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.     

基于RELAP5的双功能液态锂铅实验包层模块安全分析

利用嵌入了液态锂铅（LiPb）的热工水力子模块的系统程序RELAP5/MOD3,对双功能液态锂铅（DFLL）实验包层模块（TBM）的安全特性进行评价。对DFLL-TBM及其辅助冷却系统的稳态运行工况、预期运行事件和相关事故工况进行了建模、计算和分析。计算结果表明,稳态运行时第一壁（FW）结构材料表面最高温度低于允许值550 ℃。事故工况下氦气泄漏引起的ITER真空室（VV）、窗口设备室（port cell）以及托卡马克冷却水系统大厅拱顶（TCWS vault）的增压均低于ITER要求的限值0.2 MPa。实验包层钢结构不会熔化且可通过辐射换热有效地导出衰变余热。DFLL-TBM的设计可满足ITER对其热工水力安全方面的要求。     

LOCA analysis for Korean helium cooled solid breeder TBM

One of the major ITER goals is test blanket module (TBM) program which is for the demonstration of the breeding capability that would lead to tritium self-sufficiency in a reactor and the extraction of high-grade heat suitable for electricity generation under the ITER fusion environment. While the engineering design of Korean helium cooled solid breeder (HCSB) TBM and its ancillary systems has been performed, a safety assessment on different possible accident scenarios should be carried out for the purpose of licensing. In this paper, accident analyses for several loss of coolant accident (LOCA) cases were performed in order to assess safety aspects of the TBM design using RELAP5/MOD3.2. Since the TBM forms a loop with helium cooling system (HCS) which is one of ancillary systems required for removing heat deposited in the TBM by neutron wall loading and surface heat flux from plasma, it is necessary to model the complete loop for accident analysis. In this study, the helium passage including the TBM and HCS was nodalized for each accident scenario. The TBM and HCS components were modeled as the associated heat structures provided by RELAP5 to include heat transfer across solid boundaries. Based on computational results it was found that current design of the TBM is robust from the safety point of view.     

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.     

Development of TSACO and application to Chinese HCCB TBM cooling system

In this study, a thermal-hydraulic and safety analysis code (TSACO) for helium cooling system has been developed using Fortran 90 language, and the simulation has been performed for the cooling system of the Chinese helium cooled ceramic breeder test blanket module (CH HCCB TBM). The semi-implicit finite difference technique was adopted for the solution of the dynamic behavior of helium cooling system. Furthermore, a detailed illustration of the numerical solution for heat structures and critical model was presented. The code was verified by the comparison of RELAP5 code with the same initial condition, boundary condition, heat transfer and flow friction models. The TBM inlet/outlet temperatures and pressure drop were obtained and the results simulated by TSACO were shown in good agreement with those by RELAP5. Thereafter, the design basis accident in-vessel loss of coolant accident (LOCA), was investigated for the CH HCCB TBM cooling system. The critical flow model was also verified by comparing with RELAP5 code. The results indicated that the TBM can be cooled down effectively. The vacuum vessel (VV) pressure and the mass of helium spilled into the VV maintained below the design limits with a large margin.     

聚变驱动次临界堆双冷嬗变包层热工水力学参数设计与分析

聚变驱动次临界堆双冷嬗变包层是一个以氦气和液态金属LiPb为冷却剂,以嬗变核废料为主要目的的多功能包层。依据功率平衡模型对不同工况优化的基础上,对该包层热工系统参数进行了设计分析。采用三维商用计算流体力学程序对第一壁和高功率密度区中液态LiPb的流场进行数值模拟计算,给出了优化的典型热工水力参数。     

基于RELAP5的中国氦冷固态包层真空室外破口瞬态特性分析

利用RELAP5/MOD3.4对中国氦冷固态包层、氦气冷却剂回路和二次侧水冷系统进行建模和系统热工水力安全评价。依据ITER事故分析制定的事故序列,对设计基准真空室外破口进行了瞬态分析,并对比了不同破口位置、面积和停堆方式对第一壁的影响。结果表明：真空室外破口发生在风机的下游较上游危险,且小破口较大破口更危险;若真空室外破口同时包层第一壁破口,也可通过自然循环和辐射换热带走衰变热冷却包层;真空室外破口事故中采用聚变停堆系统的3s停堆方式,可避免第一壁熔化。     

