ITER屏蔽包层活化分析

作为国际热核聚变实验堆(ITER)的重要部件之一,屏蔽包层承受高强度聚变中子辐照,需要定期更换和维修。当活化的屏蔽包层从ITER托卡马克装置移到热室时,可能会给工作人员造成严重的辐射照射,是ITER大厅和热室屏蔽设计的重要辐射源。文中基于ITER最新中子学分析基准模型和"二步法"停堆剂量计算方法,使用超级蒙特卡罗核计算仿真软件系统SuperMC针对15号屏蔽包层建立精细的中子学模型,并计算分析包层的活化情况及最严重情况下的周围辐射剂量率,并初步应用于ITER赤道窗口室的屏蔽分析。计算结果显示,单个包层周围最大剂量率为350 Sv/hr,当传送小车停留在赤道窗口室内时,窗口室屏蔽门外剂量率高于10 mSv/hr,不足以满足设计要求。     

Radiation facilities for fusion reactor first wall and blanket structural materials development

Present and future irradiation facilities for the study of fusion reactor irradiation damage are reviewed. Present studies are centered on irradiation in accelerator-based neutron sources, fast- and mixed-spectrum fission reactors, and ion accelerators. The accelerator-based neutron sources are used to demonstrate damage equivalence between high-energy neutrons and fission reactor neutrons. Once equivalence is demonstrated, the large volume of test space available in fission reactors can be used to study displacement damage, and in some instances, the effects of high-helium concentrations and the interaction of displacement damage and helium on properties. Ion bombardment can be used to study the mechanisms of damage evolution and the interaction of displacement damage and helium. These techniques are reviewed, and typical results obtained from such studies are examined. Finally, future techniques and facilities for developing damage levels that more closely approach those expected in an operating fusion reactor are discussed.     

Beryllium/ferritic martensitic steel joining for the fabrication of the ITER test blanket module first wall

The first wall of an international thermonuclear experimental reactor (ITER) test blanket module (TBM) is a multilayered component consisting of plasma facing armor and structural materials including the cooling channels. One of the main issues about the R&D on the TBM is to develop the joining technologies for a fabrication of the TBM first wall. The objectives of this study are to optimize the hot isostatic pressing (HIP) conditions and the interlayer combination for the fabrication of beryllium (Be)/ferritic martensitic steel (FMS) joints without a degradation of the mechanical properties of the FMS. Effects of HIP joining conditions including the temperature and interlayer types were investigated. The HIP temperature was selected for the anticipated tempering condition for FMS to avoid a grain coarsening which would deteriorate the mechanical properties of FMS. Several interlayer materials were applied in order to manufacture high strength joints. Be and FMS were joined successfully by the application of a Ti/Cu interlayer and it showed a relatively high bending strength, 257 MPa, among the interlayer types studied. The fracture was caused by a delamination of the reaction layer between FMS and the coated interlayer without a plastic deformation. This paper summarizes the results of a Be/FMS joints manufacturing and an investigation of their properties.     

CFD analysis of flow boiling in the ITER first wall

This paper compares two Computational Fluid Dynamic (CFD) approaches for the analysis of flow boiling inside the first wall (FW) of the International Thermonuclear Experimental Reactor (ITER): (1) the Rohsenow model for nucleate boiling, seamlessly switching to the Volume of Fluid (VOF) approach for film boiling, as available in the commercial CFD code STAR-CCM+, (2) the Bergles–Rohsenow (BR) model, for which we developed a User Defined Function (UDF), implemented in the commercial code FLUENT. The physics of both models is described, and the results with different inlet conditions and heating levels are compared with experimental results obtained at the Efremov Institute, Russia. The performance of both models is compared in terms of accuracy and computational cost.     

ITER中国液态锂铅实验包层模块结构热应力数值模拟

使用有限元程序对中国向国际热核实验堆ITER实验包层工作组提交的双功能锂铅实验包层模块(DFLL-TBM)的两种结构设计方案即双冷LiPb包层DLL和单冷准静态LiPb包层SLL进行热应力数值模拟,在包层结构设计、热工水力学设计和中子学计算基础上,给出包层结构温度场和应力场分布,依据ITER高温结构设计标准,进一步对包层高温部件进行力学性能分析.根据这些模拟结果,分析两种结构基本设计方案的合理性和可行性,并作为进一步优化分析的基础.     

