Dual-functional Lithium Lead Test Blanket Module Research Articles

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.

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.

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.

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.

Preliminary numerical analysis of MHD flow of DFLL-TBM on the single coolant stage

The liquid lithium–lead (PbLi) breeder blanket concept has been explored extensively due to their potential attractiveness. To check and validate the feasibility, the China dual-functional lithium lead test blanket module (DFLL-TBM) system, which is designated to demonstrate the integrated technologies of both He single coolant (SLL: single-cooled lithium lead) and He–LiPb dual-coolant (DLL: dual-cooled lithium lead) blankets, is proposed for test in ITER. One of the key feasibility issues is the impact of liquid metal MHD effect which will influence the pressure drop, flow distribution, and heat transfer in a DFLL-TBM.To reduce MHD effect, an electrically insulating coating is applied onto the inner surface of the flow channel for single coolant blanket. In this work, a preliminary numerical study of MHD flows in a simplified DFLL-TBM model on the single coolant stage has been carried out to assess the performance of such a concept with regard to the above mentioned MHD problems and constraints. The flow distribution and MHD pressure drop of LiPb flow in the SLL stage TBM are analyzed.

Numerical investigation of buoyant effect on flow and heat transfer of Lithium–Lead Eutectic in DFLL–TBM

Thermal–hydraulic behaviors of Lithium–Lead Eutectic (PbLi) in Chinese ITER Dual Functional Lithium–LeadTest Blanket Module (DFLL–TBM) were investigated numerically by using CFD code ANSYS CFX. The influences of buoyant effect on flow and heat transfer of PbLi were analyzed and the validity of Boussinesq approximation for liquid PbLi with non-uniform heat generation was discussed. Comparisons between two scenarios with and without buoyancy were made and the results shown that the flow field was significantly modified compared with the case of non-buoyancy effect. The local reverse flow phenomena and recirculation flows in the front duct near the first wall (FW) were found. Furthermore, a fractional outflow was also observed at inlet under the condition of inlet mass flow rate 2.165 kg/s. The results indicated that the buoyancy-induced mixed convection occurred in front duct can enhance the heat transfer process and made the temperature field of PbLi more uniform that can improve the operational safety of DFLL–TBM. The present work is expected to be helpful for the designs of DFLL–TBM.

Failure Modes and Effects Analysis on ITER DFLL-TBM system

As required for licensing process, accident analyses of International Thermonuclear Experimental Reactor (ITER) accounting for site specifications and design changes will be updated. Chinese Dual-Functional Lithium-Lead-Test Blanket Module (DFLL-TBM) system is a key safety-related component of ITER, its detailed safety analysis, which was designated to demonstrate the integrated technologies of both Helium single coolant (SLL) blanket and Helium-LiPb dual coolant (DLL) blanket, was performed. Failure Modes and Effects Analysis (FMEA) was applied to perform the safety analysis of DFLL-TBM. This study described the process of FMEA studies on DFLL-TBM system. All safety-related Postulated Initiating Events (PIEs) was identified. And a set of PIEs recommended to be taken into account in the further deterministic transient analyses were defined for both SLL and DLL blanket concepts separately.

Numerical investigation of heat transfer enhancement in ribbed channel for the first wall of DFLL-TBM in ITER

As an important component of Dual Functional Lithium Lead-Test Blanket Module (DFLL-TBM), the first wall (FW) must withstand and remove the heat flux from the plasma (q″=0.3MW/m2) and high nuclear power deposited in the structure at normal plasma operation scenario of ITER. In this paper the transverse ribs arranged along the plasma facing inner wall surface were used to enhance the heat transfer capability. After the validation compared with empirical correlations the Standard k–ω model was employed to do the numerical simulation using FLUENT code to investigate the heat transfer efficiency and flow performance of coolant in the ribbed channel preliminarily. The perforation on the bottom of rib was proposed near the lower heat transfer area (LHTA) to improve the heat transfer performance according to results of analyses.

