Lithium Lead Test Blanket Module Research Articles

RELAP5/Mod3.3 MHD module development and validation: WCLL-TBM mock-up model

Magnetohydrodynamic (MHD) phenomena significantly impact the design of magnetic-confinement fusion reactors, particularly in the context of breeding blankets (BB) utilizing liquid metals (LMs) as working fluids. These effects arise from the interaction between the electro-conductive flowing metal and the magnetic field used for plasma confinement in the reactor chamber. Induced electrical currents in the liquid generate Lorentz forces, thereby altering flow behaviour in comparison to standard hydrodynamic conditions. For example, MHD effects modify velocity distribution and mass transport within ducts, amplify pressure losses, and influence heat transfer mechanisms. Accurate estimation of these impacts is crucial for the effective design of a liquid metal breeding blanket. While computational tools are essential for fusion-related physical analyses, no comprehensive MHD code currently exists for simulating all relevant phenomena in a liquid metal blanket. In this context, models predicting both distributed and concentrated MHD pressure drops have been integrated into the thermal-hydraulic system code RELAP5/Mod3.3. The Verification and Validation (V&V) process compares code results to direct numerical simulations and experimental data. For validation, RELAP5 recreates experimental results of a Water Cooled Lithium Lead test blanket module at magnetic field intensities ranging from Ha=500−3000, confirming the reliability of the newly implemented MHD subroutines for predicting pressure drops within this parameter range.

Preliminary assessment of tritium permeation and retention in the European Water Cooled Lithium Lead Test Blanket Module with TESSIM-X

• T inventory in the ITER EU Water-Cooled Lithium-Lead Test Blanket Module investigated. • T from vacuum vessel (gas, neutrals) and solute T in PbLi included in modeling. • Effect of neutron damaging on trap concentrations included in models. • T retention in the range of 0.75 g is expected over plasma operation period. • Comparison with the TMAP7 code performed. Tritium self-sufficiency is a key requirement for future fusion power plants. Therefore, it will be necessary to minimize tritium losses to the surrounding systems, such as the tritium breeding modules, plasma-facing components, cooling system, etc. These components and systems will act also as a tritium sink, as tritium will be retained in the metal walls in traps produced during manufacturing or induced by neutron irradiation. The design of the tritium breeding systems and plasma-facing components will therefore have a direct impact on the performance, operational safety and cost of any future power plant. Consequently, accurate modeling of tritium transport and retention in these systems is needed for future engineering designs of the tritium breeding systems and supporting safety requirements. In this work, tritium permeation and retention was modeled with the diffusion-transport code TESSIM-X for the current European WCLL (Water Cooled Lithium Lead) Test Blanket Module (TBM) for the ITER TBM Programme, with the inclusion of neutron-induced traps. Trap saturation is considered for dpa values of 0.25 and above, leading to an increase in trap concentrations relative to undamaged material of roughly a factor 3. The code was also benchmarked against the well-established TMAP7 code.

Preliminary MHD pressure drop analysis for the prototypical WCLL TBM with RELAP5/MOD3.3

In fusion reactor development, the design of the breeding blanket (BB) represents an important technical challenge due to the extreme operative conditions that the component must withstand. One of the critical issues of a liquid metal BB is the occurring of magnetohydrodynamic (MHD) effects within the piping network that transports the electro-conductive fluid. Predicting MHD phenomena is fundamental for a reliable and efficient blanket design and computational tools are indispensable to achieve a fusion-relevant physical analysis. For this reason, a dedicated MHD module for the System Thermal-Hydraulic (STH) code RELAP5/MOD3.3 (RELAP5) is under development at Sapienza University of Rome. In its current state, RELAP5 can evaluate MHD pressure losses in many common layouts for BB piping system. In a previous work of the Authors, the verification and validation (V&V) procedure of the MHD module has been discussed. In this paper, the code is used to assess the MHD pressure loss in the Water-Cooled Lithium Lead Test Blanket Module (TBM). Furthermore, the geometry of the TBM box is optimized to reduce the MHD pressure loss and to equalize the flow distribution among the Breeding Units.

