Towards the development of workable acceptance criteria for the divertor CFC monoblock armour

The plasma-facing components (PFCs) of the ITER divertor will be subjected to high heat flux (HHF). Carbon–fibre composite (CFC) is selected as the armour for the region of highest heat flux where the scrape-off layer of the plasma intercepts the vertical targets (VT). Failure of the armour to heat sink joints will compromise the performance of the divertor and could ultimately result in its failure and the shut down of the ITER machine. There are tens of thousands of CFCs to CuCrZr joints. The aim of the PFC design is to ensure that the divertor can continue to function even with the failure of a few joints. In preparation for writing the procurement specification for the ITER vertical target PFCs, a programme of work is underway with the objective of defining workable acceptance criteria for the PFC armour joints.     

Silicon doped carbon/Cu joints based on amorphous alloy brazing for first wall application

Amorphous ribbon-type filler-metals represent a promising selection for joining heterogeneous materials together. In this work, rapidly solidified ribbon-type Ti based amorphous filler with a melting temperature of 850 °C and a thickness up to 20 μm is used to join silicon doped carbon to pure copper. SEM examinations demonstrate that a high quality brazed joints could be acquired. The brazed seam has a uniform structure and pore free along its entire length. TiC and ZrC are formed near the interface of carbon and filler-metal when the brazing holds enough time. Using very thin Mo and Cu foil (0.2 mm in thickness) as multiple interlayer are very effective to mitigate the thermal stress occurred in the interface between carbon and copper. The shear strength of this carbon-multiple interlayer-copper joint is more than 20 MPa, and the rupture is mainly occurred on the carbon side.     

Design and evaluation of an optimized W/Cu interlayer for W monoblock components

Divertor plasma-facing components of future fusion reactors should be able to withstand heat fluxes of 10-20 MW/m² in stationary operation. Tungsten blocks with an inner cooling tube made of CuCr1Zr, so-called monoblocks, are potential candidates for such water-cooled components. To increase the strength and reliability of the interface between the W and the cooling tube of a Cu-based alloy (CuCr1Zr), a novel advanced W-fibre/Cu metal matrix composite (MMC) was developed for operation temperatures up to 550 °C. Based on optimization results to enhance the adhesion between fibre and matrix, W fibres (W_f) were chemically etched, coated by physical vapour deposition with a continuously graded W/Cu_PVD interlayer and then heated to 800 °C. The W_f/Cu MMC was implemented by hot-isostatic pressing and brazing process in monoblock mock-ups reinforcing the interface between the plasma-facing material and the cooling channel. The suitability of the MMC as an efficient heat sink interface for water-cooled divertor components was tested in the high heat flux (HHF) facility GLADIS. Predictions from finite element simulations of the thermal behaviour of the component under loading conditions were confirmed by the HHF tests. The W_f/Cu MMC interlayer of the mock-ups survived cyclic heat loads above 10 MW/m² without any damage. One W block of each tested mock-up showed stable thermal behaviour at heat fluxes of up to 10.5 MW/m².     

Potential and limits of water cooled divertor concepts based on monoblock design as possible candidates for a DEMO reactor

In this paper water-cooled divertor concepts based on tungsten monoblock design identified in previous studies as candidate for fusion power plant have been reviewed to assess their potential and limits as possible candidates for a DEMO concept deliverable in a short to medium term (“conservative baseline design”). The rationale and technology development assumptions that have led to their selection are revisited taking into account present factual information on reactor parameters, materials properties and manufacturing technologies.For that purpose, main parameters impacting the divertor design are identified and their relevance discussed. The state of the art knowledge on materials and relevant manufacturing techniques is reviewed. Particular attention is paid to material properties change after irradiation; phenomenon thresholds (if any) and possible operating ranges are identified (in terms of temperature and damage dose). The suitability of various proposed heat sink/structural and sacrificial layer materials, as proposed in the past, are re-assessed (e.g. with regard to the possibility of reducing peak heat flux and/or neutron radiation damages). As a result, potential and limits of various proposed concepts are highlighted, ranges in which they could operate (if any) defined and possible improvements are proposed.Identified missing point in materials database and/or manufacturing techniques knowledge that should be uppermost investigated in future R&D activities are reported.This work has been carried out in the frame of EFDA PPPT Work Programme activities.     