Updated thermal-mechanical analysis of DFLL-TBM for ITER

The Dual Functional Lithium Lead - Test Blanket Module (DFLL-TBM) was developed by China as one of candidate concepts for ITER. This paper is focused on evaluation of thermal-mechanical performance under integrated load, such as thermal, pressure, and electromagnetic load. The latest 3D finite element model slice for DFLL-TBM was developed. The steady state thermal-mechanical analyses were performed under the normal plasma operational. The primary stress and secondary stress were evaluated according to Structural Design Criteria for ITER In-vessel Components (SDC-IC). It was concluded that the primary plus secondary stress were satisfied with structural material permission under normal operation conditions. The possible recommendations for optimization and analysis of DFLL-TBM are also discussed and proposed.

Effects of tube rolling and heat treatment on microstructure and mechanical properties of CLAM rectangular tube

The first wall (FW) of China dual functional lithium lead-test blanket module (DFLL-TBM) for International Thermonuclear Experimental Reactor (ITER) contains rectangular helium flow channels, which will be fabricated by hot isostatic pressing (HIP)-diffusion welding with China low activation martensitic (CLAM) steel plates and rectangular tubes. Study on the rectangular tube fabrication process are being conducted. Rectangular tubes of 15 mm × 20 (inter dimensions of cross section) mm × 2.2 (wall thickness) mm × 1500 (length) mm were produced by hot piercing and cold rolling, and its dimensional deviation met the requirement of TBM assembly. The effects of tube rolling and heat treatment on microstructure and mechanical properties of CLAM rectangular tube were analyzed. It is found that soft annealing heat treatment after tube piercing was a suitable way to improve the ductility of the CLAM steel. The elongation should be less than 15% at every tube rolling step, and the favorable tempering temperature to release the stress caused by tube rolling was 800 °C.

Analysis of Plasma-Driven Tritium Permeation Through the First Wall of DFLL-TBM in ITER

Tritium permeation through the first wall (FW) from the plasma into helium coolant is evaluated for a dual-functional lithium-lead test blanket module (DFLL-TBM). The effect of the surface conditions on the plasma facing and coolant sides, both temperature gradient and beryllium layer clad on the plasma facing side, as well as trapping in defects on the tritium permeation is considered. The results show that most of the tritium implanted in FW re-entered the plasma. The plasma-driven tritium permeation is very sensitive to the surface conditions on the plasma facing side. With a higher sticking coefficient on the plasma-facing side, the tritium permeation into helium coolant is significantly reduced. The tritium permeation is strongly reduced with a beryllium layer clad on the front side of FW. The plasma driven tritium permeation will not seriously impact the tritium safety of DFLL-TBM. Based on tritium safety, it is reasonable to clothe the beryllium layer on FW and keep the surface clean to reduce the plasma driven tritium permeation.

Preliminary safety analysis for the Chinese ITER Dual Functional Lithium–Lead Test Blanket Module

Safety analysis is part of the ITER Test Blanket Module (TBM) design process ensuring that the TBM does not adversely affect the safety of ITER. To get the licence for TBM as a whole with ITER, relevant safety analysis is required for each TBM system proposed by each party. The safety analysis for the Chinese Dual Functional Lithium–Lead Test Blanket Module (DFLL-TBM) has been performed based on the latest DFLL-TBM design. In this paper, the following safety considerations, such as source terms, operational releases, accident sequence analyses and waste assessment, were analysed. Both the deterministic approach and the complementary systematic approach starting with failure mode and effects analysis studies were adopted in the accidental analysis. The preliminary results showed that the DFLL-TBM system at normal operating conditions and under accident scenarios did not add additional safety hazards to the ITER machine and could meet the ITER safety requirements and additional safety requirements for the TBM system.

Thermo-Mechanical Analyses of the High Heat Flux Component for ITER Dual Functional Lithium Lead Test Blanket Module

The finite element code ANSYS is used to calculate the temperature and stress distributions for the first wall of DFLL-TBM (dual functional lithium lead-test blanket module), for testing in ITER. Preliminary analyses indicate that not only the low temperature design rules, the well-known 3Sm rules, are satisfied for the first wall, but the additional high temperature structural design criteria for the creep damage limits and creep-ratcheting limits are met as well.