Severe accident analysis of the dual-functional lithium-lead test blanket module using ISAA-DFLL code

The international thermonuclear experimental reactor (ITER) project aims to build a tokamak fusion test reactor to verify the feasibility of fusion reactors. The test blanket module (TBM) is the key component of the international thermonuclear experimental reactor. The fusion design study team proposed the concept of the dual-functional lithium-lead test blanket module (DFLL-TBM). The integral severe accident analysis (ISAA) is a self-developed ISAA code that models the progression of severe accidents in nuclear power plants. A broad spectrum of severe accident phenomena including fission product release and transport behavior is modeled in ISAA. In this paper, a new version of ISAA, referred to as ISAA-DFLL is introduced for the application of DFLL-TBM test blanket module into the treatment of multi fluids and the modules of the new physical property and heat transfer. The modification is verified by comparing the steady-state temperature distribution of the DFLL-TBM first wall with the design parameters. Then accident analysis of In-vessel loss of helium coolant in first wall and TBM pipe is conducted by using the ISAA-DFLL code. By comparing the calculation results with the general safety requirements for TBM, it is concluded that the design of the DFLL-TBM system and the modifications of the current version are reasonable and accurate.

HyPer-QuarCh II: A laboratory-scale device for hydrogen isotopes permeation experiments

In a D-T fusion reactor, the correct estimation of the tritium inventory and permeation fluxes towards the coolant and to the external environment is a crucial issue for the reactor licensing. Within this frame, a fast and reliable sensor for the online measurement of hydrogen isotopes concentration in the breeder is therefore necessary. At ENEA Brasimone research centre, Italy, the development, qualification and characterization of hydrogen isotopes permeation sensors (HPS) were carried out since the early 2000s. A new experimental laboratory-scale device, named Hyper-Quarch II (Hydrogen Permeation Quartz Chamber), was developed on the basis of the experience gained in the past experimental campaigns. This device is characterised by an upgraded test section in quartz and new instrumentation equipment, and will be used to test advanced hydrogen permeation sensor in both gas phase and in stagnant LiPb eutectic alloy (15.7 at.% Li). Hydrogen or deuterium will be adopted instead of tritium to simulate the operative conditions of the Water-Cooled Lithium-Lead (WCLL) Test Blanket Module of ITER and the WCLL Breeding Blanket of the European DEMO reactor. Moreover, the upgrade was performed to also allow the measurement of the Sieverts’ constant of hydrogen or deuterium solubilised in the LiPb with absorption or desorption techniques in a temperature range from 300 to 550∘C and pressure range 0.1 to 100 hPa.

Conceptual design overview of the ITER WCLL Water Cooling System and supporting thermal-hydraulic analysis

In this paper, the conceptual design of the International Thermonuclear Experimental Reactor (ITER) Water-Cooled Lithium-Lead (WCLL) Test Blanket Module (TBM) Water Cooling System (WCS) from Europe is presented. The system consists of two loops in series. This design feature allows the removal of heat from the TBM box avoiding at the same time the release of radionuclides into the ITER Component Cooling Water System (CCWS), that acts as WCLL Test Blanket System heat sink. For this purpose, the WCS primary loop deals with the direct heat removal from the ITER TBM and the secondary one implements physical separation between the contaminated primary loop coolant and the CCWS. The insertion of an economizer into the primary loop determines the characteristic “eight” shape of the circuit. This choice was done in order to reduce the temperature difference on the intermediate heat exchanger. Hairpin type and steam bubble are the technologies selected for heat exchangers and pressurizers, respectively. Pressure and temperature control systems are foreseen to limit excursions from rated values in normal operational states and abnormal transients.A computational activity was promoted to assess the WCLL-WCS conceptual design, using a modified version of the RELAP5 Mod3.3 system code. A detailed thermal-hydraulic model was developed on the basis of design outcomes. The nodalization scheme includes the TBM, the WCS, a portion of the CCWS and the lithium-lead circuit. The computational campaign involved both the normal operational state and selected abnormal transients. In all the scenarios simulated, the conceptual design has highlighted the capability of operating the system respecting all the thermal-hydraulic requirements. The abnormal transient selected and presented is the loss of flow in the CCWS (loss of heat sink). In these conditions, TBM cooling function has been verified, keeping standard control strategies without any external action.