Activation analysis of coolant water in ITER blanket and divertor

Coolant water in blankets and divertor cassettes will be activated by neutrons during ITER operation. ¹⁶N and ¹⁷N are determined to be the most important activation products in the coolant water in terms of their impact on ITER design and performance. In this study, the geometry of cooling channels in blanket module 4 was described precisely in the ITER neutronics model ‘Alite-4’ based on the latest CAD model converted using MCAM developed by FDS Team. The ¹⁶N and ¹⁷N concentration distribution in the blanket, divertor cassette and their primary heat transport systems were calculated by MCNP with data library FENDL2.1. The activation of cooling pipes induced ¹⁷N decay neutrons was analyzed and compared with that induced by fusion neutrons, using FISPACT-2007 with data library EAF-2007. The outlet concentration of blanket and divertor cooling systems was 1.37 × 10¹⁰ nuclide/cm³ and 1.05 × 10¹⁰ nuclide/cm³ of ¹⁶N, 8.93 × 10⁶ nuclide/cm³ and 0.33 × 10⁵ nuclide/cm³ of ¹⁷N. The decay gamma-rays from ¹⁶N in activated water could be a problem for cryogenic equipments inside the cryostat. Near the cryostat, the activation of pipes from ¹⁷N decay neutrons was much lower than that from fusion neutrons.     

Detailed 3-D nuclear analysis of ITER blanket modules

In ITER, the blanket modules (BM) are arranged around the plasma to provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. As a part of the BM design process, nuclear analysis is required to determine the level of nuclear heating, helium production, and radiation damage in the BM. Additionally, nuclear heating in the VV is also important for assessing the BM design. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40-degree partially homogenized ITER global model. The regions analyzed include BM01, the neutral beam injection (NB) region, and the upper port region. For BM01, the results show that He production meets the limit necessary for re-welding, and the VV heating behind BM01 is acceptable. For the NBI region, the VV nuclear heating behind the NB region exceeds the design limit by a factor of two. For the upper port region, the nuclear heating of the VV exceeds the design limit by up to 20%. The results presented in this work are being used to modify the BM design in the cases where limits are exceeded.     

Electrical connectors for blanket modules in ITER

Blanket electrical connectors (E-straps, ES) are low-impedance electrical bridges crossing gaps between blanket modules (BMs) and vacuum vessel (VV). Similar ES are used between two parts on each BM: the first wall panel (FW) and shield block (SB). The main functions of E-straps are to: (a) conduct halo currents intercepting some rows of BM, (b) provide grounding paths for all BMs, and (c) operate as electrical shunts which protect water cooling pipes (branch pipes) from excessive halo and eddy currents. E-straps should be elastic enough to absorb 3-D imposed displacements of BM relative VV in a scale of ±2 mm and at the same time strong enough to not be damaged by EM loads. Each electrical strap is a package of flexible conductive sheets made of CuCrZr bronze. Halo current up to 137 kA and some components of eddy currents do pass through one E-strap for a few tens or hundreds milliseconds during the plasma vertical displacement events (VDE) and disruptions. These currents deposit Joule heat and cause rather high electromagnetic loads in a strong external magnetic field, reaching 9 T. A gradual failure of ES to conduct Halo and Eddy currents with low enough impedance gradually redistributes these currents into branch pipes and cause excessive EM loads. When branch pipes will be bent so much that will touch surrounding structures, the Joule heating in accidental electrical contact spots will cause local melting and may lead to a water leak.The paper presents and compares two design options of E-straps: with L-shaped and Z-shaped elastic elements. The latter option was developed in 2012 on the basis of more thoughtful analysis of bi-directional cyclic loading conditions influencing a fatigue lifetime. Detail comparative simulations of current and field patterns and subsequent analysis of the fatigue strength and technological assessment allowed make a final choice for the E-strap design in ITER.     

The design of the ITER first wall panels

The ITER blanket is in the final stage of design completion. The issues raised during the 2007 ITER design review about the first wall (FW) heat loads and remote handling strategy have been addressed, while integrating the recently confirmed in-vessel coils. This paper focuses on the FW design, which is nearing completion. Key design justifications are presented, followed by a summary of the current status of the manufacturing plan and R&D activities.     

ITER中国液态锂铅实验包层模块中子学分析

基于ITER装置全模型,借助于MCNP自动建模程序MCAM,将TBM模块插入该模型的赤道窗口,使用MCNP/4C和FENDL1.0数据库,对DLL和SLL两个典型子模块设计进行三维中子学计算和分析,给出TBM模块核热功率密度分布以及氚增殖能力.     

