Fusion Reactor Research Activities at SWIP (1)

Transcription

1 Fusion Reactor Research Activities at SWIP (1) Design Studies of DEMO Reactor 1. Introduction Considering that there is still a long way to go towards an economically competitive commercial fusion power plant, thus, the DEMO in Chinese fusion developing strategy should be an indispensable step prior to the commercial one after ITER. The DEMO in China is to demonstrate the safety, reliability and environment feasibility of the fusion power plant, meanwhile to demonstrate the prospective economic feasibility of the commercial fusion power plant. Nowadays, two options of breeding blanket with ceramic and lead lithium breeders have been chosen as China DEMO concepts under the conditions of meeting the requirements of the neutronics, thermal-hydraulics and mechanics aspects. The conceptual design for the ceramic breeder DEMO (HCSB-DEMO) have already been carried out, the three-dimensional scheme of HCSB-DEMO is shown in Fig.1. Fig.1 Conceptual View of HCSB-DEMO 2 Plasma Parameters Under different plasma limited conditions, the fusion power, the neutron wall loading and the plasma burn time in the fusion reactors with different sizes are investigated. In this way, those reactors which fit our requirements are selected, and the POPCON analyses for them are also done. At last, three group plasma parameters for HCSB-DEMO with different limited conditions are given in Table g= g= P fus =2700MW 0MW 20MW MW 60MW 80MW 100MW 140MW P AUX =172MW Fig.2 Plasma operation contour 1

2 The following figure is the POPCON analyses for one of the selected reactor. The following table is the main plasma parameters of HCSB-DEMO. Table 1 The main plasma parameters Fig.3 the equilibrium configurations of the OP1 Parameter OP1 OP2 OP3 R/a (m/m) 9.0/ / /2.1 BT (T) Ip (MA) Volume(m 3 ) Surface(m 2 ) / / / /0.45 N <n e >/n G q <T e >( k e V) P AUX (MW) P FUS (MW) p n (MW m -2 ) H H98 (y,2) * He / E Z eff I bs / I p (%) Q Operating model steady steady steady Using the EFIT code under fixed boundary conditions to reconstructed the equilibrium configurations of the OP1, the figure as following: 3. DEMO Blanket Blanket design is a very important section in developing DEMO. The blanket of HCSB-DEMO is intended to be designed as module structure. The fourteen modules are connected each other along the poloidal direction, and there are all 72 rows of this connection along the toridal direction, so the total number of the modules is about The module mounting/dismounting operations are carried out through the vacuum vessel vertical ports. The radial thickness of inboard and outboard modules is 630mm and 800 mm respectively. On the basis of the design parameters of HCSB-DEMO blanket (listed in Table 2), the concrete design for the blanket modules were performed. On the whole, the couple structure is adopted, that is, the two modules are symmetrically welded to the back-plate which fixed the surface of the vacuum vessel. This design scheme makes the blanket modules compact and firm, and can increase the cooling effect. Table 2 Design Parameters of HCSB-DEMO Blanket Neutron Wall Loading, NWL Max. Surface Heat Load, SHL Tritium Breeding Ratio, TBR MW/m 2 MW/m Coolant He 2

3 Temperature Range, T C Pressure, P MPa Tritium Breeder [% 6 Li enrichment] Packing Factor, PF Temperature Range, T C Neutron Multiplier Temperature Range,T C Structure Material Temperature Range, T C Li 4 SiO 4 [80%] pebbles 63% Be pebbles at 80% PF CLAFM The blanket modules contain the first-wall and the back-plate, between them (cooling plate), there are 13 rows of circular coolant channels, 6 rows of lithium orthosilicate pebble-bed and 8 rows of berylium pebble-bed. The general structure scheme of HCSB-DEMO blanket is shown in Fig.4. (a) Plain Plot Fig.4 Structure Scheme of HCSB-DEMO Blanket (b) 3-D Plot 4. Tchnology R&D The R&D of the relevant technology about DEMO blanket such as material technology (structure material and breeder), tritium technology et al., which are undergoing development by cooperating with other domestic institutions. 3

