Blanket

The blanket is a component installed to surround the plasma in a fusion reactor, serving three critical functions: producing tritium fuel, converting fusion energy into heat, and shielding against neutrons. For fusion reactors using the D-T reaction, the blanket is one of the most important and technically challenging components, and its design determines the overall performance, economics, and safety of the reactor.

The term “blanket” derives from the fact that it “wraps around the plasma like a blanket.” Most of the high-energy neutrons (14.1 MeV) generated by fusion reactions are slowed and absorbed in this blanket, and their kinetic energy is converted to heat. Simultaneously, nuclear reactions between lithium in the blanket and neutrons produce tritium, enabling fuel self-sufficiency for the fusion reactor.

Fundamental Roles of the Blanket

Three Primary Functions

The blanket must simultaneously fulfill three essential functions in a fusion reactor. These functions are interrelated, and design optimization must consider trade-offs among them.

Tritium Breeding Function

In D-T fusion reactions, one tritium atom is consumed per reaction. Tritium exists in negligible quantities in nature (only trace amounts are produced by cosmic ray reactions with the atmosphere), and with a relatively short half-life of 12.32 years, continuous external supply is impractical. Therefore, the fusion reactor must have the capability to produce its own tritium.

By loading lithium-containing materials in the blanket, tritium is generated through nuclear reactions between neutrons produced by fusion reactions. This “tritium breeding” function allows the fusion reactor to produce more tritium than it consumes, achieving fuel self-sufficiency.

Heat Conversion Function

Approximately 80% of the energy released in D-T fusion reactions is carried by neutrons with kinetic energy of 14.1 MeV. Converting this neutron energy efficiently to heat for power generation is a crucial role of the blanket.

{}^{2}\text{D} + {}^{3}\text{T} \rightarrow {}^{4}\text{He}\ (3.5\ \text{MeV}) + n\ (14.1\ \text{MeV})

Of the total 17.6 MeV energy, the 3.5 MeV carried by alpha particles (helium-4) is used for plasma self-heating, while the 14.1 MeV carried by neutrons is recovered in the blanket. Neutrons are slowed through elastic and inelastic scattering with nuclei within the blanket, ultimately transferring thermal energy to the coolant.

Shielding Function

High-energy neutrons and accompanying gamma rays cause severe damage to equipment and structures outside the blanket. Superconducting coils are particularly vulnerable to neutron irradiation and require strict dose limits. Sufficient shielding performance is also needed to reduce worker exposure during maintenance operations.

The blanket protects external equipment by slowing, absorbing neutrons, and shielding gamma rays. Inadequate shielding performance leads to serious problems such as reduced superconducting coil lifetime, increased vacuum vessel activation, and difficulties in remote maintenance.

Detailed Performance Requirements

The specific performance requirements for blankets are summarized below.

Function Category	Requirement	Design Target	Notes
Tritium Production	Tritium Breeding Ratio (TBR)	$\geq 1.05$	Margin for losses
Tritium Production	Tritium Recovery Rate	$> 99\%$	Extraction efficiency from breeder
Heat Extraction	Energy Multiplication Factor	1.1 - 1.5	Additional energy from nuclear reactions
Heat Extraction	Coolant Outlet Temperature	300 - 700 C	Directly affects power generation efficiency
Shielding	Superconducting Coil Neutron Flux	$< 10^{9}$ n/cm $^2$ /s	To ensure coil lifetime
Shielding	Coil Nuclear Heating	$< 1$ mW/cm $^3$	To reduce cryogenic load
Structural Integrity	Neutron Wall Load Tolerance	1 - 2 MW/m $^2$	DEMO conditions
Structural Integrity	Design Lifetime	$> 5$ years	To reduce replacement frequency
Maintainability	Module Removal Time	$< 1$ week/module	By remote operation
Safety	Tritium Containment	Multiple barriers	Below public exposure limits

Physics of Tritium Breeding

Tritium as Fusion Fuel

Tritium (hydrogen-3, ${}^{3}\text{H}$ or T) is a radioactive isotope of hydrogen consisting of one proton and two neutrons. It undergoes beta decay with a half-life of 12.32 years, transforming into helium-3.

{}^{3}\text{T} \rightarrow {}^{3}\text{He} + e^{-} + \bar{\nu}_e + 18.6\ \text{keV}

The amount of tritium on Earth is extremely small; in nature, only about 4 kg is produced annually through spallation reactions between cosmic rays and atmospheric nuclei. In contrast, a 1 GW fusion power plant is estimated to consume approximately 56 kg of tritium per year.

\dot{m}_T = \frac{P_{\text{fusion}}}{E_{\text{fusion}}} \times m_T \times t

Here, $P_{\text{fusion}}$ is the fusion power output, $E_{\text{fusion}} = 17.6$ MeV is the energy per reaction, and $m_T$ is the mass of a tritium atom. For 1 GW of fusion power, the tritium consumption rate is approximately $1.8 \times 10^{-6}$ g per second, reaching about 56 kg annually.

Therefore, “tritium breeding” that produces more tritium than consumed within the reactor is essential for sustained fusion reactor operation.

Lithium-6 Reaction

The reaction between lithium-6 ( ${}^{6}\text{Li}$ ) and neutrons is the primary pathway for tritium breeding.

{}^{6}\text{Li} + n \rightarrow {}^{4}\text{He}\ (2.05\ \text{MeV}) + T\ (2.73\ \text{MeV})

Characteristics of this reaction:

Q-value: +4.78 MeV (exothermic reaction)
Reaction threshold: None (can occur with thermal neutrons)
Cross-section: Approximately 940 barns at thermal energy (0.025 eV), follows 1/v law

The energy dependence of the reaction cross-section is approximated by:

\sigma(E) \approx \sigma_0 \sqrt{\frac{E_0}{E}} \quad (\text{low energy region})

Here, $\sigma_0 = 940$ barns and $E_0 = 0.025$ eV.