Russian development of enhanced heat flux technologies for ITER first wall

Recently the ITER first wall (FW) design has been significantly upgraded to improve resistance to electromagnetic loads, to facilitate FW panel replacement and to increase FW ability to withstand higher (up to 5 MW/m²) surface heat loads. The latter has made it necessary to re-employ technologies previously developed for the now-abandoned port limiters. These solutions are related to the cooling channel with CuCrZr–SS bimetallic walls and hypervapotron type cooling regime, optimization of Be-tiles dimensions and Be to CuCrZr joining technique. A number of representative mockups were tested at high heat flux (HHF) at the Tsefey electron-beam facility to verify the thermo-hydraulic characteristics of the reference cooling channel design at moderate water flow velocities (V = 1–3 m/s, P = 2–3 MPa, T = 110–170 °C). The heat flux was gradually varied in the range of 1–10 MW/m² until the critical heat flux was registered. The mockups of hypervapotron structure demonstrated the required cooling efficiency and critical heat flux margin (1.4) at a water velocity of ≥2 m/s. Dimensions of Be armor tiles strongly affect the thermo-mechanical stresses both in the CuCrZr cooling wall and at the Be–CuCrZr interface. Results of tile dimensions optimization (variable in the range 12 mm × 12 mm × 6 to 50 mm × 50 mm × 8 mm) obtained by the HHF (variable in the range of 3–8 MW/m²) experiments are presented and compared with analysis. It is shown that optimization of the tile geometry and joining technology provides the required cyclic fatigue lifetime of the reference FW design.     

CFETR第一壁及赤道面外包层中子辐照损伤初步分析

中国聚变工程实验堆(Chinese Fusion Engineering Testing Reactor,CFETR)的包层和偏滤器第一壁面向堆芯等离子体,第一壁辐照损伤分析对于托克马克安全运行至关重要。赤道面外包层较其它包层距离堆芯等离子体中心更近,其结构材料承受中子辐照大。因此,进行中子辐照损伤评估十分必要。基于此目的,采用计算机辅助设计(Computer Aided Design,CAD)模型和蒙特卡罗中子学建模转换接口Mc CAD完成中子学建模,并用蒙特卡罗方法的粒子输运程序计算第一壁和氦冷固态外包层结构材料辐照损伤。此外,对比了铍和钨作为面向等离子体材料两种情况下第一壁的受损情况。计算结果表明,氦冷固态包层模型下结构材料可以满足CFETR一期的运行要求。     

ITER中国液态锂铅实验包层模块热工水力学设计与分析

依据ITER堆芯物理参数和中子学计算结果,给出了双功能锂铅实验包层模块热工水力学设计方案,并对氦气系统和液态金属锂铅系统压降和驱动功率进行了初步计算.相关中子学、结构、安全及MHD模拟结果证实了设计方案的可行性.     

ITER中国液态锂铅实验包层模块活化特性分析与废料处理

使用中子学程序系统VisualBUS和活化数据库EAF-99对DFLL-TBM的高级子模块DLL-TBM的活化特性进行了计算和分析,包括DLL-TBM各部件在不同停堆时间的活度、衰变余热和剂量率.活化计算所需要的三维中子能谱通过MCNP/4C中子/光子输运程序和国际原子能机构发布的FEND1.0数据库计算得到.在活化计算分析的基础上,参照欧洲聚变堆安全和环境评估(SEAFP)策略中有关核废料的处理标准评估了TBM各区材料在退役后的废料处理工作,包括核废料应该采用何种适当的方式进行处理及其被完全清除干净的可行性.     

Fabrication of Be/CuCrZr/SS mock-ups for ITER first wall

Three types of the mock-ups consisting of Be, CuCrZr and stainless steel (SS) were fabricated by a HIP joining method to demonstrate the fabricability of the ITER first wall. The effects of the mock-up design and the interlayer type on the joining strength of the Be/CuCrZr joint were investigated on the basis of the shear test results for the Be/CuCrZr joint specimens. The joining strength of the Be/CuCrZr ranged from 78 MPa to 204 MPa depending on the types of mock-ups and interlayers. The highest shear strength was obtained from the Be/CuCrZr joint specimens which were taken from the Be/CuCrZr/SS mock-ups with three 80 mm × 80 mm × 10 mm Be tiles and a Cr/Cu interlayer. The effect of the size and the number of Be tiles used in the mock-ups was not as significant as the interlayer selection. Even though the shear test is considered to be very useful to elucidate which interlayer is the most applicable for the joining of Be and CuCrZr, the final interlayer selection should be based on the results obtained from the high heat flux test for the mock-ups fabricated in this work.     