4 Fusion Reactor Research Activities at SWIP (2) Design-relevant study of blanket liquid curtain Owing to its high heat removal and self-refreshing capability, liquid metal first wall has been seen as a potential first wall for a fusion reactor in the future. Blanket liquid curtain is actually a special liquid metal wall to protect blanket. Study of blanket liquid curtain contains flow analyses of liquid metal, electromagnetic property analyses of liquid, magnetic circuit design, analyses of kinetic effects of fluid and so on. 1. Programming Liquid metal flows MHD effects should include two parts: analyses of flowing properties in a force field and calculations of induced electromagnetic forces with a velocity distribution. Based on this, new code used to analyze properties of blanket liquid curtain has been developed. Fig.1 Loop control diagram of the new calculation code 2. Flowing analyses Now lithium is chosen as liquid material. If there is no external magnetic field, owing to gravity the speed of fluid will increase rapidly. And height of fluid will decrease obviously. With external field 2T perpendicular to sidewalls, owing to hindering action of induced electromagnetic forces, flow speed will decrease. Choosing suitable material properties and dimensions of the duct, we can get that the height of fluid do not vary in the range of our calculation. Fluid flow has two steams: very fast surface layer and behind it slow layer. 4

5 a)with no magnetic field b)with external field 2T Fig.2 velocity distribution Obviously velocity distribution relies on distribution of induced electromagnetic forces. According to the equation F = J B = σ [ φ + V B] B [2,3], induced force is composed of the forces owing to interaction of induced electric field and magnetic field and owing to movement in magnetic field. In the free surface layer, although its velocity is very fast, induced force is very small because of electric field effect. 3. Thermal analyses Layer I of blankety liquid curtain is faced to plasma while Layer II is attached to a blanket.t 0 is the temperature of the lithium at the top of the blanket and T1,T2 are those of the two layers at the distance h from the top. V1,V2 and d1,d2 are speeds and widths of the layers. T c is constant temperature for the blanket. Heat flux coming from plasma is q. To the zones with unit lengths perpendicular to axis x, following equations can be obtained: Q 1 =C p *ρ*1*1*d1*(t1-t 0 ) Q 2 =C p *ρ*1*1*d2*(t2-t 0 ) Q left =q*h/v1 Q right = 0 h ( λ * ( T 2() T c ) / V 2) dh In which, Q left, Q right and Q 1, Q 2 are heat irradiated to lithium and got by lithium respectively. Cp, ρ and λ are specific heat, density and thermal conductivity of liquid lithium. For liquid lithium curtain, if d2=2*d1 and v1=3*v2 then T1+2*T2=286.6h+600in which T2<= h. With T2=400 C and T1>400 C, h=2.09m. It means blanket liquid curtain can effectively remove all of the heat from plasma with up to 5MW/m 2. It will be good to thermal design of blankets. 4. Designing Liquid metal flows down along front of blanket, if only gravity exists, velocity of fluid will becomes bigger and bigger and it will be thinner and thinner. Finally liquid metal cannot cover the blanket and blanket will be exposed in very severe external circumstance. It is lucky that liquid metal is acted not only by gravity but also by magnetic field. Toroidal field exists around first wall and the component of force produced by liquid metal flow in magnetic field will alleviate the influence of gravity. According to the results above, it is possible to make liquid metal flow have a certain height along surface of blanket with suitable design. Based on our works, a preliminary design has been provided. 5

6 Fusion Reactor Research Activities at SWIP (3) High Temperature He Experiment Loop (HTHEL) 1. Introduction CH ITER HCSB TBM program is a long-term activity in SWIP. For the qualification of the whole property of TBM s Helium Cooling System, it is indispensable to build a experimental helium cooling loop under high pressure and temperature profiles, in order to validate design parameters, especially regarding mass flow, temperature change, pressure drop and heat transfer process in narrow cooling channel. In addition, the similar testing must be performed for developing the DEMO blanket in future. For these purposes, a construction plan of experimental helium cooing loop (HTHEL) with high temperature and pressure has been scheduled in China. 2. Requirements The primary requirements defining the nominal operating condition for each of the fusion components that are planed to be tested in China constructing High Temperature & Pressure He Experiment Loop (HTHEL) are presented in Table 1. Test section Table 1 Requirements for test section in HTHEL He mass flow rate (kg/s) Pressure (MPa) Pressure difference (MPa) He inlet/outlet Temperature ( C) ITER-TBM < /500 DEMO Blanket ~4.0 8 < /500 The goals of HTHEL are: To validate the design of the TBM; To test important components and sub-systems (Circulator, HX) of the HCS; To assess reliability and availability of the HCS in ITER; To simulate and examine abnormal operating conditions ( LOFA, LOCA); And to validate the DEMO blanket helium loop at very high mass flow rate. 3. Conceptual Design The primary conceptual design of the piping and instrumentation arrangement for HTHEL is being performed currently. Figure 1 shows the flow diagram of HTHEL and interfaces to ancillary equipment. 6