Due to this 1/v law, ${}^{6}\text{Li}$ has a higher reaction probability with low-energy neutrons (thermal neutrons). Therefore, it is efficient to first slow down fusion neutrons (14.1 MeV) before reacting them with ${}^{6}\text{Li}$ .

The natural abundance of ${}^{6}\text{Li}$ in lithium is 7.59%, and enriched lithium-6 materials may be used to enhance tritium breeding efficiency.

Lithium-7 Reaction

The reaction between lithium-7 ( ${}^{7}\text{Li}$ ) and neutrons proceeds only with high-energy neutrons.

{}^{7}\text{Li} + n \rightarrow {}^{4}\text{He} + T + n'

Characteristics of this reaction:

Q-value: -2.47 MeV (endothermic reaction)
Reaction threshold: 2.47 MeV (effectively about 4 MeV or higher)
Cross-section: Approximately 0.3 barns at 10 MeV

Being an endothermic reaction, the incident neutron energy must exceed the threshold. Fusion neutrons at 14.1 MeV satisfy this condition, and a neutron remains after the reaction. This “surplus neutron” effectively multiplies neutrons.

The natural abundance of ${}^{7}\text{Li}$ is 92.41%, comprising most of lithium, but its contribution to tritium production is limited due to its smaller cross-section compared to ${}^{6}\text{Li}$ . However, the neutron multiplication effect by high-energy neutrons cannot be ignored.

Kinematics of Breeding Reactions

Let us examine the kinematics of the ${}^{6}\text{Li}(n,\alpha)T$ reaction in detail. Consider the reaction in the center-of-mass frame.

With incident neutron energy $E_n$ and reaction Q-value $Q = 4.78$ MeV, the total energy in the center-of-mass frame is:

E_{\text{cm}} = E_n \times \frac{m_{\text{Li}}}{m_n + m_{\text{Li}}} + Q

The energy distribution of product particles (alpha particle and triton) follows from momentum conservation:

E_\alpha = \frac{m_T}{m_\alpha + m_T} \times (E_{\text{cm}} - Q') \approx 2.05\ \text{MeV}

E_T = \frac{m_\alpha}{m_\alpha + m_T} \times (E_{\text{cm}} - Q') \approx 2.73\ \text{MeV}

For thermal neutron incidence ( $E_n \approx 0.025$ eV), the triton produced has an energy of approximately 2.73 MeV, which is rapidly slowed down within the blanket and converted to thermal energy.

Principles of Neutron Multiplication

Neutron Balance

To achieve TBR $> 1$ in a fusion reactor, detailed examination of neutron balance is necessary. One neutron is generated per D-T reaction, but the effective number of neutrons is reduced by the following factors:

Neutron absorption by structural materials
Neutron leakage through blanket openings (ports, plasma heating devices, etc.)
Neutron capture reactions (parasitic absorption)
Leakage to vacuum vessel and shielding

To compensate for these losses, neutron multiplier materials that undergo $(n,2n)$ reactions are used. The neutron balance is expressed as:

\text{TBR} = \int_V \left[ \sigma_6(E) N_6 + \sigma_7(E) N_7 \right] \phi(E) \, dE \, dV

Here, $\sigma_6$ and $\sigma_7$ are the reaction cross-sections for ${}^{6}\text{Li}$ and ${}^{7}\text{Li}$ respectively, $N_6$ and $N_7$ are nuclear densities, and $\phi(E)$ is the neutron flux spectrum.

Neutron Multiplication by Beryllium

Beryllium ( ${}^{9}\text{Be}$ ) is the most efficient neutron multiplier.

{}^{9}\text{Be} + n \rightarrow 2{}^{4}\text{He} + 2n - 1.57\ \text{MeV}

Or more precisely:

{}^{9}\text{Be} + n \rightarrow {}^{8}\text{Be}^* + n' \rightarrow 2{}^{4}\text{He} + 2n

Characteristics of the beryllium $(n,2n)$ reaction:

Property	Value
Reaction threshold	1.85 MeV
Maximum cross-section	Approximately 0.6 barns (3-4 MeV)
Average multiplication factor	Approximately 1.8 (for 14 MeV neutron incidence)
Melting point	1287 C
Density	1.85 g/cm $^3$

Advantages of beryllium:

Low reaction threshold (1.85 MeV) enables multiplication even with moderated neutrons
Being a light element, it also functions as a neutron moderator
High melting point allows high-temperature operation

Challenges with beryllium:

Limited resources (global confirmed reserves of approximately 100,000 tons)
Toxic dust (risk of chronic beryllium disease)
Helium embrittlement and swelling due to irradiation

Neutron Multiplication by Lead

Lead (Pb) is also used as a neutron multiplier that undergoes $(n,2n)$ reactions. Natural lead is a mixture of four stable isotopes ( ${}^{204}\text{Pb}$ , ${}^{206}\text{Pb}$ , ${}^{207}\text{Pb}$ , ${}^{208}\text{Pb}$ ).

{}^{A}\text{Pb} + n \rightarrow {}^{A-1}\text{Pb} + 2n - Q

$(n,2n)$ reaction thresholds and cross-sections for each isotope:

Isotope	Abundance	Reaction Threshold	Cross-section (14 MeV)
${}^{204}\text{Pb}$	1.4%	8.4 MeV	Approximately 2.1 barns
${}^{206}\text{Pb}$	24.1%	8.1 MeV	Approximately 2.2 barns
${}^{207}\text{Pb}$	22.1%	6.7 MeV	Approximately 2.3 barns
${}^{208}\text{Pb}$	52.4%	7.4 MeV	Approximately 2.2 barns

Advantages of lead:

Abundant and inexpensive resources
Low melting point (327 C) facilitates use in liquid state
Lithium-lead alloy (Li-Pb) can serve as both breeder and multiplier

Challenges with lead:

High reaction threshold (6.7-8.4 MeV), multiplication only with high-energy neutrons
Being a high-Z element, has limited neutron moderation effect
${}^{210}\text{Po}$ production from activation (from ${}^{209}\text{Bi}$ impurities)

Optimization of Neutron Multiplier Placement

The placement of neutron multipliers must be optimized to maximize TBR. The basic principles are:

Position multipliers close to the plasma (to utilize high-energy neutrons)
Position breeder material (lithium) behind the multiplier (to utilize moderated neutrons)
Alternate breeder and multiplier placement (multi-layer structure)

Optimization by one-dimensional transport calculation yields a typical arrangement:

\text{Plasma} \rightarrow \text{First Wall} \rightarrow \text{Be Layer} \rightarrow \text{Li Layer} \rightarrow \text{Be Layer} \rightarrow \text{Li Layer} \rightarrow \cdots \rightarrow \text{Shielding Layer}

Such multi-layer structures can achieve TBR = 1.1 - 1.4.

Tritium Breeding Ratio (TBR) Design

Definition and Calculation of TBR

The Tritium Breeding Ratio (TBR) is a crucial indicator that quantitatively represents the tritium self-sufficiency capability of a fusion reactor.

\text{TBR} = \frac{\text{Tritium production rate per unit time}}{\text{Tritium consumption rate per unit time}}

More precisely, it is obtained by integrating the tritium production rate from ${}^{6}\text{Li}$ and ${}^{7}\text{Li}$ reactions within the blanket.

\text{TBR} = \frac{1}{S_n} \int_0^\infty \int_V \left[ \sigma_6(E) N_6(\mathbf{r}) + \sigma_7(E) N_7(\mathbf{r}) \right] \phi(\mathbf{r}, E) \, dV \, dE

Here, $S_n$ is the fusion neutron source strength (neutrons/second), and $\phi(\mathbf{r}, E)$ is the neutron flux at position $\mathbf{r}$ and energy $E$ .

Basis for TBR Design Targets

While TBR $\geq 1$ is necessary for a fusion reactor to be self-sufficient in tritium, actual designs require margins accounting for the following factors.

Tritium Loss Factors

Loss due to radioactive decay

\frac{dN_T}{dt} = -\lambda N_T, \quad \lambda = \frac{\ln 2}{t_{1/2}} = \frac{0.693}{12.32\ \text{years}}

Approximately 5.5% of tritium is lost to decay annually.

Losses in fuel processing systems

Small losses occur at each stage of tritium recovery, purification, and fuel supply. Typical loss rates are:

Recovery efficiency from breeder: 99% (1% loss)
Processing system loss: 0.1-0.5%
Fuel supply system loss: 0.1%

Tritium inventory

“Inventory” tritium present in fuel processing systems, inside the vacuum vessel, and in structural materials is not available during reactor operation. Tritium accumulation in plasma-facing materials (tungsten, beryllium) is particularly problematic.

Fuel supply for future reactors

During the deployment phase of fusion power, tritium must be supplied for initial fuel loading of new reactors. Approximately 1-2 kg of tritium is needed for initial loading of a 1 GW reactor.

Estimation of Required TBR

The required TBR considering these factors is estimated by:

\text{TBR}_{\text{required}} = 1 + \epsilon_{\text{decay}} + \epsilon_{\text{process}} + \epsilon_{\text{margin}}

Typical values are:

$\epsilon_{\text{decay}} \approx 0.01$ (decay loss component)
$\epsilon_{\text{process}} \approx 0.02$ (processing loss component)
$\epsilon_{\text{margin}} \approx 0.02$ (design margin)

Therefore, TBR $\geq 1.05$ is the design target. More conservative targets of TBR $\geq 1.10$ may be adopted for DEMO reactor designs.

Design Parameters Affecting TBR

The major design parameters determining TBR are shown below.

Parameter	Effect on TBR	Optimization Direction
Blanket coverage	Higher coverage increases TBR	Minimize port area
Breeder lithium density	Higher density increases TBR	Use Li $_2$ O, liquid Li
${}^{6}\text{Li}$ enrichment	Optimization at moderate enrichment	30-90% (concept dependent)
Neutron multiplier amount	Optimal amount maximizes TBR	Be volume fraction 60-80%
Structural material volume fraction	Lower fraction increases TBR	Thin walls, low-absorption materials
Blanket thickness	Thicker increases TBR	50-80 cm (concept dependent)

TBR Calculation Methods

Monte Carlo neutron transport calculations are used for precise TBR calculations. Representative calculation codes include MCNP, Serpent, and TRIPOLI.

Calculation procedure:

Construction of 3D geometric model (conversion from CAD data)
Setting material compositions and nuclear cross-section libraries
Setting neutron source distribution (toroidal plasma distribution)
Execution of neutron transport calculation
Integration of ${}^{6}\text{Li}(n,\alpha)T$ and ${}^{7}\text{Li}(n,n'\alpha)T$ reaction rates
Statistical error evaluation

For ITER TBM design calculations, $10^8$ - $10^9$ neutron histories are tracked to keep TBR statistical error below 1%.

Energy Multiplication and Heat Conversion

Physics of Energy Multiplication

In the blanket, not only is the kinetic energy of fusion neutrons converted to heat, but energy is also generated or absorbed by various nuclear reactions. The energy multiplication factor $M$ represents the overall energy balance.

M = \frac{E_{\text{total}}}{E_n} = \frac{E_n + E_{\text{reaction}}}{E_n}

Here, $E_n = 14.1$ MeV is the fusion neutron energy, and $E_{\text{reaction}}$ is the net energy generation (exothermic reaction) or absorption (endothermic reaction) from nuclear reactions.