NUCLEAR DATA SENSITIVITY/UNCERTAINTY PRE-ANALYSIS OF FNG WCLL FUSION BENCHMARK

To assure tritium self-sufficiency in future fusion reactors such as DEMO the accuracy of TRP calculations has to be demonstrated within the design uncertainties. A new neutronics experiment representing a mock-up of the Water Cooled Lithium Lead (WCLL) Test Blanket Module (TBM) is under preparation at the Frascati neutron generator (FNG) with the objective to provide an experimental validation of accuracy of nuclear data and neutron transport codes for the tritium production rate (TPR) calculations. The mock-up will consist of LiPb bricks, EUROFER plates and Perspex substituting water. The mock-up will be irradiated by 14 MeV neutrons at the FNG facility, and the TPR and detector reaction rates will be measured using Li2CO3 pellets and activation foils placed at different positions up to about 55 cm inside the mock-up. Computational pre-analyses for the design of the WCLL neutronics experiment using the SUSD3D sensitivity/uncertainty (S/U) code system is described and compared with the results of some similar FNG experiments performed in the past, in particular the FNG HCPB Tritium Breeder Module Mock-up (2005) and FNG-HCLL Tritium Breeder Module Mock-up (2009). The objective of the pre-analysis is to provide the calculated nuclear responses including the uncertainties due to the uncertainties in nuclear data and thus contributes to the optimisation of the design of the experimental set-up.

Neutronic calculations of the China dual-functional lithium–lead test blanket module with the parallel discrete ordinates code Hydra

The China dual-functional lithium–lead test blanket module (DFLL-TBM) is a liquid LiPb blanket concept developed by the Institute of Nuclear Energy Safety Technology of the Chinese Academy of Sciences for testing in ITER to validate relevant tritium breeding and shielding technologies. In this study, neutronic calculations of DFLL-TBM were carried out using a massively parallel three-dimensional transport code, Hydra, with the Fusion Evaluated Nuclear Data Library/MG. Hydra was developed by the Nuclear Engineering Computational Physics Lab based on the discrete ordinates method and has been devoted to neutronic analysis and shielding evaluation for nuclear facilities. An in-house Monte Carlo code (MCX) was employed to verify the discretized calculation model used by Hydra for the DFLL-TBM calculations. The results showed two key aspects: (1) In most material zones, Hydra solutions are in good agreement with the reference MCX results within 1%, and the maximal relative difference of the neutron flux is merely 3%, demonstrating the correctness of the calculation model; (2) while the current DFLL-TBM design meets the operation shielding requirement of ITER for 4 years, it does not satisfy the tritium self-sufficiency requirement. Compared to the two-step approach, Hydra produces higher accuracies as it does not rely on the homogenization technique during the calculation process. The parallel efficiency tests of Hydra using the DFLL-TBM model also showed that this code maintains a high parallel efficiency on O(100) processors and, as a result, is able to significantly improve computing performance through parallelization. Parameter studies have been carried out by varying the thickness of the beryllium armor layer and the tritium breeding zone to understand the influence of the beryllium layer and breeding zone thickness on tritium breeding performance. This establishes a foundation for further improvement in the tritium production performance of DFLL-TBM.

Characterization of Pb-15.7Li Hydrogen Isotopes Permeation Sensors and Upgrade of Hyper- Quarch Experimental Device

The water-cooled lithium-lead (WCLL) test blanket module (TBM) of ITER and the respective breeding blanket (BB) of DEMO need to be efficiently characterized from the point of view of hydrogen isotopes inventory. An in situ detection is required to guarantee tritium self-sufficiency and to control the radiological hazards toward the external environment. In this context, a reliable hydrogen isotopes permeation sensor is essential both for tritium management at the several steps of the reactor fuel cycle and for monitoring the tritium processing systems, e.g., the tritium extraction system (TES). Hence, the hydrogen isotopes permeation sensors were characterized by protium experiments in a 1:1 scale experimental facility named TRIEX-II, located in ENEA C. R. Brasimone, Italy. The results obtained were used for preliminary validation of the novel hydrogen isotopes transport (HIT) code. Moreover, in order to test the sensor in different operative conditions, a new experimental device named Hyper-Quarch II (Hydrogen Permeation Quartz Chamber) is under development and will be installed in ENEA C. R. Brasimone, on the basis of the experience gained in the past experimental campaigns. This device, characterized by an upgraded quartz chamber and new instrumentation equipment, will be used to test and qualify an advanced hydrogen permeation sensor at a laboratory-scale in both dynamic and equilibrium modes: protium or deuterium will be used instead of tritium to simulate the operative conditions of ITER and DEMO reactors.

