Studies on cooling by two-dimensional confined jet flow of high heat heat flux surface in fusion reactor

Divertor surface of a magnetic confinement fusion reactor is exposed to strong radiative heating. According to standard design of the ITER, maximum heat flux on the divertor surface becomes locally near 30 MW m⁻². To cool such high heat flux surface by water flow, it is necessary to establish a cooling method which enhanced the critical heat flux (CHF). We proposed a cooling by a planar impinging jet with free surface in the previous report. In the jet cooling on flat surface, high CHF was obtained in the limited region where the jet flow hits directly. As apart from the region, the CHF decreases abruptly with the distance from the center. To overcome this difficulty, it was proposed that the planar jet is applied to cool concave surface where the centrifugal force is efficiently used to enhance the CHF. In this study, the CHFs were investigated in the confined jet flow which was guarded by a wall on the other side of the heated wall, because the guard wall works to protect splash of water from liquid film by violent boiling and expects further enhancement of the CHF. In this study, the CHFs were investigated in the confined flow of two-dimensional jet on flat and concave surfaces in the various flow conditions and got a correlation for the CHF. Applicability of this cooling for divertor surface was assessed by using the experimental results.     

Preliminary study of high neutron flux fusion heating

A heating scheme for nuclear fusion is proposed based on the availability of a high flux, low energy neutron source. The heat is derived in the reaction ⁶Li (n, T) ⁴He resulting from the incidence of a low energy neutron beam on a sample of ⁶Li D. The energy release per reaction, Q = 4.6 MeV, is converted through electron Coulomb collisions thereby quickly dissociating the solid sample to the plasma state. For 10⁻³ eV neutrons it is estimated that this dissociation occurs in 7 ms for an incident flux of 10¹⁷ cm⁻² · s⁻¹. The possibility of further driving the heated fuel to fusion is also discussed.     

Thermal-hydraulic analysis techniques for axisymmetric pebble bed reactor cores

A mathematical model for the analysis of coupled thermal-hydraulic problemsin steady-state pebble bed nuclear reactor cores is presented. The bed has been treated macroscopically as a generating, conducting porous medium. The model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, and new coefficients of the viscous and inertial loss terms are presented. The remaining equations in the model make use of continuity and thermal energy balances on the solid and fluid phases. None of the usual simplifying assumptions such as constant properties, constant velocity flow or negligible conduction and/or radiation are used. A computer program based on this model has been constructed; it has been validated by comparing predictions with measured values of previous experiments. Validation of the nonlinear fluid flow model is reported in a companion paper.     

Initiation and propagation of damage in actively cooled CFC armoured high heat flux components in fusion machines

Plasma facing components (PFCs) in magnetic confinement controlled fusion machines are armoured with carbon fibre composite (CFC) bonded to a copper alloy heat sink. The manufacturing process induces high level of residual stresses due to the thermal expansion mismatch between CFC and copper and PFCs have to withstand strong stress ranges during operation. To study the initiation and propagation of damage in the CFC part, the ONERA damage model is used to describe the behaviour of the N11 material. The finite element simulations show that the damage is located near the interface and develops during the manufacturing of the PFCs as a consequence of the high amplitude of shear stresses. Under high heat flux, stresses decrease and the damage does not evolve. Further studies will take into account the damageable behaviour of the composite/copper interface, which will lead to geometrical optimisations and better knowledge of the link between damage and conductivity.     

Super X divertors for solving heat and neutron flux problems of fusion devices

We present a new magnetic geometry, called the Super X divertor (SXD), that could potentially solve the enormous heat exhaust problem of next-generation high power-density experiments and fusion reactors. With only small changes in net coil currents, the axisymmetric SXD modification of the standard divertor (SD) coils greatly increases the divertor radius, the line length, and the plasma-wetted area. The lower B at large R decreases parallel heat flux and hence lowers the plasma temperature at SXD plates to below 10 eV, allowing higher divertor radiation fractions. The SXD could safely exhaust five times more heat than an SD, is unique in allowing adequate shielding of divertor target from neutron damage, and can enable much improved, reactor-relevant core plasma performance.     