Major reactions and energy balance:

Reaction	Energy Balance	Typical Contribution
${}^{6}\text{Li}(n,\alpha)T$	+4.78 MeV	+0.3 - 0.4
${}^{7}\text{Li}(n,n'\alpha)T$	-2.47 MeV	-0.02
${}^{9}\text{Be}(n,2n)$	-1.57 MeV	-0.05 - 0.1
Structural material $(n,\gamma)$ reactions	+several MeV	+0.1 - 0.2
Neutron elastic scattering	0 (kinetic energy transfer only)	-

Typical solid breeder blankets yield $M \approx 1.15$ - 1.25, while liquid metal blankets yield $M \approx 1.10$ - 1.20.

Thermal Power Calculation

The blanket thermal power of a fusion reactor is calculated as:

P_{\text{blanket}} = P_{\text{fusion}} \times \frac{E_n}{E_{\text{fusion}}} \times M = P_{\text{fusion}} \times \frac{14.1}{17.6} \times M

For 1 GW fusion power output with $M = 1.2$ :

P_{\text{blanket}} = 1000\ \text{MW} \times 0.80 \times 1.2 = 960\ \text{MW}

Power generation is performed from the total thermal output including this plus heat recovery at the divertor (a portion of alpha particle energy).

Power Generation Efficiency

Higher blanket coolant temperatures improve power generation efficiency (Carnot efficiency constraint). Thermodynamic efficiency is expressed as:

\eta_{\text{Carnot}} = 1 - \frac{T_{\text{cold}}}{T_{\text{hot}}}

Actual power generation efficiency, considering turbine and generator efficiencies:

\eta_{\text{plant}} = \eta_{\text{Carnot}} \times \eta_{\text{turbine}} \times \eta_{\text{generator}} \times (1 - \epsilon_{\text{aux}})

Here, $\eta_{\text{turbine}} \approx 0.85$ - 0.90, $\eta_{\text{generator}} \approx 0.98$ , and $\epsilon_{\text{aux}}$ is the auxiliary power ratio.

Cooling Method	Outlet Temperature	Theoretical Efficiency	Effective Efficiency
Pressurized water	325 C	40%	33%
Helium	500 C	52%	42%
Liquid Li	600 C	55%	45%
Li-Pb	500 C	52%	40%

Cooling Methods

Helium Cooling

Helium (He) is chemically inert and has minimal reactions with neutrons, making it an excellent coolant for fusion reactors.

Characteristics and Advantages

Property	Value/Characteristic
Chemical stability	Completely inert, no corrosion
Neutron absorption	Extremely small (minimal effect on TBR)
Induced radioactivity	None (does not activate)
Phase change	None (single-phase flow)
Operating pressure	8 - 10 MPa (typical)
Operating temperature	350 - 550 C (outlet)

Thermal-Hydraulic Characteristics

Helium heat transfer depends on Reynolds and Prandtl numbers. Heat transfer coefficient in the turbulent regime is estimated by the Dittus-Boelter correlation.

\text{Nu} = 0.023 \times \text{Re}^{0.8} \times \text{Pr}^{0.4}

h = \frac{\text{Nu} \times k}{D_h}

Due to helium’s low density and low heat transfer coefficient, high flow velocities (50-100 m/s) are required for adequate heat removal, resulting in large pressure drops.

\Delta P = f \times \frac{L}{D_h} \times \frac{\rho v^2}{2}

Here, $f$ is the friction factor (Blasius equation), $L$ is the flow path length, and $D_h$ is the hydraulic diameter.

Design Challenges

Structural loading from high-pressure operation (8-10 MPa)
Coolant tube surface temperature rise due to low heat transfer coefficient
Increased pumping power (auxiliary power) due to high flow rates
Larger piping and equipment

Water Cooling

Water (pressurized water) is a reliable coolant with extensive experience in nuclear power generation.

Characteristics and Advantages

Property	Value/Characteristic
Heat transfer coefficient	High (approximately 10 times that of helium)
Technology maturity	Accumulated LWR technology
Operating pressure	15 - 15.5 MPa
Operating temperature	280 - 325 C (inlet/outlet)
Phase change	None (subcooled water)

Design Challenges

Low coolant temperature limits power generation efficiency
TBR reduction due to neutron moderation effect (considered in design)
Hydrogen generation risk during loss of coolant (structural material reaction)
Tritium dissolution and permeation into water

In water-cooled blankets, tritium concentration management in coolant water is an important safety issue. Tritium water (HTO) concentration in primary coolant is limited considering biological effects.

Liquid Metal Cooling

This method uses liquid lithium or lithium-lead alloy (Li-Pb) as coolant. These can also serve as breeder material, enabling simplified blanket structures.

Liquid Lithium

Property	Value/Characteristic
Melting point	180.5 C
Boiling point	1342 C
Li density	Maximum (pure lithium)
Thermal conductivity	High (approximately 50 W/m·K)
Electrical conductivity	High (significant MHD effect)

Main challenges with liquid lithium are chemical reactivity (violent reactions with water and air) and MHD effects.

Lithium-Lead Alloy (Li-Pb)

The eutectic composition Li $_{17}$ Pb $_{83}$ (17 at% Li) is primarily considered.

Property	Value/Characteristic
Melting point	235 C (eutectic)
Li density	Approximately 1/5 of pure Li
Neutron multiplication	$(n,2n)$ reaction by Pb
Vapor pressure	Lower than pure Li
MHD effect	Smaller than pure Li

Advantages of Li-Pb include serving as both breeder and neutron multiplier, and lower chemical reactivity than pure lithium.

Magnetohydrodynamic (MHD) Effects

When liquid metals flow in strong magnetic fields, magnetohydrodynamic (MHD) effects cause large pressure drops. This is an important challenge unique to fusion reactors.

When conductive fluid flows across magnetic field lines, induced currents are generated, and Lorentz forces (braking forces) arise from their interaction with the magnetic field.

\mathbf{F} = \mathbf{j} \times \mathbf{B}

\mathbf{j} = \sigma (\mathbf{E} + \mathbf{v} \times \mathbf{B})

MHD pressure drop is estimated by:

\Delta P_{\text{MHD}} = c \sigma v B^2 L

Here, $c$ is a coefficient depending on channel geometry, $\sigma$ is the fluid electrical conductivity, $v$ is the flow velocity, $B$ is the magnetic field strength, and $L$ is the channel length.