Preliminary neutronics analyses of China Dual-Functional Lithium-Lead (DFLL) test blanket module for CFETR

China Fusion Engineering Test Reactor (CFETR), as one of the option fusion reactors in China, has been proposed to fill the gap between ITER and fusion DEMO plant. In order to evaluate the neutronics performance in the framework of National Magnetic Confinement Fusion Science Program of China, the neutronics analyses of the detailed model of CFETR with China dual-functional lithium lead (DFLL) test blanket module (TBM) were carried out with the Super Multi-functional Calculation Program for Nuclear Design Safety Evaluation (SuperMC) and JEFF-3.2 nuclear data library. To meet the self-sufficiency target of CFETR, the tritium breeding ratio has been improved to 1.2. Meanwhile, the nuclear heating result showed a potential for energy multiplication factor improvement. From this previous study, an optimized model was proposed to improve the shielding performance and the results showed that the DFLL-TBM with the optimized model could meet the tritium self-sufficiency target and radiation design requirements.

Modification on the contact model of LiPb and noncondensable gas in RELAP/SCDAPSIM/MOD4.0 and application to LOCA of China DFLL-TBM

In-TBM breeder box coolant leak is one of four reference accidents identified for China Dual-functional Lithium Lead Test Blanket Module (DFLL-TBM), which will result in the pressurization of the TBM LiPb breeding zones and cooling system. In order to analyze In-TBM breeder box coolant leak accident, there is a need to simulate the mixing of liquid metal and non-condensable gas. While the current system safety code RELAP/SCDAPSIM/MOD4.0 which was initially designed to predict the behavior of light water reactor systems is incapable of modeling the mixture of liquid metal fluids and non-condensable gas. This paper first briefly introduce the reason for RELAP/SCDAPS/MOD4.0’s incapability of modeling liquid metal in contact with a non-condensable gas. Then, a solution to solve the problem and the modification of the RELAP/SCDAPSIM/MOD4.0 code is proposed. Several typical problems involving liquid metal in contact with helium were simulated and the results demonstrate the feasibility and validity of the modified RELAP/SCDAPS/MOD4.0 in modeling the mixing of liquid metal and non-condensable gas. Last but not least, the modified RELAP/SCDAPS/MOD4.0 is used for transient analysis of In-TBM breeder box coolant leaks. Since both lead-lithium eutectic alloy and helium serve as TBM coolants, the Lithium Lead Cooling System (LLCS) and the Helium Cooling System (HCS) were modeled to investigate the thermal-hydraulic characteristic of the TBM system and its influence on ITER safety under the accident conditions. Results show that LLCS pressurization during In-TBM breeder box coolant leak is adequately handled by the safety accessories provided in LLCS and the highest temperature of the first wall is far below its melting point during the accident process.

CFD simulation of pressure wave propagation in the helium coolant tube break accident of DFLL-TBM

Dual-functional Lithium Lead Test Blanket Module (DFLL-TBM) was proposed by China for testing in the International Thermonuclear Experimental Reactor (ITER). When an in-TBM helium coolant tube breaks, high pressure helium will discharge into the Pb-Li breeding zones. The pressure shock in the breeding box will threaten the safety of TBM. The main objective of this work was to study the pressure wave propagation in the DFLL-TBM due to the In-box LOCA. CFD simulations and analysis on helium coolant tube break accident were performed in this paper. Six different break locations each with a small and a large break size (9 mm * 9 mm and 9 mm * 27 mm) were considered. Computational results indicate that intense pressure waves spread quickly from the break to the surrounding structure. The maximum pressure during the accident is even higher than the operating pressure of helium and its value is related to the location of the break. The longer distance between the break and the farthest inlet/outlet, the higher the peak pressure. Compared to the small size break accident, the pressure variation in TBM is more intense in large break accident.

Non-linear assessment of critical failure modes in the first wall of the European TBM

The assessment of the different components of the Helium Cooled Pebble Bed (HCPB) and Helium Cooled Lithium Lead (HCLL) Test Blanket Modules (TBMs) based on elastic FE analyses has been performed considering main failure modes and respective structural design criteria. This study has been carried out for the operation at full power and for the so-called Approach 2 cooling that is set up at reduced heat extraction capability for compliance with temperature targets to be relevant to DEMOnstration power plant blanket system. Consequently this situation is the worst condition in terms of cooling performance as a result it reveals that many locations in the First Wall (FW) become critical in the sense that certain criteria are not fulfilled, particularly those for immediate plastic flow localization, progressive deformation (ratcheting) and creep-fatigue interaction. Of course one can better reach code compliance by improving the cooling performance (so called Approach 1) but it will degrade the DEMO relevancy.The critical failure modes are re-assessed performing non-linear analysis using an elasto-viscoplatic model, particularly developed for RAFM steels such as the structural material EUROFER97 by Aktaa & Schmitt, and considering the appropriate design criteria of the inelastic route of RCC-MRx and SDC-IC. Despite the conservatism in these criteria, the results show all-clear signal with respect to immediate plastic flow localization and at least much smaller breaches of the limits with respect to the other failure modes. Reporting details and results of the non-linear failure assessments demonstrates their potential and capability in supporting specific and efficient improvements to the design of highly loaded components. The conservatism in the considered design criteria and ways for its reduction are discussed in addition.