1000MW核电管板纯净钢锻件制造工艺及其性能

根据1000MW核电管板用纯净钢锻件性能和组织的要求，利用合金化原理确定了冶炼时钢中各合金元素的成分控制方向；采用电炉加钢包炉加真空浇注进行冶炼浇注，真空浇注过程中加保护防止二次氧化；采用特殊的镦粗工艺避免了管板镦粗过程心部产生超标缺陷；采用合理的热处理工艺，保证管板锻件的组织和性能。经检验表明，锻件用钢的质量达到了纯净钢的要求，管板锻件的综合性能达到世界领先水平。     

Thermal analysis and tests of W/Cu brazing for primary collimator scraper in CSNS/RCS

To the transverse beam collimation system in a rapid cycling synchrotron,an important component is the primary collimator,which improves emittance of the beam halo particles such that the particles outside the predefined trajectory can be absorbed by the secondary collimators.Given the material properties and power deposition distribution,the beam scraper of the primary collimator is a0.17 mm tungsten foil on a double face-wedged copper block of 121.5 mm x 20 mm.The heat is transferred to the outside by a φ34 mm copper rod.In this paper,for minimizing brazing thermal stress,we report our theoretical analysis and tests on brazing the tungsten and copper materials which differ greatly in size.We show that the thermal stress effect can be controlled effectively by creating stress relief grooves on the copper block and inserting a tungsten transition layer into the copper block.This innovation contributes to the successful RD of the primary collimator.And this study may be of help for working out a brazing plan of similar structures.     

The impact of thermal fatigue and carbidization on the W coatings deposited on CFC tiles for the ITER-like Wall project at JET

Since August 2011 JET operates with the ITER-like wall comprising bulk Be tiles, bulk W tiles and W coated CFC tiles with a thickness of 10–15 μm and 20–25 μm. In order to evaluate behavior of the W coatings to a cyclic thermal loading relevant to JET operation, high heat flux (HHF) tests have been carried out up to 5100 pulses with an electron beam facility at peak temperatures of 1000 °C, 1250 °C and 1450 °C. The pulse duration was 24 s. Optical inspections of the W layer performed periodically by interrupting the test revealed small delaminations with the size of 50–500 μm. The dependence of the delamination percentage on the number of pulses can be seen as a degradation curve for each particular W coating. In this way the thermo-mechanical properties of the W coatings can be characterized quantitatively. Thermal fatigue and carbidization of the tungsten due to the diffusion of the carbon from the substrate have been recognized as mechanisms for degradation of the coatings. Tungsten carbides have been identified by using TEM (transmission electron microscopy) diffraction analysis on FIB (focused ion beam) prepared cross-section samples subjected to HHF tests. Nano-pores developed at the CFC–Mo and Mo–W interfaces during the tests might be also responsible for the degradation of the coating.     

Design and setup of a testing device to investigate a reduced sized attachment system mock up for the ITER EU HCPB-TBM under different mechanical loading conditions

In order to determine the forces acting on the EU-Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) during operation, a measurement system is developed. Therefore, two force reconstruction (FR) methods using measured strain signals are selected that are suitable for the application to the TBM. The first one, the augmented Kalman filter is a combined deterministic-stochastic approach. A second FR method based on the concept of a model predictive controller is proposed in this paper, which uses an optimization algorithm. In order to test the selected methods a testing device has been built which can be used to apply different force excitations on a reduced sized TBM mock up and measure the resulting strain signals of 16 strain gages. A simple tube mock up has been designed and manufactured to test and calibrate the FR algorithms. In addition, a second TBM mock up with attachment system is described. Finally, first results of the FR of a worst-case test case from simulated strain data of the simple tube mock up are presented.     