Flowing liquid lithium at 1 m/s in a 10 T magnetic field can result in MHD pressure drops of several MPa, more than 100 times the normal fluid friction losses.

MHD Countermeasures

Insulating coatings on channel walls (alumina, aluminum nitride, etc.)
Flow channel geometry optimization (flow parallel to magnetic field)
Use of low electrical conductivity materials (SiC/SiC composites)
Low flow velocity design (challenge to balance with heat transfer)

Blanket Concepts

HCPB (Helium Cooled Pebble Bed)

HCPB (Helium Cooled Pebble Bed) is a solid breeder blanket concept developed in Europe.

Structure and Materials

Component	Material	Specification
Structural material	EUROFER97 (reduced activation ferritic/martensitic steel)	Thickness 4-6 mm
Breeder	Li $_4$ SiO $_4$ pebbles	Diameter 0.25-0.63 mm
Multiplier	Be pebbles	Diameter 1 mm
Coolant	He gas	8 MPa, 300-500 C
Tritium purge gas	He + 0.1% H $_2$	0.1 MPa

Operating Principle

Fusion neutrons enter the blanket
Neutron multiplication in Be pebble layers
Tritium production in Li $_4$ SiO $_4$ pebble layers
Heat recovery through helium cooling tubes
Tritium recovery from pebbles by purge gas

Design Parameters

TBR: Approximately 1.15 (considering port openings)
Energy multiplication factor $M$ : Approximately 1.20
Coolant outlet temperature: 500 C
Surface heat flux tolerance: 0.5 MW/m $^2$
Neutron wall load tolerance: 2 MW/m $^2$

HCLL (Helium Cooled Lithium Lead)

HCLL (Helium Cooled Lithium Lead) is a liquid breeder blanket concept developed in Europe.

Structure and Materials

Component	Material	Specification
Structural material	EUROFER97	Thickness 4-6 mm
Breeder/Multiplier	Li $_{17}$ Pb $_{83}$	Liquid state
Coolant	He gas	8 MPa, 300-500 C
Insulating coating	Al $_2$ O $_3$	MHD countermeasure

Operating Principle

Fusion neutrons enter the blanket
Neutron multiplication by Pb in Li-Pb
Tritium production by Li in Li-Pb
Circulating Li-Pb to recover tritium
Heat recovery through helium cooling tubes

Design Parameters

TBR: Approximately 1.10 (considering port openings)
Energy multiplication factor $M$ : Approximately 1.15
Coolant outlet temperature: 500 C
Li-Pb outlet temperature: 450 C

HCLL features a relatively simple internal blanket structure as Li-Pb serves as both breeder and multiplier. However, MHD effect countermeasures are a significant challenge.

WCLL (Water Cooled Lithium Lead)

WCLL (Water Cooled Lithium Lead) is a liquid breeder blanket concept developed in Europe using pressurized water as coolant.

Structure and Materials

Component	Material	Specification
Structural material	EUROFER97	Thickness 4-6 mm
Breeder/Multiplier	Li $_{17}$ Pb $_{83}$	Liquid state
Coolant	Pressurized water	15.5 MPa, 295-328 C

Design Parameters

TBR: Approximately 1.10
Energy multiplication factor $M$ : Approximately 1.12
Cooling water outlet temperature: 328 C
Li-Pb outlet temperature: 328 C

WCLL has the advantage of utilizing mature LWR technology. However, power generation efficiency is limited by the low coolant temperature. Additionally, since structural material separates Li-Pb from water, design must consider Li-water reaction risks in case of cooling tube failure.

WCCB (Water Cooled Ceramic Breeder)

WCCB (Water Cooled Ceramic Breeder) is a solid breeder blanket concept developed in Japan.

Structure and Materials

Component	Material	Specification
Structural material	F82H (reduced activation ferritic/martensitic steel)	Thickness 5-7 mm
Breeder	Li $_2$ TiO $_3$ pebbles	Diameter 0.2-2 mm
Multiplier	Be, Be $_{12}$ Ti	Pebble/plate form
Coolant	Pressurized water	15 MPa, 280-325 C
Tritium purge gas	He + H $_2$	Low pressure

Operating Principle

Similar to HCPB but using pressurized water as coolant. Water’s excellent heat transfer characteristics simplify cooling system design.

Design Parameters

TBR: Approximately 1.10
Energy multiplication factor $M$ : Approximately 1.18
Cooling water outlet temperature: 325 C

Liquid Blanket (Self-Cooled Concept)

“Self-cooled” concepts using liquid breeder material (Li or Li-Pb) as coolant can significantly simplify blanket structures.

DCLL (Dual Coolant Lithium Lead)

A concept developed in the United States using Li-Pb as breeder and coolant, with only the first wall and structural materials cooled by helium.

Component	Material	Specification
Structural material	RAFM steel / SiC/SiC
Breeder/Coolant	Li-Pb	Target outlet 700 C
First wall cooling	He gas	8 MPa
Flow channel insulation	SiC insert	MHD countermeasure

SiC/SiC composite Flow Channel Inserts (FCI) significantly reduce MHD pressure losses.

\Delta P_{\text{MHD}} \propto \frac{1}{R_{\text{wall}}}

Here, $R_{\text{wall}}$ is the wall resistance, which can be increased by insulating materials.

If high-temperature Li-Pb operation (700 C or higher) becomes possible, power generation efficiency exceeding 45% may be achievable.

ITER Test Blanket Module (TBM)

Purpose of the TBM Program

At ITER, the Test Blanket Module (TBM) program is being advanced to demonstrate blanket technology for the prototype reactor (DEMO). The main blanket at ITER is a shielding blanket without tritium breeding function, but TBMs are installed in dedicated ports to verify breeding blanket performance in the fusion environment.