CFD analysis of the ITER first wall 06 panel. Part II: Thermal-hydraulics

The computational fluid dynamics (CFD) analysis of the FW06 panel of the ITER shielding blanket is presented in two companion papers. In this Part II we concentrate on the thermal-hydraulics of the water coolant, driven by the nuclear volumetric and plasma surface heat loads discussed in Part I. Both the detailed steady state analysis of a single cooling channel and the coarse transient analysis of the whole panel are considered. The compatibility of the hot spots with the maximum recommended temperatures for the different materials is confirmed. The heat transfer coefficient between coolant and walls is obtained post-processing the results of the simulation and compared with the results of available correlations, which may be used for simpler analyses: in the fully developed flow regions of the cooling pipes, it turns out to be well approximated by the Sieder–Tate correlation. The operation margin with respect to the critical heat flux is also computed and turns out to be sufficiently large compared with the design limit.     

Thermo-hydraulic design verification of the neutral beam liner for the ITER vacuum vessel

The ITER vacuum vessel has upper, equatorial and lower port structures used for equipment installation, utility feedthroughs, vacuum pumping and access inside the vessel for maintenance. A neutral beam (NB) port of equatorial ports provides a path of neutral beam for plasma heating and current drive. An internal duct liner is built in the NB ports, and copper alloy panels are placed in the top and bottom of the liner to protect duct surface from high-power heat loads. Global NB liner models for the upper panel and the lower panel have been developed, and flow field and conjugate heat transfer analyses have been performed to find out pressure drop and heat transfer characteristics of the liner. Heat loads such as NB power, volumetric heating and surface heat flux are applied in the analyses for beam steering and misalignment conditions. For the upper panel, it is found that unbalanced flow distribution occurs in each flow path, and this causes poor heat transfer performance as well. In order to improve flow distribution and to reduce pressure losses, hydraulic analyses for modified cooling path schemes have been carried out, and design recommendations are made based on the analysis results. For the lower panel, local flow distributions and pressure drop values at each header and branch of the tube are obtained by applying design coolant flow rate. Together with the coolant flow field, temperature and heat transfer coefficient distributions are also acquired from the analyses. Based on the analysis results, it is concluded that the lower panel for the NB liner is relatively well designed even though the given heat loads are very severe.     

Fabrication and high heat flux test of large mockups for ITER first wall semi-prototype

The main topic of an ITER blanket first wall procurement is to qualify whether each party has the key technology needed for the fabrication and joining of first wall components. A semi-prototype qualification project will be released requiring that the single components of a full-scale first wall must be fabricated and successfully pass high heat flux tests using a hypervapotron cooling channel. In this work, various mockup types have been modeled and fabricated to develop the joining technology for a semi-prototype. The semi-prototype, which has three double-fingered panels, is a scaled-down component of a full-size first wall. The standard or slit mockups with a 80 mm × 80 mm single beryllium tile joined to a CuCrZr heat sink were fabricated to qualify our HIP (Hot Isostatic Pressing) technology for the joining of semi-prototype. These standard mockups were installed to perform a high heat flux test in the Korea heat load test facility (KoHLT). For a preliminary test of a semi-prototype, thermo-hydraulic mockups of 710 mm × 100 mm were designed and fabricated to verify the Cu/SS cooling performance, such as hypervapotron. For the high heat flux test in our KoHLT facility, the normal cycle is based on an expected heat flux of 300 s in accordance with the ITER qualification specifications. These tests will be performed to qualify the joining technologies, which is required for an ITER blanket first wall and a semi-prototype.     

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.     

Validation of R5 assessment procedure for ITER test blanket module by finite element analysis

Part of the ITER experimental setup will be testing of the first wall blanket module built from EUROFER 97 steel. Lifetime of this component should be predicted with help of defect assessment procedure. This work verifies the application of the R5 assessment code on the 2D test blanket module geometry by using finite element simulations. Verification is based on the evolution of C(t) parameter. Results exhibit good correspondence in predictions provided by R5 and finite element method for different thermo-mechanical loading conditions. These results therefore show applicability of R5 procedure on such complex geometries.