Fatigue lifetime of repaired high heat flux components for ITER divertor

Thermal fatigue behaviour of repaired monoblocks was assessed from High Heat Flux (HHF) tests up to 20 MW m^?2 on 11 components. Among these components, 8 monoblocks were repaired (2 CFC and 6 tungsten). These components were manufactured by two EU industries: ANSALDO Ricerche and PLANSEE. Non destructive examination was performed on SATIR thermography test bed before and after HHF tests. SATIR results show that repaired monoblocks have a good thermal exhaust capability before HHF tests. For all monoblocks, no degradation of thermal properties was noticed during cycles at 10 MW m^?2. After hundreds of cycles at 20 MW m^?2, two W repaired monoblock melted. Post-HHF SATIR examination revealed a degradation of thermal properties which is systematic for W melted monoblocks and non-systematic for W repaired ones. For CFC repaired monoblocks, no damage was observed up to 20 MW m^?2. For the first ITER divertor set, specifications for the pre-qualification are that CFC (Resp. W) components have to sustain in steady state 1000 cycles at 10 MW m^?2 (Resp. 3 MW m^?2) followed by 1000 cycles at 20 MW m^?2 (Resp. 5 MW m^?2). For the first ITER divertor set, the repair process is validated for CFC and W monoblocks.     

Improvement of a temperature response function for divertor heat flux monitoring in fusion devices

Temperature response functions have been developed to investigate sensor design and divertor heat flux estimation in magnetically confined plasmas. The time-dependent heat flux can be derived by fitting the response function to experimental thermocouple(TC) data. Because the TC signals have a time delay to transit events such as discharge start or confinement transition, the time delay is taken into account in a temperature response function. Such a function accurately describes the signal from ...     

Results and consequences of high heat flux testing as quality assessment of the Wendelstein 7-X divertor

The series manufacturing of the first 282 Wendelstein 7-X divertor elements was concluded in 2011. The divertor is designed to remove a steady-state heat load of 10 MW/m². 940 target elements of five different types made of CuCrZr heat sinks and covered with 16,000 CFC NB31 flat-tiles have to be produced. Additional to quality assessment during the manufacturing process, a final assessment of the delivered elements with operational heat load is indispensable to ensure a constant high thermal performance of the installed divertor.Based on the results of the pre-series testing a statistical quality assessment method has been developed for the series production. The application of this method to the series elements ensures their thermal performance with reasonable high heat flux test effort.     

Testing of high heat flux components manufactured by ENEA for ITER divertor

ENEA is involved in the International Thermonuclear Experimental Reactor (ITER) R&D activities and in particular in the manufacturing of high heat flux plasma-facing components, such as the divertor targets. During the last years ENEA has manufactured actively cooled mock-ups by using different technologies, namely brazing, diffusion bonding and HIPping. A new manufacturing process that combines two main techniques PBC (Pre-Brazed Casting) and the HRP (Hot Radial Pressing) has been set up and widely tested.A full monoblock medium scale vertical target, having a straight CFC armoured part and a curved W armoured part, was manufactured using this process.The ultrasonic method was used for the non-destructive examinations performed during the manufacturing of the component, from the monoblock preparation up to the final mock-up assembling. The component was also examined by thermography on SATIR facility (CEA, France), afterwards it was thermal fatigue tested at FE200 (200 kW electron beam facility, CEA/AREVA France).The successful results of the thermal fatigue testing performed according the ITER requirements (10 MW/m², 3000 cycles of 10 s on both CFC and W part, then 20/15 MW/m², 2000 cycles of 10 s on CFC/W part, respectively) have confirmed that the developed process can be considerate a candidate for the manufacturing of monoblock divertor components. Furthermore, a 35-MW/m² Critical Heat Flux was measured at relevant thermal–hydraulics conditions at the end of the testing campaign.This paper reports the manufacturing route, the thermal fatigue testing results, the pre and post non-destructive examination and the destructive examination performed on the ITER vertical target medium scale mock-up.These activities were performed in the frame of EFDA contracts (04-1218 with CEA, 93-851 JN with AREVA and 03-1054 with ENEA).     