7 Fig.1 Flow design diagram of HTHEL Fig.2 3-D schematic view of HTHEL 7

8 Fusion Reactor Research Activities at SWIP (4) Neutronics Studies for Fusion Reactor Neutronics is one of the most important areas for design of fusion reactor. It lays the basis on other analysis such as thermal-hydraulics, tritium systems, shield, radioactivity safety, etc. Research of neutronics for fusion reactor has been proceeding for about 30 years in the division of fusion reactor of SWIP. In the past, neutronics for TETB series, TCB and FEB serials and so on have been performed. There are plenty of much valuable research results and conclusions achieved as well as published on domestic and foreign magazines or presented at relevant conferences. Presently, main tasks of neutronics in the division focus on neutronics design for Chinese ITER HCSB TBM. Main fields of current neutronics analysis in the division includes: Calculation of Nuclear Power and Tritium; Radioactivity and Afterheat; Shielding Neutronics; Nuclear Data Library. Main calculation tools include 1) One-dimension neutron transport codes: BISON, ANISN, ONEDANT; 2) Two-dimension neutronics codes: TWODANT, DORT4.2; 3) Three-dimension neutronics codes: NJOY2000, TRANSX2.5. Now, there is a neutronics team consisting of eight persons in field of neutronics research of ITER TBM and DEMO blanket in the division: Fig.1 MCNP Model for HCSB TBM Fig.2 Neutron flux at Be tile, Li 4 SiO 4 of HCSB TBM 8

9 Fusion Reactor Research Activities at SWIP (5) Fusion Reactor Safety Research 1. Introduction As any other nuclear devices, fusion reactor safety research is one of important contents of fusion reactor design and development. Compared to fission reactor, fusion has its own unique safety characteristic and contents, including test strategy of all components in pile or out pile, environment evaluation, radioprotection, emergency preparation, remote handling, waste management, codes development and validation, and data establishment, etc. Series of studies for fusion reactor safety research have been developed for years in SWIP in the framework of various Chinese national programs. Now the reactor safety focuses its programme on the development of China TBMs, DEMO and future commercial fusion reactor. The reactor safety research is an important direction of the fusion reactor research division in SWIP. 2. Key technology At present, the fusion reactor safety research gathers the following aspect: (1) China helium cooled solid breeder (HCSB) test blanket module (TBM) safety analyses Since China became a member of international thermo-nuclear experiment reactor (ITER), we have begun to design and analyze China HCSB TBM, including its safety. The TBM safety analyses focus on resource terms analyses, tritium safety management, accident analysis, waste dispose, QA, codes development and verification, TBM safety report, and so on. Some results have been reported on ITER test blanket work group (TBWG) web. (2) China fusion reactor safety analyses Radioactivity and radiation: Fusion radioactivity research includes swell and hydrogen-embrittlement of fusion structure materials, tritium safety management, radioprotection, and environment effect, etc. For tritium breeder and neutron multiplier, it is necessary to study their thermo-mechanical and radiation stability. Electromagnetic analysis: Electromagnetic stress on components and electromagnetic radiation harm to person are must studied under fusion strong electromagnetic field condition. Plasma rupture: Components safety: System accident: (3) Material safety research The objectives of the fusion reactor materials research are: To evaluate the integrity and behavior of structural materials used in nuclear power industry; To perform research to unravel and understand the parameters that determine the material behavior under or after irradiation; To contribute to the interpretation, the modeling of the materials behavior and to develop and assess strategies for optimum life management of nuclear power plant components. (4) Codes development 9

