聚变实验堆ITER实验包层瞬态热工安全分析

在中国向ITER(International Thermonuclear Experiment Reactor)实验包层工作组提交的双功能锂铅实验包层模块(DFLL-TBM)设计分析的基础上,通过对DFLL-TBM系统相关的瞬态事故如真空室内部冷却剂泄漏、TBM(实验包层模块)内部冷却剂泄漏以及真空室外部冷却剂泄漏事故进行计算分析,评价DFLL-TBM对ITER在热工方面对安全的影响.结果表明:当发生瞬态事故时,DFLL-TBM有能力通过热辐射将余热排出,且包层结构不会熔化.DFLL-TBM可满足ITER在热工方面对安全的要求.     

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.     

Preliminary thermal-hydraulic and safety analysis of China DFLL-TBM system

China has proposed the dual-functional lithium-lead (DFLL) tritium breeding blanket concept for testing in ITER as a test blanket module (TBM), to demonstrate the technologies of tritium self-sufficiency, high-grade heat extraction and efficient electricity production which are needed for DEMO and fusion power plant. Safety assessment of the TBM and its auxiliary system should be conducted to deal with ITER safety issues directly caused by the TBM system failure during the design process. In this work, three potential initial events (PIEs) – in-vessel loss of helium (He) coolant and ex-vessel loss of He coolant and loss of flow without scram (LOFWS) – were analyzed for the TBM system with a modified version of the RELAP5/MOD3 code containing liquid lithium-lead eutectic (LiPb). The code also comprised an empirical expression for MHD pressure drop relevant to three-dimensional (3D) effect, the Lubarsky–Kaufman convective heat transfer correlation for LiPb flow and the Gnielinski convective heat transfer correlation for He flow. Since both LiPb and He serve as TBM coolants, the LiPb and He ancillary cooling systems were modeled to investigate the thermal-hydraulic characteristic of the TBM system and its influence on ITER safety under those accident conditions. The TBM components and the coolants flow within the TBM were simulated with one-dimensional heat structures and their associated hydrodynamic components. ITER enclosures including vacuum vessel (VV), port cell and TCWS vault were also covered in the model for accident analyses. Through this best estimate approach, the calculation indicated that the current design of DFLL-TBM and its auxiliary system meets the thermal-hydraulic and safety requirements from ITER.     

Design of a multipurpose laboratory scale apparatus for the investigation of hydrogen isotopes in PbLi and permeation technologies

The knowledge of the tritium transport parameters in lead lithium is fundamental for the design of the HCLL (helium cooled lead lithium) blanket. In fact, the inventory of tritium in fusion reactors blankets and the permeation of tritium into the blanket coolant, with the consequent leaks toward the environment, are strongly depending on its solubility and diffusivity in the lead alloy PbLi. Several experiments, devoted to investigate the function linking the tritium solubilised in lead lithium with the corresponding tritium partial pressure at equilibrium, were carried out in the past, but significant uncertainties still remain.A detailed analysis of the past experimental works is carried out in this paper with the aim to investigate the main problems occurred in the facilities used to measure the tritium solubility in PbLi that caused such a big spread in the achieved results. On the basis of this analysis, a new a multipurpose laboratory scale apparatus has been designed. The apparatus is able to measure the tritium solubility and diffusivity in PbLi in the range of temperature 300–550 °C and it will be operated with hydrogen partial pressure in the range 10²–10⁴ Pa. The facility can work with desorption and absorption technique.Moreover, the apparatus has been designed to allow the testing of H/D concentration sensors in Pb–15.7Li in operative conditions relevant to the HCLL–TBM and the characterisation of hydrogen permeation barrier.     

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.     

Design and development of hydrogen isotope sensor in liquid PbLi

Indian LLCB – TBM uses liquid Lead-Lithium (Pb-Li) as tritium breeder, neutron multiplier and coolant. Tritium bred in liquid PbLi stream has to be recovered by tritium extraction system. Therefore, a reliable sensor with quick response time for measurement of hydrogen isotope is essential.A hydrogen isotope sensor in liquid Pb-Li, based on permeation of hydrogen isotopes through metal (sensor material) is designed. The capsule shaped sensor, made of iron membrane coated with Pd from inside (downstream side), allow hydrogen isotope to permeate through it. The design work mainly includes the selection of proper material, its thickness and surface conditions, which is to be supported by numerical calculations for optimization of maximum permeation flux, fast response and fabricability. The numerical calculation utilizes a physical model having recombination of two hydrogen isotope atoms at the surface and atomic diffusion through the bulk. In this work, design calculations based on numerical simulation and fabrication procedure of the hydrogen isotope sensor are presented.     