Significance of TBM Testing

TBM testing will demonstrate the following:

Tritium production in actual fusion neutron environment
Operational verification of tritium recovery and measurement systems
Irradiation behavior of breeder and multiplier materials
Thermal-hydraulic performance of cooling systems
Structural integrity (thermal stress, electromagnetic forces, irradiation damage)
Remote handling for removal/maintenance

These data are extremely important as design basis for DEMO blankets.

TBM Concepts by Party

Japan (WCCB-TBM)

Developed by Japan Atomic Energy Agency (JAEA) and National Institutes for Quantum Science and Technology (QST).

Item	Specification
Structural material	F82H
Breeder	Li $_2$ TiO $_3$ pebbles (enriched Li)
Multiplier	Be pebbles, Be $_{12}$ Ti
Coolant	Pressurized water (15 MPa, 280-325 C)
Module dimensions	680 mm x 1940 mm x 600 mm
Expected TBR	Approximately 1.0 (TBM unit)

A feature of the Japanese TBM is the use of advanced beryllium compound (Be $_{12}$ Ti). This has smaller swelling under neutron irradiation than pure Be and superior mechanical properties.

Europe (HCPB-TBM, WCLL-TBM)

Europe is developing two types of TBMs in parallel.

HCPB-TBM:

Item	Specification
Structural material	EUROFER97
Breeder	Li $_4$ SiO $_4$ pebbles
Multiplier	Be pebbles
Coolant	He gas (8 MPa, 300-500 C)

WCLL-TBM:

Item	Specification
Structural material	EUROFER97
Breeder/Multiplier	Li-Pb alloy
Coolant	Pressurized water (15.5 MPa)

China (HCCB-TBM, HCLB-TBM)

China is developing both solid breeder (HCCB) and liquid breeder (HCLB) types.

HCCB (Helium Cooled Ceramic Breeder):

Structural material: CLF-1 steel (Chinese-developed reduced activation ferritic/martensitic steel)
Breeder: Li $_4$ SiO $_4$ pebbles
Coolant: He gas

HCLB (Helium Cooled Lithium-Lead Blanket):

Structural material: CLF-1 steel
Breeder/Multiplier: Li-Pb alloy
Coolant: He gas

Korea (HCSB-TBM)

HCSB (Helium Cooled Solid Breeder) TBM developed by Korea Atomic Energy Research Institute (KAERI).

Structural material: RAFM steel
Breeder: Li $_2$ TiO $_3$ pebbles
Multiplier: Be pebbles
Coolant: He gas

India (LLCB-TBM, HCCB-TBM)

India is developing both liquid breeder (LLCB) and solid breeder (HCCB) types.

TBM Ports and Installation Locations

TBMs are installed in three equatorial plane ports at ITER (Ports 16, 18, and 2). Two TBMs can be installed in parallel in each port.

TBMs will be installed in phases according to ITER operational phases.

Phase	Period	TBM Testing Content
Stage 1	Hydrogen/Helium operation	Structural integrity, cooling system performance
Stage 2	D-D operation	Tritium production at low neutron flux
Stage 3	D-T operation	Full-scale tritium breeding tests

DEMO Blanket Design

DEMO Blanket Requirements

The prototype reactor (DEMO) blanket must meet the following stringent requirements based on demonstration results from ITER TBM.

Item	ITER TBM	DEMO	Notes
TBR	Reference value	$\geq 1.05$	Self-sufficiency required
Neutron wall load	0.78 MW/m $^2$	1-2 MW/m $^2$	2-3 times increase
Annual fluence	0.1 MWa/m $^2$	2-3 MWa/m $^2$	20+ times increase
Operating time	Hundreds of second pulses	Continuous operation	Steady-state
Lifetime	Demonstration test	5+ years	Limited replacements
Power generation efficiency	-	30-45%	Economic requirement

European DEMO Blanket

European DEMO design (EU-DEMO) considers two concepts in parallel: HCPB and WCLL.

EU-DEMO HCPB

Item	Design Value
Breeder	Li $_4$ SiO $_4$ pebbles (enriched ${}^{6}$ Li 60%)
Multiplier	Be pebbles (70% volume fraction)
Structural material	EUROFER97
Coolant	He (8 MPa, 300-500 C)
TBR	1.12
M	1.25
First wall heat load	1 MW/m $^2$
Neutron wall load	1.2 MW/m $^2$
Design lifetime	5 fpy (full power years)

fpy stands for “full power year,” expressing cumulative operating time at rated power in years. 5 fpy corresponds to 10 years of operation at 50% of rated power.

EU-DEMO WCLL

Item	Design Value
Breeder/Multiplier	Li-Pb ( ${}^{6}$ Li enrichment 90%)
Structural material	EUROFER97
Coolant	Pressurized water (15.5 MPa, 295-328 C)
TBR	1.09
M	1.17
Power generation efficiency	33%

WCLL can leverage mature LWR technology, but HCPB is considered more favorable economically due to lower power generation efficiency with WCLL.

Japanese DEMO Blanket

Japanese prototype reactor design (JA-DEMO) is based on the WCCB concept, with advanced concepts using SiC/SiC composites for high-temperature blankets also being studied.

JA-DEMO WCCB

Item	Design Value
Breeder	Li $_2$ TiO $_3$ pebbles
Multiplier	Be $_{12}$ Ti
Structural material	F82H
Coolant	Pressurized water (15 MPa)
TBR	1.05 or higher
Neutron wall load	1.5 MW/m $^2$

Advanced SiC Blanket

For the long term, high-temperature blankets using SiC/SiC composites as structural material are being considered.

Item	Target Value
Structural material	SiC/SiC composites
Coolant outlet temperature	900-1000 C
Power generation efficiency	50% or higher

SiC/SiC has advantages of excellent high-temperature strength and low activation under neutron irradiation. However, many technical challenges remain, including joining technology, hermeticity assurance, and large structure fabrication.

