Vacuum Vessel

The vacuum vessel is not only the container that confines plasma in a fusion reactor but also the structural backbone that houses and supports in-vessel components such as the blanket and divertor. In tokamak-type fusion reactors, a torus (doughnut)-shaped vacuum vessel is employed, within which high-temperature plasma exceeding 100 million degrees is confined by magnetic fields.

The vacuum vessel is one of the most critical structural components of a fusion reactor, and its design and fabrication require extremely advanced technologies. This chapter provides detailed coverage of the functions required of vacuum vessels, material selection, structural design, manufacturing technologies, and the specific specifications for ITER.

Basic Functions of the Vacuum Vessel

Multiple functions are required of the vacuum vessel. These functions are interrelated and sometimes require meeting conflicting demands, making the design a multifaceted optimization problem.

Function as a Vacuum Boundary

Maintaining Ultra-High Vacuum

The plasma density in a fusion reactor is approximately $10^{20}$ m $^{-3}$ . When impurities are generated from the wall due to plasma-wall interactions, radiation losses increase and become obstacles to plasma confinement and heating. Therefore, the vacuum level inside the vacuum vessel must be below $10^{-5}$ Pa (ultra-high vacuum) before introducing fuel gas.

The required vacuum level is determined from the perspective of keeping impurity-induced radiation losses within acceptable limits. The relationship between impurity concentration $n_Z$ in the plasma and background gas pressure $p$ is expressed using the recycling coefficient $R$ and particle confinement time $\tau_p$ as follows:

n_Z \approx \frac{p \cdot S}{k_B T} \cdot \frac{\tau_p}{1 - R}

Here, $S$ is the vessel inner surface area, $k_B$ is the Boltzmann constant, and $T$ is the wall temperature.

The radiation loss power $P_{rad}$ due to impurities is expressed using electron density $n_e$ , impurity density $n_Z$ , and radiation cooling coefficient $L_Z(T_e)$ as:

P_{rad} = n_e \cdot n_Z \cdot L_Z(T_e) \cdot V

where $V$ is the plasma volume. For light elements such as carbon and oxygen, $L_Z \sim 10^{-31}$ W m $^3$ , but for high-Z elements such as tungsten, $L_Z$ reaches $\sim 10^{-27}$ W m $^3$ , making heavy element impurity contamination particularly problematic.

Ultimate Vacuum and Pumping Systems

The ultimate vacuum of the vacuum vessel is determined by the balance between effective pumping speed $S_{eff}$ and outgassing rate $Q$ :

p = \frac{Q}{S_{eff}}

The outgassing rate $Q$ is given by the product of surface area $A$ and outgassing rate per unit area $q$ :

Q = q \cdot A

The typical outgassing rate from stainless steel surfaces is approximately $q \sim 10^{-6}$ Pa m $^3$ /(s m $^2$ ) at room temperature, but can be reduced to $q \sim 10^{-9}$ Pa m $^3$ /(s m $^2$ ) after baking.

For ITER-class vacuum vessels, where the inner surface area reaches approximately 1,000 m $^2$ , achieving a vacuum level of $10^{-5}$ Pa requires, even after baking:

S_{eff} = \frac{Q}{p} = \frac{10^{-9} \times 1000}{10^{-5}} = 0.1 \text{ m}^3/\text{s} = 100 \text{ L/s}

or higher effective pumping speed. In practice, turbomolecular pumps or cryopumps in the several thousand L/s class are used with safety margins included.

Leak Rate Requirements

The leak rate of the vacuum vessel is strictly limited to maintain the ultimate vacuum. The allowable leak rate $Q_L$ must satisfy:

Q_L < S_{eff} \cdot p_{target}

ITER imposes an extremely stringent requirement of $Q_L < 10^{-8}$ Pa m $^3$ /s. This corresponds to approximately 0.3 cm $^3$ or less per year when converted to air at standard conditions.

Achieving this leak rate in large vacuum vessels with total weld lengths reaching several kilometers requires strict quality control of welds and helium leak testing of all welded joints.

Structural Support Function

Weight Support for In-Vessel Components

The vacuum vessel supports in-vessel components such as the blanket, divertor, and shield. In the case of ITER, these total approximately 8,000 tonnes, and together with the vacuum vessel itself, must support loads exceeding approximately 16,000 tonnes.

Structural strength against static loads is evaluated by the relationship between allowable stress $\sigma_{allow}$ and acting stress $\sigma$ :

\sigma < \sigma_{allow} = \frac{\sigma_y}{S_F}

Here, $\sigma_y$ is the yield stress of the material, and $S_F$ is the safety factor (typically 1.5-3).

Electromagnetic Force Support

In fusion reactors, large electromagnetic forces are generated by the interaction between plasma current and magnetic fields. Particularly during disruptions (sudden collapse of the plasma), electromagnetic forces on the order of hundreds of MN are generated instantaneously, requiring the vacuum vessel to have structural strength to withstand these forces.

Electromagnetic forces during disruptions depend on the time rate of change of plasma current $dI_p/dt$ and plasma inductance $L_p$ , expressed as:

F_{em} \propto L_p \cdot I_p \cdot \frac{dI_p}{dt}

ITER’s vacuum vessel structural design accounts for cases where plasma current decreases from 15 MA to zero in a few milliseconds.