10 Multi-functional/multi-dimensional Neutronics Analysis Code; Multi-dimensional Neutronics/Thermal-hydraulics Coupling Code; Tritium permeation and cycling simulation code; MHD simulation code; System analysis code. Flow chart of 3-Dimensional activation calculation by MCNP coupling with FDKR (5) R&D for safety analysis Tritium cycle and tritium permeation; Detritiation system; Remote handling facilities; Neutronics integral experiment; Helium test loop (8~10MPa, 500~700 o C); Test platform for high heat flux properties of material; Tritium processing equipments; Tritium emergency-response system. 3. Development programme and technology route Short term: Aim at fusion reactor design and China ITER TBM, to develop fusion reactor safety and protection research; to develop primary safety analysis based on safety analysis report of ITER design and experience of fusion experiment reactor design, including radioactivity calculation, LOCA & LOFA accident analysis, and E-M calculation. Middle term: To primarily master fusion reactor safety analysis approach; to develop environment and shielding analysis; to develop experiment related to safety such as helium cooled loop and material performance experiment, etc.; to develop, transplant and fetch in some safety analysis codes often used in the world for risk probability and reliability analysis. Long term: To have full safety analysis ability without any difficulty; to build perfect system of fusion reactor safety design, technology analysis and safety assessment; to build a set of experiment establishment to provide data for safety analysis; to complete safety design for China DEMO, commercial fusion power plant. Technology route: To mater ITER safety design, safety assessment technology and apply to fusion reactor safety analysis and assessment; to develop necessary related 10

11 experiment and provide data for safety design; to develop China fusion reactor safety design and analysis referring domestic fission reactor safety related technology, China nuclear safety code and ITER related design; to reinforce international cooperation and invite domestic or overseas nuclear safety experts to instruct or participate in our future fusion reactor safety design. 11

12 Fusion Reactor Research Activities SWIP (6) ITER Test Blanket Module (TBM) Program 1. Introduction ITER will play a very important role in first integrated blanket testing in fusion environment. Some of related technologies of DEMO blanket, such as tritium self-sufficiency, the extraction of high-grade heat, design criteria and safety requirements and environmental impacts will be demonstrated in ITER test blanket modules (TBMs). China has planned to develop own the ITER TBM modules for testing during ITER operation period. Although different concepts for ITER TBMs have been proposed by other parties [1]. However, the he-cooled solid breeder (HCSB) blanket with Ferritic /Martensitic steel (FMs) is still main stream for the fusion DEMO blanket and has foundation of the world R&D database. Therefore, a helium-cooled solid pebble bed concept has been adopted, as one of options, in China ITER TBM modules design. The preliminary design and performances analysis as well as a draft Design Description Document (DDD) based on the definition and the strategy of DEMO fusion reactor in China have been carried out recently. Preliminary design and analysis have shown that the proposed TBM module concept is feasible within the existing technologies. 2. Design description The schematic structure and 2-D calculation model of the HC-SB TBM are shown in Figs.1. CH TBM has the following overall sizes: 1660mm (H) 484mm(W) 630mm (D). The low-activation ferritic/martensitic steel (LAFMs) is chosen as design structural material for the first wall and the main components. The HCSB TBM consists of first wall, breeding sub-modules, back plate, caps, grid and support plate. A beryllium armor of 2 mm-thickness on the FW is used as the plasma facing component (PFC). Fig.1 Schematic view of HCSB TBM 12

13 A U-shaped structure with the thickness of 30mm is used for the first wall with 87 cooling channels with dimensions of 14mm 7mm. As shown in Fig.2, Arrangement of the sub-modules by 3 6 is adopted in the structure design. The integral HCSB TBM consists of 18 breeding sub-modules. Each has independent cooling circuit and purge gas circuit. The back plate forms the back support, where the return manifolds used to collect the helium flow and the purge gas manifolds are connected. The lithium orthosilicate, Li 4 SiO 4 of with 1mm diameter and with enriched lithium of 80% 6 Li are selected as the tritium breeder. To assure an adequate Tritium Breeding Ratio (TBR), beryllium pebble is adopted as neutron multiplier in the neutron breeding zone. In order to increase the filling ratio in neutron multiplier zone, Be pebble of diameters mm are used. The helium gas is used as coolant of the Helium Cooling System (HCS) and the purge gas of Tritium Extraction System (TES). The pressure of the helium cooing system and the tritium extraction system are 8 MPa and 0.1 MPa, respectively. Main parameters of the HC-SB TBM design are shown in Table 1. Fig.3 shows the general layout of CH HC-SB TBMs in ITER test port. Fig.2 overall isometric view of sub-module Fig 3 General layout of CH TBMs in ITER port C 13