Design and R&D progress of Korean HCCR TBM

Korea plans to test a Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in ITER. The HCCR TBM adopts a four sub-module concept considering the fabricability and the transfer of irradiated TBM for post irradiation examination. Each sub-module has seven-layer breeding zone, including three neutron multiplier layers packed with beryllium pebbles, three lithium ceramic pebble bed packed tritium breeder layers, and a reflector layer packed with graphite pebbles. Based on this configuration, neutronic and electromagnetic calculations were performed and their results were applied for the conceptual design of HCCR TBM that considers manufacturing feasibility. Also, a design and safety analysis of HCCR Test Blanket System (TBS) was performed using integrated design tools modifying nuclear system codes for helium coolant and tritium behavior evaluation. The Advanced Reduced Activation Alloy (ARAA) is being developed as a structural material. A total of 73 candidate ARAA alloys were designed and their out-of-pile performance was evaluated. The graphite pebbles as the neutron reflector were fabricated by using mechanical machining and grounding method with the surface coated with SiC. The hydrogen permeation characteristics of structural materials were evaluated using the Hydrogen PERmeation (HYPER) facility. The recent design and R&D progress on these areas are addressed in this paper.     

A tritium permeation model for conceptual fusion reactor designs

A transient tritium permeation model is developed based on a simplified conceptual DT-fueled fusion reactor design. The major design features described in the model are a solid breeder blanket, a low pressure purge gas in the blanket, and a high pressure helium primary coolant. Tritium inventory in the breeder is considered to be due to diffusive hold-up and solubility effects. It is assumed that diffusive hold-up is the dominant factor in order to separate the solution for the breeder tritium concentration. The model was applied to the STARFIRE-Interim Reference Design, whose system parameters yielded a breeder tritium inventory on the order of grams, based on an average pellet radius of 10^?3 cm. The breeder pellets reach their steady-state tritium content in approximately 1.4×10⁴ s from system start-up, assuming continuous full power operation. Both the steady-state breeder tritium concentration and the time to reach that steady-state are proportional to the pellet radius squared. Other candidate solid breeders were considered, and their effect on the blanket tritium inventory was noted. The addition of oxygen to the primary coolant loop was required in order to keep the tritium losses through the heat exchanger to within the design goal of 0.1 Ci/day.     

Numerical Study on Tritium Behavior by Using Isotope Exchange Reactions in Thermochemical Water-Splitting Iodine—Sulfur Process

One potential problem in the hydrogen production system coupled with the high-temperature gascooled reactor (HTGR) is transmission of tritium from the primary coolant to the product hydrogen by permeation through the heat transfer tubes. Tritium accumulation in the process chemicals in the components of a hydrogen plant, a thermochemical water-splitting iodine-sulfur (IS) process, will also be a critical issue in seeking to license the hydrogen plant as a non-nuclear plant in the future. A numerical analysis model for tritium behavior in the IS process was developed by considering the isotope exchange reactions between tritium and the hydrogen-containing process chemicals, i.e., H₂O, H₂SO₄ and HI. The tritium activity concentration in the IS process coupled with the high-temperature engineering test reactor (HTTR), the HTTR-IS system, was preliminarily evaluated in regard to the effects of some indeterminate parameters, i.e., equilibrium constants of the isotope exchange reactions, permeability of tritium through heat transfer tubes, tritium and hydrogen concentrations in the secondary helium coolant, and the leak rate from the secondary coolant loop. The results describing how the tritium activity concentration changes with variations in these parameters and which component has the maximum tritium activity concentration in the IS process are described in this paper.     

Estimation of the Tritium Behavior in the Pebble Type Gas Cooled Reactor for Hydrogen Production

In Korea, a nuclear hydrogen program has been established to develop and demonstrate mass production system for hydrogen generation. The objective of this study is to establish the evaluation procedure for predicting the tritium behavior in the 300 MWth Pebble type gas cooled reactor which is the one of the candidate reactors for nuclear hydrogen development and demonstration plant. The tritium generated by the fission reaction can be leaked to the helium coolant from the coated ceramic particles and fuel elements. The annual total release rate of the tritium is estimated as 0.47% from the fuel kernel to the helium coolant by the numerical method. Tritium attributed by ⁶Li existing as impurities in the reflector can be released to the helium coolant by the diffusion process and the total annual release rate of the tritium is estimated as 5.3% through the reflector to the helium coolant. Based on the Siverts' law, tritium permeation from the primary coolant to the hydrogen production system is also evaluated and the result is calculated as 76?0.23 Bq/g-H₂ with respect to the PRF (Permeation Reduction Factor= 10?1000) in case of the normal operation of the 300 MWth Pebble type reactor.     