Support Structure

The vacuum vessel typically transmits gravitational loads to the foundation structure through 3-9 support legs. The support structure must meet the following requirements:

Absorption of displacement due to thermal expansion (vacuum vessel temperature varies from room temperature to approximately 200 degrees C)
Response to horizontal forces during earthquakes
Transmission of electromagnetic forces during disruptions

The thermal stress $\sigma_{th}$ in support legs is evaluated using temperature change $\Delta T$ , thermal expansion coefficient $\alpha$ , and Young’s modulus $E$ :

\sigma_{th} = E \cdot \alpha \cdot \Delta T

For SUS316L, with $\alpha \approx 16 \times 10^{-6}$ /K and $E \approx 200$ GPa, a temperature change of 100 K can produce thermal stresses as high as 320 MPa. To mitigate this, flexible supports or sliding supports are employed.

Radiation Shielding Function

Protection of superconducting coils and radiation workers from 14 MeV neutrons and gamma rays generated by D-T reactions is required. Since the insulation performance and mechanical properties of superconducting coils degrade due to neutron irradiation, the vacuum vessel serves to shield against radiation.

Neutron Shielding Mechanism

Shielding of fast neutrons is accomplished through a combination of moderation by elastic scattering and absorption reactions. The attenuation of neutron flux $\phi$ is expressed using the macroscopic cross-section $\Sigma$ and shielding thickness $x$ as:

\phi(x) = \phi_0 \cdot B(x) \cdot e^{-\Sigma x}

Here, $B(x)$ is the buildup factor, representing the contribution from scattering.

The macroscopic cross-section of stainless steel for 14 MeV neutrons is approximately $\Sigma \approx 0.2$ cm $^{-1}$ , and the half-value layer (thickness that halves the flux) is:

x_{1/2} = \frac{\ln 2}{\Sigma} \approx 3.5 \text{ cm}

For effective shielding, a shielding thickness of more than 10 half-value layers (approximately 35 cm) is necessary.

Thermal Neutron Absorption

Moderated thermal neutrons are efficiently absorbed by the $(n, \alpha)$ reaction with boron-10:

^{10}\text{B} + n \rightarrow ^{7}\text{Li} + ^{4}\text{He} + 2.79 \text{ MeV}

Since the thermal neutron absorption cross-section of boron-10 is extremely large at approximately 3,840 barns, effective thermal neutron shielding can be achieved with approximately 2% boron addition. ITER uses borated stainless steel (SUS430B7, SUS430B4) as shielding material.

Allowable Dose for Superconducting Coils

The insulating materials of superconducting coils (epoxy resins and polyimides) degrade due to neutron irradiation. The allowable cumulative dose is approximately $10^{7}$ Gy for glass fiber reinforced epoxy and approximately $10^{8}$ Gy for polyimide.

To ensure the allowable dose is not exceeded throughout ITER’s operational period (approximately 20 years), the design provides shielding by the vacuum vessel and blanket to suppress the neutron flux at the coil position to below $10^{9}$ n/(m $^2$ s).

Function as a Safety Barrier

The vacuum vessel functions as a barrier to confine radioactive materials. This is an important safety function based on the concept of “defense in depth” in nuclear facilities.

Tritium Confinement

Fusion reactors use tritium (hydrogen-3) as fuel. Since tritium is a radioactive substance, its confinement is a critical safety issue.

The diffusion coefficient $D$ of tritium in metallic materials is expressed as a function of temperature $T$ :

D = D_0 \cdot \exp\left(-\frac{E_a}{k_B T}\right)

For stainless steel, $D_0 \approx 5 \times 10^{-7}$ m $^2$ /s and activation energy $E_a \approx 0.5$ eV.

The permeation flux $J$ is proportional to the concentration gradient $dc/dx$ :

J = -D \cdot \frac{dc}{dx} \approx D \cdot \frac{c_0}{d}

Here, $c_0$ is the surface concentration and $d$ is the wall thickness. To reduce tritium permeation, barrier coatings such as alumina or chromium oxide are under consideration.

Confinement During Accidents

Confinement of radioactive materials must be maintained during loss of coolant accidents (LOCA) and loss of vacuum accidents (LOVA). The vacuum vessel is required to be resistant to the following accident scenarios:

Pressurization due to coolant rupture (maximum 0.2 MPa for ITER)
Hydrogen generation from water-beryllium reactions
Dust resuspension due to air ingress

The integrity of the vacuum vessel against design basis accidents (DBA) is evaluated by the combination of pressure and temperature:

\frac{p}{p_{design}} + \frac{T - T_{ref}}{T_{design} - T_{ref}} \leq 1

Here, $p_{design}$ is the design pressure, $T_{design}$ is the design temperature, and $T_{ref}$ is the reference temperature.

Material Selection

Austenitic Stainless Steel

Austenitic stainless steel is the most commonly used structural material for vacuum vessels. ITER employs SUS316L(N)-IG (ITER Grade).