National Ignition Facility TestController for automated and manual testing

The Controls and Information Systems (CIS) organization for the National Ignition Facility (NIF) has developed controls, configuration and analysis software applications that combine for several million lines of code. The team delivers updates throughout the year, from major releases containing hundreds of changes to patch releases containing a small number of focused updates. To ensure the quality of each delivery, manual and automated tests are performed using the NIF TestController test infrastructure. The TestController system provides test inventory management, test planning, automated and manual test execution, release testing summaries and results search, all through a web browser interface. As part of the three-stage software testing strategy, the NIF TestController system helps plan, evaluate and track the readiness of each release to the NIF production environment.After several years of use in testing NIF software applications, the TestController's manual testing features have been leveraged for verifying the installation and operation of NIF Target Diagnostic hardware. The TestController recorded its first test results in 2004. Today, the system has recorded the execution of more than 160,000 tests and continues to play a central role in ensuring that NIF hardware and software meet the requirements of a high reliability facility. This paper describes the TestController system and discusses its use in assuring the quality of software delivered to the NIF.     

A reactor system for radiation research and materials testing

Translated from Atomnaya Énergiya, Vol. 68, No. 6, pp. 438–441, June, 1990.     

Effects of the fabrication process parameters on the precipitates and mechanical properties of a 9Cr-2W-V-Nb steel

The effects of the fabrication process parameters such as a tempering temperature, cold rolling and annealing condition on the precipitates and mechanical properties of a normalized 9Cr-2W-V-Nb steel were evaluated. Nb-rich MX precipitates were found in the specimen tempered at 550 °C while M₂₃C₆, Nb- and V-rich MX ones were observed in the specimen tempered at 750 °C. A cold rolling and an annealing at 750 °C of the specimen tempered at 550 °C induced the formation of large inhomogeneous M₂₃C₆ carbides, causing a reduced tensile strength. However, the cold rolling of the specimen tempered at 750 °C provided fine precipitates due to a fragmentation of some of the M₂₃C₆ carbides, and an annealing at 700 °C for 30 min was found to be suitable to recover the degraded mechanical properties from a cold working.     

Joint resistance measurements of pancake and terminal joints for JT-60SA EF coils

To evaluate the joint fabrication technology for the JT-60SA EF coils, joint resistance measurements were conducted using a sample consisting of pancake and terminal joints. Both joints are shake-hands lap joints composed of cable-in-conduit conductors and a pure copper saddle-shaped spacer. The measurements demonstrated that both joints fulfilled the design requirement. Considering these measurements, the characteristics of both joints were investigated using analytical models that represent the joints. The analyses indicated that the characteristics of the conductors used in the joints affect the characteristics of the joints.     

Combined cryogenic and mechanical pump for NET torus exhaust

This novel pumping concept consists of a 20 K stage to increase gas density followed by a mechanical pump operating at the same temperature. Advantages of the concept include order-of-magnitude reductions in size, weight, stored energy, eddy-current rotor heating, and sudden venting thrust when compared to a conventional turbomolecular pump scaled-up to perform the same duty. The device offers extremely low tritium inventory and near steady-state operation. Critical design and development issues include minimizing backstreaming in the mechanical portion of the pump, and eddy-current heating of the rotor.     

Need for and requirements for a neutron irradiation facility for fusion materials testing

The construction and operation of an intense 14-MeV neutron source is essential for the development and eventual qualification of structural materials for a fusion reactor demonstration plant (DEMO). Because of the time required for materials development and the scale-up of materials to commercial production, a decision to build a neutron source should precede engineering design activities for a DEMO by at least 20 years. The characteristic features of 14-MeV neutron damage are summarized including effects related to cascade structure, transmutation production, and dose rate. The importance of a 14-MeV neutron source for addressing fundamental radiation damage issues, alloy development activities, and the development of an engineering database is discussed. For these considerations, the basic requirements and machine parameters are derived.     

