Breeding blanket and associated systems in fusion power plants have to ensure tritium breeding self­sufficiency, show a sufficient power conversion effi­ciency, and withstand high neutron fluences.50 TBMs for ITER should be representative for such blanket modules (see Giancarli eta/.23 and Chuyanov eta/.24).

Among the technical objectives of ITER, it is specifi­cally stated that ‘‘ITER should test tritium breeding module concepts that would lead in a future reactor to tritium self-sufficiency and to the extraction of high-grade heat and electricity production.’’51 The main testing objectives shall be

• Validation of structural integrity theoretical pre­dictions under combined and relevant thermal, mechanical, and electromagnetic loads

• Validation of tritium breeding predictions

• Validation of tritium recovery process efficiency and T-inventories in blanket materials

• Validation of thermal predictions for strongly heterogeneous breeding blanket concepts with volumetric heat sources

• Demonstration of the integral performance of the blanket systems.

Table 1 also provides values for the ITERTBM loading parameters and tentative requirements for DEMO as a next step. It is seen that the neutron wall load in ITER is relatively small, which requires specific measures to make meaningful nuclear tests with TBMs. Four versions of a TBM are considered with specific objectives as follows:

1. H-phase and H-He-phase: focus on electromag­netic behavior;

2. D-phase: focus on thermal and neutronic behavior;

3. First D-T-phase: DEMO-relevant data acquisi­tion on neutronics, tritium production and man­agement, and thermomechanics;

4. Second D-T-phase: DEMO-relevant data acquisi­tion with high duty, long pulses with an integral TBM.

All ITER parties consider helium-cooled ceramic breeder (HCCB) blankets. This type of blanket requires beryllium as the neutron multiplier and ferritic-martensitic steel as the structural material. The ceramic breeder is a Li-based compound, either Li2TiO3 or Li4SiO4, and is used in pebble beds. A water-cooled ceramic breeder (WCCB) blanket is proposed for the Japanese TBM.

ITER prepares three horizontal ports for TBM testing, and two TBMs can be installed in one port. Though TBMs will be installed and tested as part of the ITER activities, each TBM will be developed under the responsibility of the distinct ITER party. Four of the ITER parties, China, European Union (EU), Japan, and Russian Federation (RF), have made TBM design proposals with solid breeder materials, while the United States and Korea propose to test submodules integrated into one of them. All ceramic breeder-based TBMs use pebble beds and ferritic — martensitic steel structures and He coolant at 8 MPa with inlet temperature of 300 °C and outlet up to 500 °C depending on the operating conditions. Only the Japanese party proposes a water-cooled concept in addition. Figure 10 shows the typical arrangement of a port cell in ITER to accommodate TBMs.

4.15.2.1 Ceramic Breeder Requirements

The development of any ceramic breeder blanket concept toward demonstration and realization of fusion power must ensure that the ceramic breeder material meets the following specific requirements:

• Though the ceramic breeder has no structural function in the blanket, the pebbles or pellets must withstand the stresses induced under reactor operating conditions (pressures, temperatures, temperature gradients and thermal shocks, irradiation-induced swelling, creep) without an excessive fragmentation, which might result in degradation of the heat transfer parameters and purge gas flow, up to end of life (EOL) peak burnup and displacement damage.

• Stability of the ceramic at the maximum operating temperature with regard to lithium transport (e. g., by evaporation or redistribution).

• Compatibility between the ceramic and the structural material in the reference purge gas con­ditions under neutron irradiation. Compatibility is one of the criteria defining the maximum inter­face temperature between ceramic breeder and the structural material.

• Sufficiently low tritium residence time to mini­mize the tritium inventory in blanket and auxiliary systems that determine source term in off-normal and accidental conditions.

• Activation as low as possible under neutron irradi­ation, including activation from impurities, so as to reduce the D-T fuel cycle back-end issues (includes the materials’ recycling aspects).