Properties of SUS316L(N)-IG

SUS316L(N)-IG is a material optimized from conventional SUS316L for fusion reactor applications, with the following characteristics:

Property	Specification	Notes
Tensile strength	$\geq$ 450 MPa	Room temperature
0.2% yield strength	$\geq$ 185 MPa	Room temperature
Elongation	$\geq$ 40%	Room temperature
Nitrogen content	0.06-0.08%	Improved weldability
Cobalt content	$\leq$ 0.05%	Reduced activation
Boron content	$\leq$ 0.002%	Helium generation suppression

The addition of nitrogen contributes to stabilizing the austenite phase and increasing strength, while also optimizing the viscosity of the molten pool during electron beam welding to improve weld quality.

Activation Characteristics

Activation of fusion reactor materials occurs through nuclear transmutation by neutron irradiation. The major activation reactions in SUS316L are:

$^{58}\text{Ni}(n, p)^{58}\text{Co}$ (half-life 70.9 days)
$^{59}\text{Co}(n, \gamma)^{60}\text{Co}$ (half-life 5.27 years)
$^{54}\text{Fe}(n, p)^{54}\text{Mn}$ (half-life 312 days)

In particular, $^{60}\text{Co}$ has a long half-life and emits high-energy gamma rays, making it problematic from the perspective of exposure during maintenance operations and waste disposal. ITER Grade materials significantly suppress $^{60}\text{Co}$ generation by limiting cobalt content to 0.05% or less.

Inconel Alloys

In areas requiring high-temperature strength or under special environmental conditions, nickel-based superalloys such as Inconel may be used.

Inconel 625

Inconel 625 (UNS N06625) has the following properties:

Property	Specification
Tensile strength	827 MPa (room temperature)
0.2% yield strength	414 MPa (room temperature)
High-temperature strength (600 degrees C)	690 MPa
Corrosion resistance	Excellent

However, due to its high nickel content (approximately 60%), it is disadvantageous from an activation perspective, and its use is limited.

Inconel 718

Inconel 718 is a nickel-based alloy that achieves high strength through age hardening. It may be used for fasteners and support members subjected to high stress.

\sigma_{0.2\%} \approx 1100 \text{ MPa (after aging treatment)}

Reduced Activation Materials

For future fusion reactors, the adoption of reduced activation materials is being considered to reduce radioactive waste.

Reduced Activation Ferritic/Martensitic Steel (RAFM)

Reduced Activation Ferritic/Martensitic Steel replaces long-lived elements (Mo, Nb, Ni, Co) with W, V, and Ta.

Representative composition (F82H):

Fe-8Cr-2W-0.2V-0.04Ta-0.1C

It is expected that the radioactivity level will decrease to 1/1000 or less of SUS316L after 100 years following irradiation, enabling near-surface disposal.

Quantitative Evaluation of Reduced Activation

The activation characteristics of materials are evaluated by the time variation of contact dose rate $\dot{D}$ :

\dot{D}(t) = \sum_i A_i(t) \cdot \Gamma_i \cdot f_i

Here, $A_i(t)$ is the radioactivity of nuclide $i$ , $\Gamma_i$ is the gamma-ray constant, and $f_i$ is the geometrical factor.

As a design guideline for reduced activation materials, the target is to satisfy:

\dot{D}(100\text{ years}) < 10 \text{ }\mu\text{Sv/h}

at 100 years after shutdown.

Shielding Materials

In double-wall vacuum vessels, shielding materials are installed between the walls.

Borated Stainless Steel

Borated stainless steel is a shielding material with excellent thermal neutron absorption properties.

Material	Boron content	Application
SUS430B7	1.75%	Standard shielding
SUS430B4	1.0%	Emphasis on workability
SUS304B7	1.75%	Emphasis on corrosion resistance

The addition of boron reduces material ductility, so material selection must balance workability.

Ferritic Stainless Steel

Ferritic stainless steel (SUS430) is ferromagnetic, which has the effect of reducing toroidal field ripple. The magnetic field ripple $\delta$ is defined as:

\delta = \frac{B_{max} - B_{min}}{B_{max} + B_{min}}

The magnetic field concentration effect of ferritic steel can reduce ripple amplitude.

Structural Design

Double-Wall Structure

The ITER vacuum vessel employs a double-wall structure consisting of inner and outer walls. This structure has the following advantages:

Structural Advantages

Increased rigidity: High bending rigidity is obtained by connecting the double walls with ribs. The equivalent moment of inertia $I_{eq}$ is:

I_{eq} = \frac{t_1 (h_1 + h_2 + g)^3}{12} - \frac{(t_1 - t_2) g^3}{12}

Here, $t_1$ and $t_2$ are the thicknesses of the outer and inner walls, $h_1$ and $h_2$ are the wall heights, and $g$ is the inter-wall gap.

Shielding space provision: Shielding materials and cooling pipes can be accommodated between the walls.
Efficient cooling: Efficient heat removal is possible by circulating coolant between the walls.

Wall Thickness Design

The wall thickness of the vacuum vessel is determined considering the following factors:

Internal pressure loads (0.2 MPa during accidents)
Electromagnetic force loads (during disruptions)
Gravity loads (support of in-vessel components)
Thermal stress

The membrane stress $\sigma$ for a cylindrical shell under internal pressure is:

\sigma_\theta = \frac{p \cdot r}{t}, \quad \sigma_z = \frac{p \cdot r}{2t}

Here, $p$ is the internal pressure, $r$ is the radius, and $t$ is the wall thickness. For the ITER vacuum vessel outer wall ( $r \approx 10$ m) with 0.2 MPa internal pressure and allowable stress of 150 MPa:

t_{min} = \frac{p \cdot r}{\sigma_{allow}} = \frac{0.2 \times 10}{150} = 13 \text{ mm}

In practice, a wall thickness of 60 mm is adopted due to electromagnetic forces and manufacturing constraints.