聚变制氢堆高温液态包层热工水力学新概念研究

在深入分析聚变堆包层设计要求和目前技术发展水平的基础上,根据热化学工艺制氢需要高温热的要求,提出了一个基于技术相对成熟的低活化铁素体/马氏体钢作为主要结构材料、高压氦气与液态LiPb合金作为冷却剂、具有创新性“多层流道插件”结构方案以获得高温热能的包层热工水力学概念,建立了热工水力学模型,在利用有限元数值模拟程序进行模拟计算的基础上分析了这种新概念包层的可行性。     

Analysis of the reference accidental sequence for safety assessment of LLCB TBM system

Safety analysis of the reference accidental sequence has been carried out for Lead Lithium cooled Ceramic Breeder (LLCB) Test Blanket Module (TBM) system; India's prototype of DEMO blanket concept for testing in International Thermonuclear Experimental Reactor (ITER). The accidental event analyzed starts with a Postulated Initiating Event (PIE) of ex-vessel loss of first wall helium coolant due to guillotine rupture of coolant pipe with simultaneous assumed failure of plasma shutdown system. Three different variants of the sequences analyzed include simultaneous additional failures of TBM and ITER first wall, failure of TBM box resulting in to spilling of lead lithium liquid metal in to vacuum vessel and reactor trip on Loss of Coolant Accident (LOCA) signal from TBM system. The analysis address specific reactor safety concerns, such as pressurization of confinement buildings, vacuum vessel pressurization, release of activated products and tritium during these accidental events and hydrogen production from chemical reactions between lead–lithium liquid metal and beryllium with water. An in-house customized computer code is developed and through these deterministic safety analyses the prescribed safety limits are shown to be well within limits for Indian LLCB-TBM design and it also meets overall safety goal for ITER. This paper reports transient analysis results of the safety assessment.     

Investigation into the in-box LOCA consequence and structural integrity of the KO HCCR TBM in ITER

Korea has developed a Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) and its auxiliary system in ITER. In parallel with its design, safety analysis has performed including accident analysis with the selected reference accidents. Among them, the effect of in-box LOCA to the structural integrity of the TBM was investigated. From the transient analysis of the GAMMA-FR on the in-box LOCA, it is found that the pressure of the internal TBM can be increased up to 8 MPa with the same pressure of the operating coolant through the Tritium Extraction System (TES) and He purge lines in the TBM. Structural analysis with ANSYS code for TBM was performed with this condition and it is confirmed that the TBM can endure and it does not affect the ITER machine by the failure.     

ITER中国液态锂铅实验包层模块氚渗透分析

对ITER中国液态锂铅实验包层模块的氚渗透途径进行了初步分析,并建立了氚渗透模型;在确保环境安全的前提下,通过计算LiPb中的氚分压分析了氚渗透量及氚总量的分配情况;在此基础上通过改变进入氚提取系统中LiPb比例(F)和涂层氚渗透减少因子(TPRF)对氚提取及渗透的影响做了灵敏性分析.     

Status and progress of Indian LLCB test blanket systems for ITER

The lead–lithium ceramic breeder (LLCB) TBM and its auxiliary systems are being developed by India for testing in ITER machine. The LLCB TBM consists of lithium titanate as ceramic breeder (CB) material in the form of packed pebble beds. The FW structural material is ferritic martensitic steel cooled by high-pressure helium gas and lead–lithium eutectic (Pb–Li) flowing separately around the ceramic breeder pebble bed to extract the nuclear heat from the CB zones. Low-pressure helium is purged inside the CB zone for in situ extraction of bred tritium. Currently the LLCB blanket design optimization is under progress. The performance of tritium breeding and high-grade heat extraction is being evaluated by neutronic analysis and thermal–hydraulic calculations for different LLCB cooling configurations and geometrical design variants. The LLCB TBM auxiliary systems such as, helium cooling system (HCS), lead–lithium cooling system (LLCS), tritium extraction system (TES) process design are under progress. Safety analysis of the LLCB test blanket system (TBS) is under progress for the contribution to preliminary safety report of ITER-TBMs. This paper will present the status of the LLCB TBM design, process integration design (PID) of the auxiliary systems and preliminary safety analysis results.     