Critical heat flux at high velocity channel flow with high subcooling

A quantitative analysis of critical heat flux (CHF) in heated channels under high mass flux with high subcooling was successfully carried out by applying a new flow model to the existing CHF model of a macro-water-sublayer on the heated wall and steam blankets over it. The CHF correlation proposed could correctly predict the existing experimental data for circular tubes of 0.33–4 mm in diameter with mass flux of 124–90 000 kg (m² s)⁻¹ and inlet water subcooling of 35–210 K at 0.1–7.1 MPa, resulting in CHF of 4.2–224 MW m⁻², and for rectangular channels of 3–20 mm gap with a mass flux of 940–27 000 kg (m² s)⁻¹ and inlet water subcooling of 13–166 K at 0.1–3.0 MPa, resulting in CHF of 2.0–62 MW m⁻². An error of the CHF correlation has also been estimated.     

Thermal-hydraulic design of water-cooled pressure tube blanket for a fusion driven subcritical reactor

A Water-cooled Pressure Tube Energy production blanket (WPTE) for fusion driven subcritical reactor has been designed to achieve 3000 MW thermal power with self-sustaining tritium cycle. Pressurized water has great advantages in energy production; however the high pressure may cause some severe structural design issues. This paper proposes a new concept of water-cooled blanket. To solve the problem of the high pressure of the coolant, the pressure tube was adopted in the design and in the meantime, the thickness of the first wall can be significantly reduced as result of adopting pressure tube. The numerically simulating and calculating of temperature, stress distribution and flow analyses were carried out and the feasibility of using water as coolant was discussed. The results demonstrated the engineering feasibility of the water-cooled fusion–fission hybrid reactor blanket module.     

Calculating decay heat removal rates via natural circulation along variable heat flux reactor fuel plates

A recently developed integral technique is applied to natural convection cooling along test reactor fuel plates. The technique is demonstrated for water and air flow. In the case of air flow, the process is characterized by a large temperature rise along the fuel channel, thereby rendering the commonly applied Boussinesq approximation invalid. This case is a heat transfer problem of particular interest in accident analyses such as determining the level of decay heat dissipation possible, without exceeding the melting temperature of the fuel, subsequent to a hypothetical loss of primary coolant.     

TACR热传输系统热工水力瞬态分析

钍燃料的利用对于缓解核燃料资源短缺具有重要意义,坎杜型反应堆(Canadian Deuterium Uranium,CANDU)在堆芯布置、中子利用效率及先进燃料循环方面具有较高的灵活性,使得其在CANDU反应堆中引入钍燃料循环更具现实意义。CANDU型反应堆中钍基燃料应用关键基础技术研究是加拿大与我国正在开展的合作课题,其中开发自主的CANDU堆堆芯热工水力设计和安全分析程序是钍基燃料应用必不可少的设计工作之一。本文针对CANDU型反应堆热传输系统结构特点,采用FORTRAN程序设计语言开发了适用于CANDU型反应堆热传输系统的热工水力瞬态分析程序CANTHAC(CANDU Thermal-Hydraulic Analysis Code)。利用CANTHAC对钍基先进CANDU堆(Thorium-based Advanced CANDU Reactor,TACR)进行了瞬态分析,计算工况包括满功率稳态、无保护蒸汽发生器(Steam Generator,SG)二次侧给水温度降低事故及完全失流事故。其中,满功率稳态计算结果与清华大学设计的钍基先进CANDU堆TACR设计值吻合较好,相对误差不超过2%,在可接受范围内;无保护SG二次侧给水温度降低事故及完全失流事故在计算条件下所得的燃料温度及系统压力等关键热工水力参数均在安全限值内,满足安全准则要求。程序为模块化编程,便于移植和改进,具有一定的通用性,为进一步研究工作奠定了基础。     