14 Table 1 Design parameters for the China HCSB TBM Configuration FW area Neutron wall loading Surface heat flux BOT (Breeder Out of Tube) 0.484m (W) 1.660m(H) Total heat deposition NT-TBM, PI-TBM 0.76MW Globe TBR Lithium orthosilicate, Li 4 SiO 4 Tritium production rate ITER operation condition Modules: 3 6 Sub-modules 0.803m MW/m MW/m (3-D), 80% Li g/d Sub-module dimension (P) (T) (R) 253.5mm 130mm 420mm Ceramic (Li 4 SiO 4 ) Neutron multiplier (Beryllium) Structure material breeder Coolant helium (He) Pipes size He purge flow (He) Diameter Thickness Max. Temperature Diameter Thickness Max. temperature RAFM Max. temperature Pressure Pressure drop Temp. range (inlet/outlet) Mass flow Diameter (OD/ID) Pressure Pressure drop 1mm, pebble bed 90mm (four zones) ~1mm, Pebble bed 200mm (five zones)+2mm (armor) 550 (Armor) 613 (Be Pebble bed) RAFM 537 8MPa 0.22MPa 300/ kg/s 101.6/85.5mm 0.12MPa 0.02MPa A design concept for the China ITER HC-SB TBM has been proposed. Preliminary design and performance analysis for the TBM module have been performed. The results show that the proposed TBM design is feasible within the existing technologies. Ii is characterised by simple structure, mature technical. The design description document (DDD) have been carried out in last year. The further HC-SB TBM design works will update and optimize the structure design as well as ancillary subsystem parameters. 14

15 Fusion Reactor Research Activities SWIP (7) Research and Development of ITER bellows assemblies In ITER facilities, there are some special expansion joints connected with the openings of Cryostat. Expect for the assemblies connected with NB openings, others are rectangular with big dimensions and severe working conditions and they are important for ITER safe runing. Owing to the class of safety of these bellows assemblies and special mechanical properties of rectangular bellows from circular ones, our research group include some companies such as Nanjing Chenguang, 725 Institute and Shenyang Huiyu. Designing, fabricating and testing of ITER bellow assemblies are guided by ITA and IT corresponding members. 1. Main design parameters for ITER expansion joints Port duct bellows from the three types of bellows asemblies, are in the circumstance of vaccuum and neutron iradiation, rectangular, big differences of temperatures between the two ends and of pressures inside and outside. And so they should be focused on. After determining the design requirements for ITER bellows asseblies, following parameters can be got for ITER Port duct bellows assemblies. Table 1 Strctual parameters for Port duct bellows asseblies(pitch90 round corner200 convolutions3)) symbol unit Port Duct Bellows upper equatorial Lower Nominal thickness t Mm Length for longer side D l Mm 3,240 3,240 3,320 Length for shorter side D s mm 2,240 2,820 3,240 Distance between the outermost ends of bellows L u mm 1,220 1, Length for intermittent duct L id mm Total length of the assemblies L t mm 1,420 1,270 1,030 Material for bellows L stainless steel Port duct bellows has four different pressure-loading. They are corresponding to 20 C,100 C,120 C,200 C and 270 C. The EJMA results show that S7l+S7s Sab, S7l+S8l 1.5Sab and S7s+S8s 1.5Sab are meeted. Besides stress analysis, calculation and evalutaion for life cycles are also done. FEB analyses are also been used in bellows design and analyses. It shows that first ten frequencies are 54.7, 64.8, 68.1, 69.7, 70.1, 84.9, 91.6, 96.2, 98.0,

16 1 NODAL SOLUTION STEP=1 SUB =4 FREQ= USUM (AVG) RSYS=0 DMX = SMX = MN AUG :50:01 MX Y Z X Fig 1 Some FEB analyses results 2. Researches for fabrication of Port duct bellows There are many fabrication processes for bellows. From those, Forming and welding are most important. Port duct rectangular bellows are with round-corner, big ratio of convolution height and pitch, big dimensions and thickness. For such bellows, three types of forming are considered: 1) Hydrostatic pressurising for rounder corner and bending for straight line and then welding; 2) forming together rounder coners and parts of staight lines, and ten bending the other part of straight lines,and then welding; 3) presure forming for all bellows. Welding for port duct bellows are including three parts: welding during forming of the bellows, welding during forming of intermittent ducts and welding between bellows and ducts. Fig 2 researches of key fabrication processes for rectangular bellows 3. Testing Testing is also key process for bellows production. Several schemes have been considered according to opinions and suggestions from expasion-joint makers. 16