RAMI comparison of two China test blanket modules

Tritium breeding blanket module is one of the most important parts of fusion reactor. One of the main goals of ITER is to test and validate design concepts of tritium breeding blankets. China has proposed two test blanket module (TBM) concepts, including Helium cooled ceramic breeder test blanket module system (HCCB TBS) and the dual functional lithium-lead test blanket module system (DFLL TBS), which will be tested in one of the three equatorial ports of ITER. A detailed comparative study on the design schemes of these two TBSs concepts was performed in this paper. In this study, the RAMI (Reliability, Availability, Maintainability, Inspectability) approach was applied on those two test blanket systems. The modeling and analyzing process were done by the Reliability and Probabilistic Safety Assessment Software System RiskA. In addition, comparative study in the phases of functions, capabilities and efficiencies of tritium breeding, heat removal and coolant purification is also performed. It is found that HCCB TBS is a better choice currently to fulfill the engineering demand of ITER construction for its lower difficulty of technical implementation. However, DFLL TBS possesses a high performance prospect from the viewpoint of long-term ITER development due to its inherent advantages compared to HCCB TBS. Discussions of priorities for future research needed for these two TBS are also proposed in the paper.

Assessment of HCLL-TBM optimum welding sequence scenario to minimize welding distortions

The ITER HCLL-TBM (Helium Cooled Lithium Lead Test Blanket Module) box assembly development implies the welding development of the following components: the Box and the Stiffening Grid (SG) made of vertical and horizontal Stiffening Plates (noted respectively v-SP and h-SP). This multi-chamber box structure in EUROFER97 steel is made of plates cooled by multiple meandering channels where circulates pressurized helium. For the assembly of these components, characterized by numerous multipass welds, Gas Tungsten Arc Welding (GTAW) is envisaged as reference process. Moreover, the TBM has large dimensions and thin plates which makes it very sensitive to welding distortions and is problematic regarding the assembly feasibility and compliance with geometric tolerances. This paper presents the numerical simulation and experimental work performed to optimize the v-SP to box assembly sequence, which is the most critical assembly regarding distortions, in order to minimize welding distortions.One of the technical lock of this study is high calculation times needed for this large component which implies to set up a simplified welding simulation method. The study is composed of three main phases: an experimental-numerical study of a T-joint fillet mock-up GTAW used to develop the preliminary welding procedure and to validate a simplified simulation method, a numerical optimization of the v-SP to box welding sequence via the simplified method, and the experimental application of the optimized v-SP to Box welding sequence on the TBM mock-up. The calculation and comparison of four different v-SP to box welding sequences allowed to identify the best sequence regarding welding distortions and to apply it experimentally.

Experimental and RELAP5-3D results on IELLLO (Integrated European Lead Lithium LOop) operation

The experimental facility IELLLO (Integrated European Lead Lithium LOop) was designed and installed at the ENEA Brasimone Research Centre to support the design of the HCLL TBM (Helium Cooled Lithium Lead Test Blanket Module).This work presents the results of the experimental campaign carried out within the framework of F4E-FPA-372 and which had three main objectives. First, to produce new experimental data for flowing LLE (Lead-Lithium Eutectic) for an analysis of the loop and the characterization of its main components. Then, to evaluate performances of commercial instrumentation as available instrumentation is not designed for use in LLE. Lastly, to use the data for validation of the model developed with the system code RELAP5-3D. The data collected could prove helpful to analyze the behavior of the LLE loop of ITER and DEMO in accidental conditions.The results show that the regenerative countercurrent heat exchanger has an efficiency ranging from 70 to 85%, mainly depending on the LLE mass flow rate. It was verified that the air cooler has the capability to keep the cold part of the loop at 623K, even in the most demanding situation (700rpm and maximum temperature of the hot part). The instrumentation tested showed good accuracy, with the exception of the turbine flow meter. Nevertheless, specific limitations in the upper operative temperatures were found for the LLE direct contact pressure transducer. RELAP5-3D simulations fit very well the associated experimental results achieved.