Rib Structure

The rib structure connecting the double walls plays an important role in achieving both structural rigidity and electrical resistance.

Rib Arrangement

The ITER vacuum vessel has ribs arranged in both poloidal and toroidal directions:

Poloidal ribs: Primarily responsible for bending rigidity
Toroidal ribs: Primarily responsible for torsional rigidity
Diagonal ribs: Improve shear rigidity

The rib pitch $\lambda$ is determined in relation to buckling loads. The critical buckling stress $\sigma_{cr}$ for a plate is:

\sigma_{cr} = \frac{k \pi^2 E}{12(1-\nu^2)} \left(\frac{t}{\lambda}\right)^2

Here, $k$ is the buckling coefficient (dependent on boundary conditions), $E$ is Young’s modulus, and $\nu$ is Poisson’s ratio.

Rib Welding

Full penetration or partial penetration welding is used for welding ribs to walls. The strength of the weld joint is determined by the throat thickness $a$ and weld length $L$ :

F_{weld} = \sigma_{allow} \cdot a \cdot L

ITER performs ultrasonic testing (UT) and penetrant testing (PT) as non-destructive inspection of rib welds.

Electrical Resistance Design

In tokamaks, the toroidal one-turn electrical resistance of the vacuum vessel must be appropriately designed for plasma current induction.

Electrical Resistance Calculation

The toroidal electrical resistance $R$ of a torus-shaped conductor is:

R = \rho \cdot \frac{2\pi R_0}{A_{eff}}

Here, $\rho$ is the electrical resistivity, $R_0$ is the major radius, and $A_{eff}$ is the effective cross-sectional area.

For double-wall structures, the inner and outer walls form a parallel circuit:

\frac{1}{R_{total}} = \frac{1}{R_{inner}} + \frac{1}{R_{outer}} + \frac{1}{R_{ribs}}

Design Trade-offs

The magnitude of electrical resistance involves the following trade-offs:

Electrical resistance	Large	Small
Plasma breakdown	Favorable	Unfavorable
Shape control	Favorable	Unfavorable
Position stabilization	Unfavorable	Favorable
AC losses	Large	Small

ITER sets the toroidal electrical resistance at 7.9 micro-ohms to balance these requirements.

Port Design

Ports are provided on the vacuum vessel for connecting various equipment and performing maintenance operations. ITER has a total of 44 large ports, and their design is a critical issue directly related to fusion reactor operation and maintenance.

Upper Ports

Functions and Specifications

Upper ports are primarily used for the following purposes:

Installation of electron cyclotron heating (ECH) launchers
Access for diagnostic instruments
Blanket module replacement (partial)

ITER’s upper ports are arranged in 18 locations in the toroidal direction, with each port having an opening dimension of approximately 1.5 m x 2.0 m.

ECH Launcher Ports

The ECH system injects 170 GHz millimeter waves into the plasma for localized heating and current drive. The port structure requires:

Support for millimeter wave transmission windows
Accommodation of cooling systems
Avoidance of interference with shield structures

CVD diamond is used for millimeter wave transmission windows, and the thermal load is evaluated as:

P_{thermal} = P_{incident} \cdot (1 - T_{trans}) \cdot (1 - R_{refl})

Here, $T_{trans}$ is the transmittance and $R_{refl}$ is the reflectance.

Equatorial Ports

Equatorial ports are the main access points of the fusion reactor and are used for the most diverse purposes.

Neutral Beam Injection (NBI) Ports

The NBI system injects high-energy (1 MeV class) deuterium beams into the plasma for heating and current drive. ITER has 2 NBIs installed, with each port having an opening dimension of approximately 2.0 m x 3.5 m.

Design challenges for NBI ports:

Reduced structural rigidity due to large openings
Connection to beamline vacuum
Shielding against neutron streaming

Neutron streaming is the phenomenon where neutrons leak to the outside through port openings and is important in shielding design. The streaming coefficient $S$ is:

S = \frac{\phi_{port}}{\phi_{wall}} \approx \frac{A_{port}}{4\pi r^2} \cdot e^{-\Sigma L}

Here, $A_{port}$ is the port opening area, $r$ is the distance, and $L$ is the port length.

Ion Cyclotron Heating (ICH) Ports

The ICH system injects 40-55 MHz RF power into the plasma for ion heating. ICH antennas are installed inside the vacuum vessel and are fed through ports.

The antenna coupling efficiency $\eta$ strongly depends on the distance $d$ between the plasma and antenna:

\eta \propto e^{-2k_\perp d}

Here, $k_\perp$ is the perpendicular component of the wave number. Therefore, designs that position the antenna as close to the plasma as possible are required.