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.     

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.     

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.     

Experimental study of efficiency of natural oxide layers for reduction of tritium permeation through Eurofer 97

Tritium permeation from the breeder through the helium coolant is a fundamental safety issue in the design of the HCLL (Helium coolant lithium lead) blanket system. The permeation of hydrogen isotopes through Eurofer in different conditions was deeply studied in the past, demonstrating that it is necessary to reduce this amount using tritium permeation barriers (TPB). A strong effort has been made to select the best technological solution for the realisation of tritium permeation barriers on complex structures not directly accessible after the completion of the manufacturing process, but after many years of activity for the qualification of different materials acting as TPBs it was demonstrated that these technologies are not yet mature for nuclear applications. An easier solution was identified in the nucleation and growth of natural oxides on the helium-exposed surface of cooling system components. The major objective of this work is the evaluation of the Permeation Reduction Factor (PRF) of natural oxides on Eurofer steel adding a known content of water and hydrogen to argon, used in substitution of helium. The PRF was measured on disk shaped specimens, in gas phase, using the PERI II apparatus, at a temperature of 550 °C. The oxide layer was produced in situ, in a well-defined range of hydrogen on water ratios. The obtained results are presented and discussed.     

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.     

压水堆主冷却剂中氚源项计算分析

压水堆主回路冷却剂流经堆芯时,水中固有及特加核素受中子辐照后会产生氚,氚几乎全部以气体和液体的形式排入环境,造成氚污染。因此,氚是压水堆辐射环境影响评价的主要关注内容之一。本文以AP1000为例,根据压水堆主回路冷却剂中氚的产生途径及其随时间的变化情况建立详细的计算模型,计算压水堆主回路冷却剂中的氚活度并分析各产氚途径对氚产生量的贡献。计算结果表明：主回路冷却剂中的氚主要来源于可溶性硼的中子活化和铀裂变,对氚产生量的贡献达80%以上;在⁷Li纯度为99.9%时,AP1000主回路中的年产氚量为5.23×10¹³ Bq,锂产氚量占总量的14.01%,随⁷Li纯度的增加,锂产氚量的贡献呈线性减小,在⁷Li纯度为99.99%时,锂产氚量占总量的3.18%。其他途径对氚的产生量贡献很小,可忽略。根据以上结果,可通过控制主回路冷却剂中添加的初始硼浓度、提高燃料包壳质量、增加LiOH中⁷Li的纯度等多种途径来降低主冷却剂中氚的产生量,从而减少氚对环境的放射性污染。     

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.     

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.     

ITER中国液态锂铅实验包层模块氦气冷却系统初步设计研究

氦气冷却系统是ITER中国液态锂铅实验包层模块(DFLL-TBM)在ITER装置内进行实验的重要辅助系统.根据ITER运行时的热工条件、安全要求、空间要求,分析了DFLL-TBM氦气冷却系统的功能,确定氦气冷却系统的设计原则和要求,在此基础上给出氦气冷却系统的初步设计方案和设备布置.该氦气系统的特点体现在:双功能,即有宽的裕量满足SLL-TBM和DLL-TBM实验;两条氦气回路共享压力控制单元和氦气净化子系统;旁路设计调节TBM和热交换器氦气的出口温度.     

Conceptual design of Tritium Extraction System for the European HCPB Test Blanket Module

The HCPB (Helium Cooled Pebble Bed) Test Blanket Module (TBM), developed in EU to be tested in ITER, adopts a ceramic containing lithium as breeder material, beryllium as neutron multiplier and helium at 80 bar as primary coolant.In HCPB-TBM the main function of Tritium Extraction System (TES) is to extract tritium from the breeder by gas purging, to remove it from the purge gas and to route it to the ITER Tritium Plant for the final tritium processing.In this paper, starting from a revision of the so far reference process considered for HCPB-TES and considering a new modeling activity aimed to evaluate tritium concentration in purge gas, an updated conceptual design of TES is reported.