Thermal-Hydraulic and Structural Analysis

Thermal-Hydraulic Analysis

Blanket thermal design considers the following heat loads:

Volumetric heating by neutrons
Volumetric heating by gamma rays
Surface heat load (radiation and particle flux from plasma)
Heating from nuclear reactions

Heating Distribution

Heating density due to neutrons decays exponentially in the blanket depth direction.

q'''(x) = q'''_0 \exp\left(-\frac{x}{\lambda}\right)

Here, $\lambda$ is the decay length (typically 5-10 cm).

In actual blankets, heating distribution becomes complex due to material combinations. 3D heating distributions are obtained by Monte Carlo calculations (MCNP, etc.) and input to CFD codes to calculate temperature distributions.

Temperature Field Calculation

Steady-state heat conduction equation:

\nabla \cdot (k \nabla T) + q''' = 0

Forced convection in cooling tubes:

q'' = h (T_{\text{wall}} - T_{\text{fluid}})

Here, $k$ is thermal conductivity and $h$ is heat transfer coefficient.

CFD analysis models effective thermal conductivity of pebble bed layers using experimental correlations.

k_{\text{eff}} = k_{\text{gas}} \left[ (1 - \sqrt{1 - \epsilon}) + \sqrt{1 - \epsilon} \cdot f(k_s/k_g) \right]

Here, $\epsilon$ is porosity, and $k_s$ , $k_g$ are solid and gas thermal conductivities.

Structural Analysis

The following loads act on blankets:

Thermal Stress

Thermal stress due to temperature gradients:

\sigma_{\text{thermal}} = E \alpha \Delta T

Here, $E$ is elastic modulus, $\alpha$ is linear expansion coefficient, and $\Delta T$ is temperature difference.

Thermal stresses from transient temperature changes during cooling start-up or shutdown often dominate structural integrity.

Pressure Loads

Loads from coolant pressure:

Water cooling: 15-15.5 MPa
Helium cooling: 8-10 MPa

Cooling tube membrane stress:

\sigma = \frac{P r}{t}

Here, $P$ is internal pressure, $r$ is tube radius, and $t$ is wall thickness.

Electromagnetic Forces

Large electromagnetic forces are generated in blanket structures due to rapid magnetic field changes during plasma disruptions.

Induced current density:

\mathbf{j} = \sigma \frac{\partial \mathbf{B}}{\partial t}

Electromagnetic force density:

\mathbf{f} = \mathbf{j} \times \mathbf{B}

Electromagnetic forces during disruptions can cause stresses of several hundred MPa in structural materials, making it an important design consideration.

Irradiation Damage

The following damage accumulates in structural materials due to neutron irradiation:

Atomic displacements (dpa: displacement per atom)
Helium production (He/dpa)
Hydrogen production (H/dpa)
Swelling (volume expansion)
Irradiation hardening and embrittlement

Under DEMO conditions (1.5 MW/m $^2$ , 5 fpy), approximately 70-80 dpa of irradiation damage accumulates in structural materials near the first wall.

\text{dpa} = \frac{\Phi_n \times \sigma_d}{N}

Here, $\Phi_n$ is neutron fluence, $\sigma_d$ is displacement cross-section, and $N$ is atomic number density.

Finite Element Analysis

Finite element method (FEM) stress analysis is used for structural integrity evaluation. Commercial codes such as ANSYS, ABAQUS, and NASTRAN, as well as dedicated analysis systems for ITER, are used.

Design rules conforming to ASME/RCC-MR standards are applied as evaluation criteria.

Primary stress: $\sigma \leq S_m$ (allowable stress)
Secondary stress: $\sigma \leq 3 S_m$
Fatigue evaluation: Cumulative fatigue damage $D \leq 1$
Creep-fatigue: Time fraction rule

Material Challenges

Structural Materials

The following characteristics are required for blanket structural materials:

Required Characteristic	Reason
High strength	Resistance to thermal stress and pressure loads
High-temperature strength	Maintaining mechanical properties at operating temperature
Irradiation resistance	Suppression of degradation under neutron irradiation
Reduced activation	Waste treatment, reduced worker exposure
Weldability	Ease of fabrication and repair
Compatibility	Chemical compatibility with coolant and breeder

Reduced Activation Ferritic/Martensitic Steel (RAFM)

These are the primary candidate materials for current DEMO designs.

Material Name	Developing Country	Main Composition
F82H	Japan	8Cr-2W-V-Ta
EUROFER97	Europe	9Cr-1W-V-Ta
CLAM	China	9Cr-1.5W-V-Ta

These materials eliminate long-lived activation elements such as molybdenum (Mo) and niobium (Nb) from conventional ferritic steels, replacing them with tungsten (W) and tantalum (Ta).

The target is for radioactivity levels to decay to hands-on (manual handling possible) levels 100 years after irradiation.

A(t) = A_0 \exp\left(-\frac{\ln 2}{t_{1/2}} \times t\right)

High-Temperature Operation Challenges

The upper temperature limit for RAFM steels is around 550 C. This is due to reduced creep strength and ductility loss from irradiation embrittlement.

DBTT (Ductile-Brittle Transition Temperature) shift:

\Delta \text{DBTT} = f(\text{dpa}, T_{\text{irr}}, \text{He})

Irradiation raises DBTT, increasing the risk of brittle fracture at low temperatures.

Breeder Material Challenges

Ceramic Breeders

Ceramic breeders such as Li $_2$ TiO $_3$ and Li $_4$ SiO $_4$ face the following challenges:

Degradation of tritium release characteristics due to irradiation
TBR reduction due to lithium burn-up (Li consumption)
Crystal structure changes
Swelling

Tritium release strongly depends on temperature and is described by Arrhenius-type diffusion-limited kinetics.

D = D_0 \exp\left(-\frac{E_a}{RT}\right)

Here, $D_0$ is the frequency factor and $E_a$ is the activation energy.