Manufacturing and testing of a HETS module for DEMO divertor

The development of a divertor concept for fusion power plants that is able to grant efficient recovery and conversion of the considerable fraction (～15%) of the total fusion thermal power incident is deemed to be an urgent task to meet in the EU Fast Track scenario. The He-cooled conceptual divertor design is one of the possible candidates. Helium cooling offers several advantages including chemical and neutronic inertness and the ability to operate at higher temperatures and lower pressures than those required for water cooling. The HETS (high-efficiency thermal shield) concept, initially developed by ENEA for water, has been adapted for use with He as coolant. This DEMO divertor concept is based on elements joined in series and protected by a hemispheric dome; it allows an increase of thermal exchange coefficient both for high speed of gas and for “jet impingement” effects of gas coming out from the internal side of hemispheric dome. It has been calculated to be capable of sustaining an incident heat flux of 10 MW/m² when operating at 10 MPa, an inlet He temperature of 600 °C, and an outlet temperature of 800 °C. The presented activity, performed in the frame of EFDA-TW5TRP-001 task, was focused on the manufacturing of a single HETS module and on its thermal–hydraulic testing. The materials used for the HETS module manufacturing were all DEMO-compatible: W for the armor material and the hemispherical-dome, DENSIMET for the exchanger body. The testing is performed by connecting the module to HEFUS3 He loop system that is a facility able to supply the He flow to the required testing conditions: 400 °C, 4–8 MPa and 20–40 g/s. The needed incident heat flux is obtained by RF inducting equipment coupled to an inductor coil installed just over the HETS module. A CFD analysis by ANSYS-CFX was performed in order to predict the thermal–mechanical behavior of the module and a final comparison with the experimental data is required to validate the CFD results. All parameters are monitored and recorded by data acquisition system.     

The nondestructive testing of tubes and pipes for nuclear application

The directive of the Reactor Safety Commission demands for all materials which are provided for the pressure bearing enclosure of the refrigerant a nondestructive testing with sufficient sensibility. The specification 3201.1 for nuclear application as well as company-internal rules of important manufacturers regulate the requirements derived from the above direction for the NDT of tubes and pipes.For an objective and reproducible testing, equipments with defined characteristics are employed. based on internal specifications, testing equipments are fabricated and then checked with a special computerized test system. Moreover probes are controlled with regard to their acoustic and electric properties.The NDT of heat exchanger tubes and of pipes is given here as an example:Heat exchanger tubes: The tests include the inspection of longitudinal and transverse defects, wall thickness, dimension and tightness. In connection with the NDT, defect catalogues are set up. By this means the chosen test sensitivity is verified, and so the high quality of the tubes is assured. Specially developed eddy-current methods prove that such tested tubes are free of corrosion-causing phases.Pipes: The pipes are tested for longitudinal and transverse defects, for laminations and for wall thickness. To fulfil the demand for an objective and reproducible testing, there was developed and installed an automatic, computer-controlled ultrasonic equipment with 40 probes.Development trends: For the NDT of heat exchanger and boiler tubes an electrodynamic excited ultrasonic test system is evolved which is also able to test curved and installed tubes. The sophisticated testing technology is completed by a qualified education and training of NDT personnel.     

CENDL-3.1铜的检验和改进

介绍了利用屏蔽基准实验OKTAVIAN以及核临界安全手册(ICSBEP)中的临界基准实验对CENDL-3.1铜的伞套中子评价数据进行的宏观检验.在屏蔽基准检验中,除了中子和γ泄漏谱上发现了由非弹性散射截面造成的与实验测量结果的分歧,计算结果与实验符合相当好.在快中子谱临界基准检验中,装置HMF072、HMF073和PMF013的keff的计算结果高出实验值大约2%,严重偏离实验结果.针对HMF072装置的灵敏度分析显示,该分歧的产生主要是由于全截面在0.1～1.3 MeV能区的评价不当引起的.在对0.1～1.3 MeV的全截面进行修正后,临界检验的结果获得了明显改善.