Diagnostic Ports

Numerous diagnostic instruments are installed in equatorial ports for plasma diagnostics:

Thomson scattering system (electron temperature and density measurement)
Charge exchange recombination spectroscopy (ion temperature measurement)
Neutron measurement system (fusion power measurement)
Bolometers (radiation loss measurement)

Each diagnostic system must maintain the vacuum boundary while ensuring optical and electrical access.

Lower Ports

Divertor Maintenance Ports

The primary purpose of lower ports is remote maintenance and replacement of divertor cassettes. The ITER divertor consists of 54 cassettes, all of which must be replaced at end of life.

Divertor replacement follows these procedures:

Insertion of remote handling equipment into the vacuum vessel
Release and extraction of cassette fixation
Insertion and fixation of new cassette
Vacuum testing and operation preparation

The lower port opening dimensions (approximately 1.5 m x 1.5 m) correspond to the cassette size (approximately 1.2 m x 1.0 m x 0.8 m).

Pumping System Connection

The vacuum pumping system is connected to the vacuum vessel through lower ports. The effective pumping speed $S_{eff}$ of cryopumps is limited by the port conductance $C$ :

\frac{1}{S_{eff}} = \frac{1}{S_{pump}} + \frac{1}{C}

The conductance of a cylindrical tube (molecular flow regime) is given by:

C = \frac{\pi d^4}{12L} \cdot \sqrt{\frac{\pi R T}{2M}} \approx 12.1 \cdot \frac{d^4}{L} \text{ [L/s, cm, air]}

Electromagnetic Force Countermeasures

The vacuum vessel of a fusion reactor is exposed to powerful electromagnetic forces. Transient electromagnetic forces during disruptions, in particular, are one of the greatest challenges in structural design.

Halo Currents

During disruptions, “halo currents” flow where part of the plasma current flows through the vacuum vessel wall.

Mechanism of Halo Current Generation

During a disruption, the plasma column contracts and moves, and when magnetic field lines contact the vacuum vessel wall, plasma current flows through the wall. Halo current $I_{halo}$ is a certain fraction of plasma current $I_p$ :

I_{halo} = f_{halo} \cdot I_p

ITER’s design assumes $f_{halo} \leq 0.4$ .

Electromagnetic Forces from Halo Currents

The force generated by the interaction of halo current with toroidal magnetic field $B_T$ is:

\vec{F} = I_{halo} \cdot \vec{L} \times \vec{B}_T

This force concentrates in the region where the plasma contacts the wall (typically near the divertor) and locally reaches several MN/m.

Toroidal Asymmetry

Halo currents may be distributed asymmetrically in the toroidal direction (peaking factor $TPF$ ), in which case a net lateral force is generated:

F_{lateral} = TPF \cdot f_{halo} \cdot I_p \cdot B_T \cdot 2\pi R_0

ITER assumes $TPF \leq 2$ and is designed to withstand approximately 50 MN of lateral force.

Eddy Currents

Rapid changes in plasma current induce eddy currents in the vacuum vessel.

Eddy Current Generation

According to Faraday’s law, magnetic flux change $d\Phi/dt$ induces eddy currents:

\oint \vec{E} \cdot d\vec{l} = -\frac{d\Phi}{dt}

The induced eddy current $I_{eddy}$ is:

I_{eddy} = \frac{1}{R} \cdot \frac{d\Phi}{dt} = \frac{M}{R} \cdot \frac{dI_p}{dt}

Here, $M$ is the mutual inductance between the plasma and vacuum vessel, and $R$ is the electrical resistance of the vacuum vessel.

Joule Heating from Eddy Currents

Joule heating $P$ from eddy currents is:

P = I_{eddy}^2 \cdot R = \frac{M^2}{R} \cdot \left(\frac{dI_p}{dt}\right)^2

During disruptions, hundreds of MJ of energy may be deposited in the vacuum vessel. This energy is absorbed by the thermal capacity of the vacuum vessel, but local temperature rises can reach several tens of degrees.

Electromagnetic Forces

The interaction between eddy currents and magnetic fields produces electromagnetic forces on the vacuum vessel:

\vec{F} = \int \vec{J} \times \vec{B} \, dV

These forces act in directions that compress or expand the vacuum vessel and are considered in structural design.

Electromagnetic Force Reduction Measures

Electrical Insulation

Electrical insulation may be provided in the toroidal direction of the vacuum vessel to reduce eddy currents. The insulation breaks the eddy current path and reduces current values.

However, insulated sections become structural weak points, requiring careful design. ITER considered designs using ceramic insulators at port connections, but this was not adopted from a reliability perspective.

Optimization of Resistance Distribution

The distribution of electromagnetic forces can be controlled by optimizing the distribution of electrical resistance. For example, reducing resistance at locations prone to stress concentration can mitigate electromagnetic force concentration at those locations.

Structural Design

The following methods are used for structural design against electromagnetic forces:

Addition of reinforcing ribs
Local increase in wall thickness
Strengthening of support structures

Electromagnetic-structural coupled analysis using the finite element method (FEM) is widely used for design verification.

Baking and Wall Conditioning

Baking (thermal outgassing) and wall conditioning are performed to achieve ultra-high vacuum and prepare wall conditions suitable for plasma operation.