Thermal characterisation of ceramic/metal joining techniques for fusion applications using X-ray tomography

This work investigates the thermal performance of four novel CFC–Cu joining techniques. Two involve direct casting and brazing of Cu onto a chromium modified CFC surface, the other two pre-coat a brazing alloy with chromium using galvanisation and sputtering processes. The chromium carbide layer at the interface has been shown to improve adhesion. Thermal conductivity across the join interface was measured by laser flash analysis. X-ray tomography was performed to investigate micro-structures that might influence the thermal behaviour. It was found that thermal conductivity varied by up to 72%. Quantification of the X-ray tomography data showed that the dominant feature in reducing thermal conductivity was the lateral spread of voids at the interface. Correlations were made to estimate the extent of this effect.     

Non-destructive methods for the defect detection in the ITER high heat flux components

This paper discusses the application of non-destructive testing (NDT) by ultrasonic technique for the control of the joining interfaces of the ITER divertor vertical target plasma facing units. The defect detection capability has to be proved for both metal to metal and metal to carbon/carbon fibre composite (CFC) joints because these two types of joints have to be realized for the manufacturing of the high heat flux units. In this paper the UT results coming from the investigation performed during the manufacturing, but also after the thermal fatigue testing (up to 20 MW/m²) of six mock-ups manufactured using the Hot Radial Pressure technology (HRP) in ENEA labs are presented and compared with the evidences from the final destructive examination. Regarding the Cu/CFC joint, the effectiveness of the ultrasonic test has been deeply studied due to the high acoustic attenuation of CFC to ultrasonic waves. To investigate the possibility to use the ultrasonic technique for this type of joint, an ‘ad hoc’ flat Cu/CFC joint sample, that reproduces the actual annular joint interfaces, was manufactured. This flat sample has the advantage of being easily tested by probes with different geometry and frequency. UT results are compared with X-ray and eddy current testing of the same sample.     

KATHELO: A new high heat flux component testing facility

In this article a new helium cooled test facility is presented. The loop, KATHELO, is designed to operate at pressures up to 10 MPa and temperatures up to 800 °C in order to be able to provide suitable testing conditions for the helium cooled divertor concept developed at KIT. The general layout of the loop is introduced and the some of the technological solutions adopted for the loop components are discussed. The thermal-dynamic behavior of the loop during start-up and steady state operation for two different scenarios is analyzed using a RELAP-3D model of the circuit. Based on these simulations the heating power needed for high temperature operation is estimated. A particular attention is given to the thermal coupling between the cold and hot leg of the loop with emphasis on overall loop efficiency.     

Hydrodynamic transition delay in rectangular channels under high heat flux

When upgrading a research nuclear reactor for a higher power output it can be expected that the cooling flow rate has to be increased. In the case of a reactor designed with a laminar cooling flow this upgrade may take the flow into the transition hydrodynamic regime.     

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.     

Thermal response of plasma sprayed tungsten coating to high heat flux

In order to investigate the thermal response of tungsten coating on carbon and copper substrates by vacuum plasma spray (VPS) or inert gas plasma spray (IPS), annealing and cyclic heat load experiments of these coatings were conducted. It is indicated that the multi-layered tungsten and rhenium interface of VPS-W/CFC failed to act as a diffusion barrier at elevated temperature and tungsten carbides were developed after 1 h incubation time when annealing temperature was higher than 1600 °C. IPS-W/Cu and W/C without an intermediate bonding layer were failed by the detachment of the tungsten coating at 900 and 1200 °C annealing for several hours, respectively. Cyclic heat load of electron beam with 35 MW/m² and 3-s pulse duration indicated that IPS-W/Cu samples failed with local detachment of the tungsten coating within 200 cycles and IPS-W/C showed local cracks by 300 cycles, but VPS-W/CFC withstood 1000 cycles without visible damages. However, crack creation and propagation in VPS-W/CFC were also observed under higher heat load.