higher than fusion environment, potentially resulting in different damage structures
3. Evaluation of bulk properties (tensile, creep, etc.) is difficult

IFMIF-DONES

IFMIF-DONES (International Fusion Materials Irradiation Facility - DEMO-Oriented Neutron Source) is a high-flux neutron source planned for fusion reactor materials development.

Facility Overview

IFMIF-DONES irradiates a liquid lithium target with 40 MeV deuteron beams generated from a deuteron accelerator, producing high-energy neutrons through spallation reactions.

$$d + \text{Li} \rightarrow n + \cdots$$

The energy spectrum of generated neutrons has a peak close to D-T fusion neutrons (14.1 MeV).

Main Parameters

Parameter	Value
Deuteron Beam Energy	40 MeV
Beam Current	125 mA
Beam Power	5 MW
Maximum Neutron Flux	10¹⁸ n/m²/s
High-Flux Irradiation Volume	500 cm³
Maximum Irradiation Dose	~50 dpa/year
He/dpa Ratio	Equivalent to fusion environment

Development Status

IFMIF-DONES is being developed through Japan-Europe international cooperation. Construction is planned in Granada, Spain, with operation targeted for the 2030s.

As a preliminary facility, the IFMIF/EVEDA (Engineering Validation and Engineering Design Activities) prototype accelerator has been constructed and operated at Rokkasho in Japan.

Materials Database and Design Codes

Materials property databases and design codes are essential for fusion reactor design and safety evaluation.

Materials Databases

JAEA Materials Database

The Japan Atomic Energy Agency (JAEA) has established a property database for fusion reactor materials.

Included data:

• Chemical composition and heat treatment history
• Mechanical properties (tensile, hardness, impact, creep, fatigue)
• Physical properties (thermal conductivity, coefficient of thermal expansion, specific heat, electrical resistivity)
• Post-irradiation properties
• Weld zone properties

MatDB (Europe)

Fusion for Energy operates the integrated materials database MatDB for fusion reactor materials. Data on European-developed materials, centered on EUROFER97, are included.

Materials Handbooks

Materials handbooks systematically compiling properties necessary for design have been compiled for each material.

Material	Handbook
F82H	F82H Handbook (JAEA)
EUROFER97	EUROFER97 Handbook (KIT)
V-4Cr-4Ti	V-4Cr-4Ti Alloy Handbook (ANL)
SiC/SiC	SiC/SiC CMC Handbook (ORNL)

Design Codes and Standards

Design of fusion reactors requires design codes that extend existing standards to the fusion environment.

SDC-IC (ITER Structural Design Criteria)

SDC-IC (Structural Design Criteria for In-Vessel Components) is a structural design standard established for ITER vacuum vessel internal components.

SDC-IC is based on the French nuclear equipment standard RCC-MR with the following fusion-specific requirements added:

1. Electromagnetic force loads (plasma disruptions, etc.)
2. Material property changes due to irradiation
3. High helium concentration environment

RCC-MRx

RCC-MRx is a high-temperature structural design standard established by the French Alternative Energies and Atomic Energy Commission (CEA), and its application to fusion reactor design is being considered.

The design allowable stress $S_m$ for RAFM steel is defined as the minimum of the following values:

$$S_m = \min\left(\frac{R_m}{3}, \frac{R_{p0.2}}{1.5}, \frac{R_{mt}}{3}, \frac{R_{pt}}{1.5}\right)$$

Here, $R_m$ is the tensile strength, $R_{p0.2}$ is the 0.2% yield strength, and subscript t indicates values at operating temperature.

Reflecting Irradiation Data in Design

Knockdown factors for properties are established to reflect irradiation-induced material property changes in design.

Tensile Properties

Post-irradiation 0.2% yield strength $R_{p0.2}^{\text{irr}}$ and tensile strength $R_m^{\text{irr}}$ are specified as functions of irradiation dose.

Example (F82H, 300°C irradiation):

Irradiation Dose (dpa)	Change in 0.2% Yield Strength	Change in Tensile Strength	Change in Elongation
0	Baseline	Baseline	Baseline
5	+200 MPa	+150 MPa	-40%
10	+280 MPa	+200 MPa	-50%
30	+300 MPa (saturated)	+220 MPa (saturated)	-60%

Fracture Toughness

Fracture toughness evaluation considering DBTT shift is necessary. Based on the Master Curve method, it is evaluated as a change in reference temperature $T_0$.

$$T_0^{\text{irr}} = T_0^{\text{unirr}} + \Delta T_0$$

The DBTT shift $\Delta\text{DBTT}$ and $\Delta T_0$ are known to be approximately equal.

Future Prospects and Materials Development Strategy

Short-Term, Medium-Term, and Long-Term Strategy

Fusion reactor materials development is being pursued through a phased strategy as follows:

Period	Target Reactor	Main Structural Material	Upper Temperature Limit	Power Generation Efficiency Target
Short-term (~2040)	ITER/DEMO	RAFM Steel	550°C	35-40%
Medium-term (2040-2060)	First Generation Commercial Reactor	ODS Steel	700°C	40-45%
Long-term (2060~)	Advanced Commercial Reactor	SiC/SiC, V Alloys	1000°C or higher	50% or higher

Advancement of RAFM Steel

In the short term, advancement of existing RAFM steel is important.

1. Optimization of manufacturing technology: Stable production of large forgings, thin sheets, and tubes
2. Advancement of welding technology: Establishment of remote welding and repair welding techniques
3. Expansion of irradiation database: Acquisition of high-dose, long-term data
4. Development of design standards: Improvement and international standardization of SDC-IC

Practical Application of ODS Steel

In the medium term, practical application of ODS steel is the goal.

1. Efficiency of manufacturing process: Shortening MA time, continuous manufacturing technology
2. Large-scale material production: Scale-up of HIP equipment, advancement of hot extrusion
3. Establishment of joining technology: Practical application of friction stir welding and diffusion bonding
4. Demonstration of long-term stability: Evaluation of high-temperature long-term exposure and post-irradiation properties

Development of Advanced Materials

In the long term, practical application of SiC/SiC composites and vanadium alloys is the goal.

SiC/SiC composites:

1. Manufacturing technology for large and complex-shaped components
2. Establishment of hermeticity and joining technology
3. Elucidation of irradiation damage mechanisms and improvement of resistance

Vanadium alloys:

1. Advancement of impurity control technology
2. Development of MHD insulating coatings
3. Demonstration of long-term compatibility with liquid lithium

Global Cooperation

Fusion reactor materials development has a scale and complexity that cannot be completed by a single country. International cooperation in development, exemplified by IFMIF-DONES, is essential.

Major international cooperation frameworks:

1. ITER international cooperation (materials database, design standards)
2. BA activities (Japan-Europe, IFMIF-DONES)
3. IEA fusion materials cooperation (data sharing, round-robin tests)

Related Topics

• Materials and Engineering: Overview - Overview of fusion reactor materials
• Plasma-Facing Materials - First wall and divertor materials
• ITER Project - Test Blanket Module plans
• DEMO Design - Structural material requirements for demonstration power reactor