Preliminary RAMI analysis of DFLL TBS for ITER

ITER is the first fusion machine fully designed to prove the physics and technological basis for next fusion power plants. Among the main technical objectives of ITER is to test and validate design concepts of tritium breeding blankets relevant to the fusion power plants. To achieve this goal, China has proposed the dual functional lithium-lead test blanket module (DFLL TBM) concept design. The DFLL TBM and its associated ancillary system were called DFLL TBS. The DFLL TBS play a key role in next fusion reactor. In order to ensure reliable and available of DFLL TBS, the risk control project of DFLL TBS has been put on the schedule. As the stage of the ITER technical risk control policy, the RAMI (Reliability, Availability, Maintainability, Inspectability) approach was used to control the technical risk of ITER. In this paper, the RAMI approach was performed on the conceptual design of DFLL TBS. A functional breakdown was prepared on DFLL TBS, and the system was divided into 3 main functions and 72 basic functions. Based on the result of functional breakdown of DFLL TBS, the reliability block diagrams were prepared to estimate the reliability and availability of each function under the stipulated operating conditions. The inherent availability of the DFLL TBS expected after implementation of mitigation actions was calculated to be 98.57% over 2 years based on the ITER reliability database. A Failure Modes Effects and Criticality Analysis (FMECA) was performed with criticality charts highlighting the risk level of the different failure modes with regard to their probability of occurrence and their effects on the availability.

Preliminary piping layout and integration of European test blanket modules subsystems in ITER CVCS area

This paper explores a possible integration of some ancillary systems of helium-cooled lithium lead (HCLL) and helium-cooled pebble-bed (HCPB) test blanket modules in ITER CVCS area. Computer-aided design and ergonomics simulation tools have been fundamental not only to define suitable routes for pipes, but also to quickly check for maintainability of equipment and in-line components. In particular, accessibility of equipment and systems has been investigated from the very first stages of the design using digital human models. In some cases, the digital simulations have resulted in changes in the initial space reservations.

Study on three-dimensional heterogeneous calculation of ITER test blanket module with deterministic method

The neutronics analysis on the test blanket module (TBM) has important significance for the ITER device and its related experiment design. Quantities of scoping-type studies and conceptual designs were published by using the Monte Carlo method. However, disadvantages like time consuming make it necessary to develop a new highly efficient method. Hence, a new two-step approach method based on the 3D deterministic method for analyzing the TBM is proposed in this paper. A code package 3DMOC-NSPn was developed. It is mainly composed of three modules, the 3DMOC for generating the homogenization cross section; the LINK code for cross section condensation and the NSPn code for blanket calculation. The detailed flux distribution throughout the whole TBM and the mainly neutronics features, such as TBR, displacement per atom (DPA), helium and hydrogen production rate can be obtained. To validate the numerical approach and the code package, the calculations on China dual functional lithium lead-test blanket module (DFLL-TBM) was performed. The reference results were obtained by using the MCNP code. The numerical results from 3DMOC-NSPn are in good agreement with the references. It indicates that the whole code package is a reliable neutronics analysis tool for the TBM design and evaluation.

Mechanical properties and microstructure evolution of CLAM Steel in tube fabrication and test blanket module assembly

The first wall of the China dual functional lithium lead–test blanket module (DFLL–TBM) will be assembled with China low activation martensitic (CLAM) steel rectangular tubes and plates by hot isostatic pressing (HIP) – diffusion welding. The objective of this study is to evaluate CLAM rectangular tubes and investigate mechanical property and microstructure evolution of CLAM steel in tube fabrication and TBM assembly. In this work, CLAM rectangular tubes with lengths of 1500mm were fabricated, and the dimensional accuracy met the requirement for HIP joining. In the tube fabrication process, the CLAM steel was annealed to improve its ductility. In addition, the anisotropy in mechanical properties and microstructure introduced by tube rolling was eliminated according to the simulation of HIP heat treatment in TBM preparation. The tensile strength of the CLAM tubes with final heat treatment was slightly higher than that of CLAM steel with the published standard heat treatment, while the total elongation was reduced. This revealed that a post-HIP heat treatment was required before the final heat treatment.