Liquid Breeders

Challenges with Li-Pb alloy:

Compatibility with structural materials (corrosion)
Low tritium solubility (difficult recovery)
${}^{210}\text{Po}$ production (from Bi impurities)

Tritium solubility in Li-Pb is extremely low (ppb order), requiring permeation extraction or gas bubbling methods for efficient recovery.

S_T = S_0 \exp\left(-\frac{\Delta H_s}{RT}\right)

Neutron Multiplier Material Challenges

Beryllium

Swelling due to irradiation (He production)
Mechanical property degradation
Tritium accumulation
Resource constraints

He is produced by $(n,\alpha)$ reactions in Be, causing irradiation swelling.

{}^{9}\text{Be} + n \rightarrow 2 {}^{4}\text{He} + 2n

Swelling rate depends on temperature and irradiation dose, increasing sharply above 500 C.

\frac{\Delta V}{V} = f(T, \phi_n t)

Advanced Beryllium Compounds

Be $_{12}$ Ti (beryllium titanide) has smaller swelling than pure Be and superior mechanical properties. Japan is developing it for ITER TBM and DEMO.

Property	Be	Be $_{12}$ Ti
Be density	100%	Approximately 90%
Swelling (600 C, 10 dpa)	Approximately 10%	Approximately 3%
Strength (room temperature)	300 MPa	400 MPa

Manufacturing Technology

Module Manufacturing

DEMO blankets consist of hundreds of modules. Each module has a surface area of approximately 1-2 m $^2$ and weighs several tons.

Manufacturing process:

First wall panel fabrication (HIP method, forging, machining)
Cooling tube fabrication and joining (welding, brazing)
Side wall and back wall fabrication
Module assembly (electron beam welding, diffusion bonding)
Breeder and multiplier loading
Leak testing, pressure testing
Non-destructive examination (UT, RT, PT)

Welding Technology

The following techniques are used for welding RAFM steels:

Welding Method	Application	Features
TIG welding	Cooling tube joining	Precision welding, thin wall
Electron beam welding	Module assembly	Deep penetration, low heat input
Laser welding	First wall panels	High speed, low distortion
Diffusion bonding	First wall/cooling tube	Dissimilar material joining

Welding of RAFM steels requires preheating (150-200 C) and post-weld heat treatment (750 C x 2h).

Quality Assurance

Strict quality control conforming to nuclear quality assurance programs (NQA-1, RCC-MR) is applied to blanket module manufacturing.

Main inspection items:

Non-destructive examination of welds (100%)
Leak testing (He leak inspection)
Hydrostatic testing (design pressure x 1.25-1.5)
Dimensional inspection
Material testing (mechanical properties, chemical composition)

Safety and Environmental Impact

Tritium Containment

Tritium produced in blankets must be properly contained and recovered. The multiple barrier concept prevents release to the environment.

Barrier hierarchy:

Breeder material matrix
Cooling tubes and structural materials
Blanket module outer shell
Vacuum vessel
Building (containment area)

Tritium permeation follows Sieverts’ law.

J = \frac{P_T}{d} \left( \sqrt{p_1} - \sqrt{p_2} \right)

Here, $P_T$ is the permeation coefficient, $d$ is wall thickness, and $p_1$ , $p_2$ are tritium partial pressures on both sides.

To reduce permeation, techniques for forming oxide layers of alumina (Al $_2$ O $_3$ ) or chromia (Cr $_2$ O $_3$ ) on structural material surfaces are being developed.

Activation and Waste

Blanket materials are activated by neutron irradiation and become radioactive waste after operation. The use of reduced activation materials alleviates waste treatment burden.

Waste classification targets:

Time Period	Radioactivity Level	Disposal Classification
Immediately after operation	High level	Shielded storage
50 years after operation	Intermediate level	Near-surface disposal
100 years after operation	Low level to clearance	Recyclable

Contact dose rate of RAFM steel decays to approximately 1 μSv/h after 100 years of irradiation, enabling hands-on maintenance.

Loss of Coolant Accident

Safety during Loss of Coolant Accident (LOCA) is also an important design consideration.

Decay heat removal:

Q_{\text{decay}}(t) = Q_0 \times 0.066 \times \left[ \left(\frac{t - t_s}{1\ \text{s}}\right)^{-0.2} - \left(\frac{t}{1\ \text{s}}\right)^{-0.2} \right]

Fusion reactors have smaller decay heat than fission reactors (approximately 1% of rated power), enabling designs where passive cooling (natural convection, radiation) can provide removal.

Future Outlook

Technology Challenge Priorities

To realize blanket technology, the following challenges must be addressed with priority:

TBR achievement demonstration (ITER TBM testing)
Establishment of tritium recovery technology
Confirmation of material integrity under long-term irradiation
Establishment of large module manufacturing technology
Demonstration of remote maintenance technology
Improved power generation efficiency (high-temperature operation)

Pathway to DEMO

Based on results from ITER TBM testing (scheduled to begin around 2035), DEMO blanket design and manufacturing will proceed.

Stage	Period	Main Activities
ITER TBM Testing	2035-2045	Breeding function demonstration
DEMO Engineering Design	2040-2050	Detailed design and prototyping
DEMO Construction	2050-2060	Module manufacturing and installation
DEMO Operation	2060-	Power generation demonstration

Advanced Concept Outlook

For the long term, the following advanced concepts are being considered:

SiC/SiC composite blankets (high-temperature operation, 50%+ power generation efficiency)
Molten salt blankets (FLiBe, etc.)
Advanced dual coolant concepts
Application of self-healing materials

If these technologies are realized, the economics of fusion power could be significantly improved.

Superconducting Coils - Coil technology required for magnetic confinement
Structural Materials - Candidate materials for blanket structures
ITER Project - The venue for test blanket testing
Tritium Management - Safe handling of tritium
Plasma-Facing Materials - Relation to first wall materials
Fuel Cycle - Coordination with tritium processing systems
Divertor - Relation to heat and particle exhaust