Preliminary tritium safety analysis on China DFLL-TBM for ITER

The dual-functional lithium–lead test blanket module (DFLL-TBM) system was proposed to be tested in ITER. A tritium permeation model of the entire DFLL-TBM system was developed, and the tritium permeation and inventory in DFLL-TBM system were done based on the model during normal operation. Three classes of off-normal situations had been preliminarily analyzed, i.e. in-vessel TBM coolant leaks, in-TBM breeder box coolant leaks and ex-vessel TBM ancillary coolant leaks. The results showed that some issues required significant R&D effort to guarantee the tritium release to the environment below the allowable level, such as the tritium extraction from LiPb and helium coolant and very efficient detritiation system. And more analyses would be carried in the future to further assess the safety of DFLL-TBM.

Design analysis of the China dual-functional lithium lead (DFLL) test blanket module in ITER

The dual-functional lithium lead-test blanket module (DFLL-TBM) system, which is designated to check and validate the relevant technologies of both the helium-cooled lithium–lead blanket and the helium/lithium–lead dual-cooled blanket for the Chinese fusion DEMO reactor, has been proposed for testing in ITER referring to the relevant several conceptual designs of liquid lithium–lead breeder blankets developed in China. The detailed description of the basic structure, cooling scheme and auxiliary systems for DFLL-TBM has been given in this paper. A series of preliminary performance analyses have been performed to validate the feasibility of the DFLL-TBM design. Preliminary results show that TBM design can satisfy the material requirements and ITER safety design requirements.

Neutronics analysis for the test blanket modules proposed for EAST and ITER

The dual-functional lithium lead-test blanket module (DFLL-TBM) system, which is designated to demonstrate the integrated technologies of both the He single coolant (SLL) blanket and the He-LiPb dual coolant (DLL) blanket, is proposed for testing in ITER to check and validate the feasibility of the Chinese LiPb blankets. Considering the conflict between the limited ITER resources for TBM testing and the requirement of various blanket concepts proposed by the parties, EAST, the superconducting tokamak being operated in China, can serve as a valuable pre-testing platform with some comparable design parameters to ITER for DFLL-TBM. In the contribution, neutronics calculations and activation of the TBMs in EAST are carried out and compared with those in the D–D and D–T phase of ITER. The results from EAST can make an analogy to ITER (to some extent).

Design of auxiliary system for the dual functional lithium–lead test blanket module in ITER

LiPb auxiliary system is an important external loop equipment of test blanket module (TBM) for liquid LiPb blanket design. According to the requirement of testing strategy of DFLL (dual functional lithium lead) TBM, which could demonstrate the technology of the SLL (single-coolant lithium lead) and the DLL (dual-coolant lithium lead) blankets, a conceptual design of the LiPb ancillary system is done, which could serve to achieve the dual functions. The auxiliary system can be used to achieve the recovery of tritium, heat rejection, monitoring of impurity in LiPb, LiPb purification and so on. It consists of detritiation unit, LiPb to He heat exchanger, plugging indicator, LiPb purification system and pumping system, etc.

Preliminary leakage reliability analysis of DFLL-TBM based on a combinational approach

Dual-functional lithium lead-test blanket module (DFLL-TBM) is designed to demonstrate and validate related technologies for the high-power density liquid metal blanket of fusion commercial demo reactors (e.g. FDS-II) and the He/LiPb dual-cooled multi-functional blanket of a fusion-driven sub-critical system (i.e. FDS-I). It is important to assess the reliability of DFLL-TBM due to its complexity. A new approach is proposed by combining the probabilistic fracture mechanics (PFM) method and the probabilistic safety assessment (PSA) method. A reliability analysis code for blanket modules is developed. One of the failure modes of DFLL-TBM, leakage failure, is analyzed as an example to demonstrate the capability of the combinational approach. Some key aspects that affect the leakage reliability of DFLL-TBM are identified.