Numerical analysis of dual-flow fields in Mini-TBM for Chinese LiPb experimental loop Dragon-IV

It is difficult to measure the detailed dual-flow fields of liquid metal lithium lead (LiPb) and helium gas in Mini-Test Blanket Module (TBM). Three dimensions numerical analysis of the LiPb and helium gas flow and heat transfer in Mini-TBM therefore has been curried out using the Computational Fluid Dynamics (CFD) code FLUENT. The detailed dual-flow fields, which include temperature, velocity, pressure and heat transfer of liquid LiPb and helium gas, are presented to support for the test of Mini-TBM, and to supply more robust database and make a significant joint contribution to the future TBM testing in EAST and ITER, and also optimize and improve the design of Dual Function Lithium Lead TBM (DFLL-TBM) system for ITER.     

Three-dimensional dual-flow fields analysis of the DFLL TBM for ITER

This paper concerns the design calculations and performance evaluation of the Dual Function Lithium Lead Test Blanket Module (DFLL TBM) for ITER. Detailed three-dimensional dual-flow field calculations of helium gas and lithium lead (LiPb) have been performed for the DFLL TBM. The commercial Computational Fluid Dynamics (CFD) code FLUENT based finite volume method Navier–Stokes solver capable of solving conjugate flow and heat transfer between dual-flow field and structure is used. The CFD calculations are conducted directly in the CAD model using the CATIA code that allows preserving the geometrical details. The computational results show that the current TBM design is reasonable under the ITER normal condition. The detailed dual-flow fields, which include temperature, velocity, pressure and heat transfer of liquid LiPb and helium gas, are presented to optimize and improve the design of DFLL TBM system for ITER, and to supply more robust database and make a significant joint contribution to the future TBM testing in EAST and ITER.     

Preliminary safety analysis of ex-vessel LOCA for the European HCPB TBM system

Ex-vessel loss of coolant accident caused by a double-ended pipe break of the helium coolant system inside port cell is considered as one of the most critical accident for the European Helium Cooled Pebble Beds Test Blanket Module (HCPB TBM) system. The resulting rapid helium blow-down causes an immediate block of the TBM cooling, which requires a prompt plasma shutdown. Even after the plasma shutdown the temperature can increase over the design limit and the accident sequence can lead up to a break of the TBM box protection after the failure of different protection systems. Thus air ingresses in the vacuum vessel from the damaged TBM system and steam from the surrounding ITER blanket and divertor structures. The evaluation of this sequence is very important for the definition of the correct protection strategy of the system. To consider all these different events a methodology has been developed in KIT combining different codes for a complete analysis of the accident. In particular, this paper shows an application of MELCOR code to model beryllium–steam reaction in a particular accidental sequence for the long term cooling.     

R&D activities of the liquid breeder blanket in Korea

A liquid breeder blanket has been developed in parallel with the International Thermonuclear Experimental Reactor (ITER) Test Blanket Module (TBM) program in Korea. The Korea Atomic Energy Research Institute (KAERI) has developed the common fields of a solid TBM such as design tools, structural material, fabrication methods, and He cooling technology to support this concept for the ITER. Also, other fields such as a liquid breeder technology and tritium extraction have been developed from the designed liquid TBM. For design tools, system codes for safety analysis such as Multi-dimensional Analysis of Reactor Safety (MARS) and GAs Multi-component Mixture Analysis (GAMMA) were developed for He coolant and liquid breeder. For the fabrication methods, Ferritic Martensitic Steel (FMS) to FMS and Be to FMS joinings with a Hot Isostatic Pressing (HIP) were developed and verified with a high heat flux test of up to 0.5–1.0 MW/m². Moreover, three mockups were successfully fabricated and a 10-channel prototype is being fabricated to make a rectangular channel FW. For the integrity of the joining, two high heat flux test facilities were constructed, and one using an electron beam has been constructed. With the 6 MPa nitrogen loop, a basic heat transfer experiment for code validation was performed. From the verification of the components such as preheater and circulator, a 9 MPa He loop was constructed, and it supplies high temperature (500 °C) and pressure (8 MPa) He to the high heat flux test facility. For an electromagnetic (EM) pump development for circulating the liquid breeder, magnetohydrodynamic (MHD) experiment, and flow corrosion test, a PbLi breeder loop was constructed. From the performance test, the EM pump and magnet show their capability, and flow and static corrosion tests including oxide coating for corrosion protection were performed. For tritium extraction from the liquid breeder, a gas–liquid contact method was adopted and a tritium extraction chamber was constructed. For measurement of the tritium amount in the liquid breeder, permeation sensors have been developed.     