High-Temperature Baking

Purpose and Principle

Baking is a treatment that desorbs adsorbed gases from the surfaces of the vacuum vessel and in-vessel components, reducing the outgassing rate. Gas desorption from metal surfaces increases exponentially with temperature rise:

q = q_0 \cdot \exp\left(-\frac{E_d}{k_B T}\right)

Here, $E_d$ is the activation energy for desorption. By forcibly releasing gases at high temperature and then returning to room temperature, adsorbed gases on the surface are reduced, achieving a low outgassing rate.

Baking Temperature

Baking temperature varies depending on the gas species to be desorbed:

Gas species	Desorption temperature	Notes
Water	100-150 degrees C	Major impurity source
Hydrocarbons	150-200 degrees C	Organic contaminants
Hydrogen	200-300 degrees C	Dissolved in metals
Oxygen	300-400 degrees C	Oxide film decomposition

ITER plans baking at 200 degrees C. At this temperature, water and hydrocarbons are effectively removed while minimizing impact on the mechanical properties of structural materials.

Heating Methods

The following methods are used for baking large vacuum vessels:

Heat transfer fluid circulation: A method of circulating hot water or nitrogen gas between the double walls. Uniform heating is possible, and this is adopted for ITER.
Electric heaters: A method of installing electric heaters on the outer surface of the vacuum vessel. Local heating is possible, but ensuring uniformity is a challenge.
Induction heating: Induction heating using high-frequency electromagnetic fields. Rapid heating is possible, but unsuitable for large structures.

Thermal Stress Management

Temperature gradients during baking generate thermal stresses. The allowable temperature gradient $\Delta T_{allow}$ is limited by:

\Delta T_{allow} = \frac{\sigma_{allow}}{\alpha \cdot E}

For SUS316L, $\Delta T_{allow} \approx 50$ K/m, and heating rates are limited to a few degrees per hour.

Discharge Cleaning

Glow Discharge Cleaning (GDC)

Glow discharge cleaning is a method that generates a glow discharge in low-pressure gas and removes surface impurities by ion sputtering.

Gases used:

Hydrogen/Deuterium: Reduction of oxides, removal of hydrocarbons
Helium: High sputtering efficiency
Argon: Effective for physical sputtering

Discharge parameters:

Gas pressure: 0.1-1 Pa
Discharge voltage: 300-500 V
Current density: 0.1-1 mA/cm $^2$

Sputtering yield $Y$ depends on the energy $E$ of incident ions:

Y = \frac{3\alpha}{4\pi^2} \cdot \frac{M_1 M_2}{(M_1 + M_2)^2} \cdot \frac{E}{U_s}

Here, $\alpha$ is the angular factor, $M_1$ and $M_2$ are the masses of the ion and target, and $U_s$ is the surface binding energy.

Ion Cyclotron Resonance Cleaning (ICRF Cleaning)

In tokamaks, efficient wall cleaning is possible by applying RF at the ion cyclotron frequency in a magnetic field. The ion cyclotron frequency $f_{ci}$ is given by:

f_{ci} = \frac{eB}{2\pi m_i}

For hydrogen at 1 T magnetic field, this is approximately 15 MHz.

Boronization

Purpose

Boronization is a technique that forms a thin boron film on wall surfaces to capture oxygen and carbon impurities. The boron film functions as an oxygen getter:

2\text{B} + \frac{3}{2}\text{O}_2 \rightarrow \text{B}_2\text{O}_3

This reaction reduces oxygen concentration in the plasma.

Method

Diborane (B $_2$ H $_6$ ) or trimethylboron (TMB) is used as the source gas and decomposed and deposited in a glow discharge:

\text{B}_2\text{H}_6 \rightarrow 2\text{B} + 3\text{H}_2

Typical conditions:

Source gas pressure: 0.1-0.5 Pa (diluted with He)
Discharge duration: Several hours to tens of hours
Film thickness: 50-200 nm

Detailed Specifications of the ITER Vacuum Vessel

ITER is the world’s largest superconducting tokamak, and its vacuum vessel is of unprecedented scale.

Main Parameters

Parameter	Value	Notes
Height	11.3 m	Highest to lowest point
Torus inner radius	3.2 m	Inner wall position
Torus outer radius	9.7 m	Outer wall position
Major radius	6.2 m	Plasma center
Wall thickness (inner)	337 mm	Double wall + shielding
Wall thickness (outer)	750 mm	Double wall + shielding
Inner wall plate thickness	60 mm	SUS316L(N)-IG
Outer wall plate thickness	60 mm	SUS316L(N)-IG
Rib plate thickness	40 mm	SUS316L(N)-IG
Total weight	Approx. 8,450 tonnes	Main body only
Weight including ports	Approx. 11,000 tonnes
Inner surface area	Approx. 1,000 m $^2$
Internal volume	Approx. 1,400 m $^3$
Vacuum level	$1 \times 10^{-5}$ Pa or less	Before operation
Leak rate	$1 \times 10^{-8}$ Pa m $^3$ /s or less	Total leak
Electrical resistance	7.9 micro-ohms	Toroidal direction
Internal pressure resistance	0.2 MPa or higher	During accidents
External pressure resistance	0.1 MPa	Under vacuum

Sector Configuration

The ITER vacuum vessel is fabricated in 9 sectors in the toroidal direction:

Sector Specifications

Item	Value
Number of sectors	9
Angle per sector	40 degrees
Weight per sector	Approx. 900 tonnes
Height per sector	11.3 m
Manufacturing sites	Korea, Europe

Manufacturing Process

Plate forming: 60 mm thick stainless steel plates are formed into D-shaped cross-sections by hot or cold forming
Welding: Combination of electron beam welding (EBW) and TIG welding
Rib assembly: Reinforcing ribs are welded between inner and outer walls
Shielding material installation: Borated stainless steel is placed between walls
Cooling pipe installation: Cooling water pipes are installed between double walls
Machining: Precision machining of port mounting surfaces and support sections
Inspection: Dimensional inspection, non-destructive testing, leak testing

Cooling System

Normal Operation

Cooling conditions during normal operation:

Parameter	Value
Cooling medium	Water
Inlet temperature	100 degrees C
Outlet temperature	120 degrees C
Pressure	1.1 MPa
Flow rate	Approx. 200 kg/s
Heat removal	Approx. 8 MW

Heat removal is mainly composed of:

Neutron heating: $\sim 4$ MW
Gamma-ray heating: $\sim 2$ MW
Heat conduction: $\sim 2$ MW

During Baking

Heating conditions during baking:

Parameter	Value
Heating medium	Water
Temperature	200 degrees C
Pressure	2.4 MPa
Heating rate	$\leq 5$ degrees C/h
Hold time	$\geq 200$ h

Port Configuration

The ITER vacuum vessel has the following ports:

Port type	Quantity	Main applications
Upper ports	18	ECH, diagnostics
Equatorial ports	17	NBI, ICH, diagnostics, maintenance
Lower ports	9	Divertor maintenance, pumping
Total	44

Installation of In-Vessel Components

The following are installed inside the vacuum vessel:

Blanket Modules

Quantity: 440
Total weight: Approx. 4,000 tonnes
Support method: Flexible supports
Installation accuracy: $\pm 3$ mm

Blanket modules are arranged along the inner wall of the vacuum vessel and fixed by flexible supports. Flexible supports absorb displacement due to thermal expansion while transmitting electromagnetic forces to the vacuum vessel.

Divertor Cassettes

Quantity: 54
Total weight: Approx. 700 tonnes
Support method: Support rails
Installation accuracy: $\pm 5$ mm

Divertor cassettes are arranged on support rails installed at the bottom of the vacuum vessel. During remote maintenance, cassettes are extracted along the rails and replaced.

Manufacturing and Assembly Technology

Welding Technology

Electron Beam Welding (EBW)

Electron beam welding is a welding method using an accelerated electron beam as a heat source in high vacuum, with the following characteristics:

High energy density ( $10^6$ W/cm $^2$ or higher)
Deep penetration (single-pass welding of thick plates possible)
Narrow heat-affected zone
Low heat input suppresses distortion

EBW is applied to major plate joints in the ITER vacuum vessel. Since 60 mm thick plates can be welded in a single pass, the number of welds is reduced, achieving improved quality and cost reduction.

The penetration depth $d$ of EBW depends on beam power $P$ , welding speed $v$ , and material thermal properties, approximated as:

d \propto \frac{P}{v \cdot \sqrt{\alpha}}

Here, $\alpha$ is the thermal diffusivity.

TIG Welding

TIG (Tungsten Inert Gas) welding is used for the following purposes:

Complex geometry parts where EBW is difficult to apply
On-site joining (sector-to-sector welding)
Repair welding

Since TIG welding has higher heat input than EBW, welding distortion management is important. For multi-layer welding, the heat input $Q$ for each layer is calculated as:

Q = \frac{\eta \cdot V \cdot I}{v}

Here, $\eta$ is the welding efficiency (approximately 0.6 for TIG), $V$ is the welding voltage, $I$ is the welding current, and $v$ is the welding speed.

Non-Destructive Testing

Testing Methods

The following non-destructive tests are applied to welds on the ITER vacuum vessel:

Testing method	Target	Detectable defects
Visual testing (VT)	All welds	Surface defects
Penetrant testing (PT)	Surface welds	Surface-breaking defects
Ultrasonic testing (UT)	Butt welds	Internal defects
Radiographic testing (RT)	Samples	Internal defects
Helium leak testing	Vacuum boundary	Leaks

Detection Limits

Detection limits for each testing method are as follows:

PT: Surface defects with opening width $\geq 1$ micrometer
UT: Internal defects equivalent to diameter $\geq 2$ mm
He leak: Leaks of $10^{-10}$ Pa m $^3$ /s or higher

Manufacturing Accuracy Control

Dimensional Accuracy

Manufacturing accuracy requirements for the ITER vacuum vessel:

Item	Tolerance
Overall manufacturing accuracy	$\pm 20$ mm
Individual sector	$\pm 10$ mm
Weld groove position	$\pm 5$ mm
Port mounting surface	$\pm 2$ mm

Shape Measurement

The following methods are used for shape measurement of large structures:

Laser tracker: Accuracy $\pm 0.1$ mm/m
Photogrammetry: Accuracy $\pm 0.5$ mm/m
Coordinate measuring machine (CMM): Accuracy $\pm 0.01$ mm

These measurement data are integrated and manufacturing accuracy is evaluated by comparison with CAD models.

On-Site Assembly

Sector Delivery

Sector delivery to the ITER site follows these procedures:

Marine transport: From manufacturing plants to Fos-sur-Mer port
Land transport: By special vehicles (SPMT) to the ITER site (approximately 100 km)
Delivery to assembly building: Lifting by large cranes

Each sector’s transport weight reaches approximately 900 tonnes, requiring special measures such as road and bridge reinforcement and nighttime transport.

Sector-to-Sector Welding

Sector joining is performed in the assembly building at the ITER site:

Positioning: Sectors are placed between 9 TF coils
Groove preparation: Cleaning and inspection of weld grooves
Root pass welding: Initial layer by TIG welding
Multi-layer welding: TIG or narrow-gap welding
Non-destructive testing: UT, PT
Leak testing: He leak testing

The total length of on-site welding reaches approximately 1 km, with an estimated construction period of approximately 2 years.

Accommodation of Remote Maintenance

The interior of a fusion reactor vacuum vessel becomes a high-level radiation environment after operation, so maintenance and replacement of in-vessel components must be performed by remote handling.

Radiation Environment

After ITER operation, the radiation environment inside the vacuum vessel is predicted as follows:

Location	Dose rate (24 hours after shutdown)
First wall surface	$\sim 100$ Gy/h
Divertor	$\sim 500$ Gy/h
Inside ports	$\sim 10$ Gy/h

Considering that the lethal dose for humans is approximately 4 Gy, direct access is impossible, and all maintenance operations must be performed by remote handling.

Remote Maintenance Systems

Blanket Replacement System

Blanket module replacement is performed by remote handling manipulators inserted through equatorial ports:

Manipulator payload capacity: Approx. 4 tonnes
Positioning accuracy: $\pm 5$ mm
Work time: Approx. 8 hours per module

Replacement procedure:

Removal of module fixing bolts
Cutting of cooling pipes
Extraction of module
Insertion of new module
Connection of cooling pipes
Tightening of fixing bolts

Divertor Replacement System

Divertor cassette replacement is performed through lower ports:

Cask weight: Approx. 50 tonnes (including cassette)
Replacement time: Approx. 1 week per cassette
Total replacement time: Approx. 6 months

Requirements for Vacuum Vessel

The following are required in vacuum vessel design to enable remote maintenance:

Port Dimensions

Port opening dimensions must be sufficient for remote handling equipment and cassettes to pass through. ITER’s equatorial ports have openings of approximately 2 m x 3.5 m.

Internal Surface Geometry

To ensure movement paths for remote handling equipment, protrusions on the inner surface of the vacuum vessel should be minimized, and smooth geometry is required.

Positioning Mechanisms

High-precision positioning is required for installation and removal of in-vessel components. Guide rails and positioning pins are installed on the vacuum vessel.

Visibility

Installation points for cameras and lighting are considered to ensure visibility during remote handling.

In-Service Inspection

Inspections are performed during operation and shutdown to confirm the integrity of the vacuum vessel:

Inspection Items

Inspection item	Method	Frequency
Leak detection	Mass spectrometer	Continuous
Wall thickness measurement	Remote UT	Periodic (5 years)
Weld inspection	Remote UT/PT	Periodic (5 years)
Deformation measurement	Laser measurement	Periodic (1 year)

Challenges in Inspection Technology

Inspection in high-radiation environments presents the following challenges:

Radiation degradation of electronic equipment
Ensuring visibility
Access routes for inspection equipment

To address these challenges, development of radiation-resistant inspection equipment and remote handling technology is progressing.

Future Challenges

This section describes the main development challenges in vacuum vessel technology.

Development of Structural Standards

The vacuum vessel of a fusion reactor is positioned as safety-critical equipment as a radioactive material confinement barrier. However, existing pressure vessel standards (ASME, JIS, etc.) do not adequately consider loading conditions specific to fusion (electromagnetic forces, neutron irradiation, etc.).

The following need to be developed as structural standards for fusion reactors:

Design criteria considering electromagnetic forces
Material property data under neutron irradiation
In-service period standards assuming remote inspection and repair

Development of Reduced Activation Materials

Reduction of radioactive waste is an important issue for future commercial fusion reactors. Application of reduced activation ferritic steel and vanadium alloys is also being considered for vacuum vessel materials.

Development challenges for these materials:

Application technology to large structures (welding, forming)
Accumulation of long-term reliability data
Cost reduction

Reduction of Manufacturing Costs

The manufacturing cost of the ITER vacuum vessel reaches several billion dollars. Commercial reactors require significant cost reduction:

Automation and robotization of welding
Improved manufacturing efficiency through modularization
Reduction of material and processing costs

Improvement of Maintainability

Efficiency of remote maintenance directly affects reactor availability:

Reduction of maintenance time
Improved reliability of maintenance systems
Consideration of maintainability from the design stage (DFM: Design for Maintenance)

Tokamak Approach - Magnetic configuration and components of tokamaks
Superconducting Coils - Superconducting magnets generating strong magnetic fields
ITER Project - The world’s largest tokamak experimental reactor
Plasma-Facing Materials - Materials for first wall and divertor
Structural Materials - Structural materials for fusion reactors
Tritium Management - Safe handling of tritium