中国氦冷固态实验包层2×6模块中子学计算分析

为了满足ITER对波纹度的要求,核工业西南物理研究院提出了新的减少低活化铁素体钢的氦冷固态（HCSB）实验包层模块（TBM）设计方案。采用MCNP程序及ITER全堆MCNP模型,对新设计的2×6HCSB-TBM进行三维中子学计算分析,给出了模块产氚率、核热沉积和功率密度分布等结果。在ITER运行因子为22%时,HCSB-TBM的产氚率为12.68mg/d。TBM内总核热沉积为522.5kW,最高功率密度为11.8W/cm³,出现在氚增殖区Li₄SiO₄中。计算结果可为TBM进一步的结构、热工水力学优化及其他系统设计提供中子学数据。     

Design optimization of first wall and breeder unit module size for the Indian HCCB blanket module

The Indian test blanket module(TBM) program in ITER is one of the major steps in the Indian fusion reactor program for carrying out the RD activities in the critical areas like design of tritium breeding blankets relevant to future Indian fusion devices(ITER relevant and DEMO).The Indian Lead–Lithium Cooled Ceramic Breeder(LLCB) blanket concept is one of the Indian DEMO relevant TBM,to be tested in ITER as a part of the TBM program.Helium-Cooled Ceramic Breeder(HCCB) is an alternative blanket concept that consists of lithium titanate(Li_2TiO_3) as ceramic breeder(CB) material in the form of packed pebble beds and beryllium as the neutron multiplier.Specifically,attentions are given to the optimization of first wall coolant channel design and size of breeder unit module considering coolant pressure and thermal loads for the proposed Indian HCCB blanket based on ITER relevant TBM and loading conditions.These analyses will help proceeding further in designing blankets for loads relevant to the future fusion device.     

Incorporation of lithium lead eutectic as a working fluid in RELAP5 and preliminary safety assessment of LLCS

The current work involves thermal hydraulic calculation of Lithium Lead Cooling System (LLCS) for the Indian test blanket module (TBM) for testing in International Thermonuclear Experimental reactor (ITER). It uses the RELAP portion of RELAP/SCDAPSIM/MOD4.0. Lithium-lead eutectic (LLE) has been used as multiplier, breeder and coolant in TBM. Thermodynamic and transport properties of the LLE have been incorporated into the code. The main focus of this study is to check the heat transfer capability of LLE as coolant for TBM system for steady state and the considered anticipated operational occurrences (AOO's), namely, loss of heat source, loss of primary flow and loss of secondary flow. The six heat transfer correlation (reported for liquid metals in the literature) has been tested for steady state analysis of LLCS loop and results are roughly same for all of them. A good agreement has been observed between the operating conditions of LLCS with those of RELAP5 calculations. Results from transient calculations show that a maximum temperature of 875 K is attained during a 300 s loss of primary flow (LLE).     

New progress on design and R&D for solid breeder test blanket module in China

ITER will be used to test tritium breeding module concepts, which will lead to the design of DEMO fusion reactor demonstrating tritium self-sufficiency and the extraction of high grade heat for electricity production. China plans to test the HCCB TBM modules during different operation phases. Related design and R&D activities for each TBM module with the auxiliary system are introduced.The helium-cooled ceramic breeder (HCCB) test blanket module (TBM) is the primary option of the Chinese TBM program. The preliminary conceptual design of CN HCCB TBM has been completed. A modified design to reduce the RAFM material mass to 1.3 ton has been carried out based on the ITER technical requirement. Basic characteristics and main design parameters of CN HCCB TBM are introduced briefly. The mock-up fabrication and component tests for Chinese test blanket module are being developed. Recent status of the components of CN HCCB TBM and fabrication technology development are also reported. The neutron multiplier Be pebbles, tritium breeder Li₄SiO₄ pebbles, and structure material CLF-1 of ton-class are being prepared in laboratory scale. The fabrication of pebble bed container and experiment of tritium breeder pebble bed will be started soon. The fabrication technology development is proceeding as the large-scale mock-up fabrication enters into the R&D stage and demonstration tests toward TBM testing on ITER test port are being done as scheduled.