Boiling heat transfer during rewetting

An analytical one-dimensional conduction model is developed for the rewetting of a hot surface. The solution scheme is not limited to a specific heat flux profile. Experimental results falling within the range of validity of the one-dimensional solution are correlated using a generalized pool boiling curve. The effect of various heat transfer parameters such as inlet velocity and local subcooling is discussed. A unique relationship is shown to exist between the dimensionless quench front velocity, which is defined here as an eigenvalue of the governing equation, the ratio of the integral of the dimensionless heat flux, and the integral of the surface axial temperature gradient.     

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.     

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.     

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).     

循环工况转换过程中的燃料棒热应变分析

利用RELAP5程序建立分析模型,结合强迫循环工况转换自然循环工况的试验数据,对转换过程中的冷却剂温度变化速率和燃料棒的热应变进行计算分析.计算结果表明:结合RELAP5程序进行燃料棒的热应变分析是合适的;自然循环转换过程中燃料棒热应变的变化剧烈,对燃料棒的机械性能产生影响;结合RELAP5瞬态分析结果,也可以分析其他工况变化过程对燃料棒的热应力冲击问题.     

Use of concretes for high-temperature shielding of nuclear reactors

The authors have studied the possibility of using chromite and chomotte heat-resistant concretes for the thermal shields of reactors. They observe neutron fluxes of various intensities (up to 10¹³ neutrons/cm²·sec, with spectrum similar to fission spectrum), absorbed by shields of these materials. They compute the transmission of neutrons and of fluxes of gamma quanta and the heat emission in the shielding. They calculate the temperatures in the shielding for various neutron fluxes, concrete thicknesses and cooling conditions. They perform a statistical calculation of the temperature stresses for shielding constructed of heat-resistant ferroconcrete.It was established that nuclear reactor shields can be made from heat-resistant ferroconcrete when the neutron fluxes on the concrete are up to 10¹³ neutrons/cm²·sec, for temperatures up to 1000–1100° C and temperature differences of up to 900° C.Translated from Atomnaya Énergiya, Vol. 19, No. 6, pp. 524–529, December, 1965Report read by G. I. Budker at the International Conference on High-Energy Accelerators (Frascati, Italy).     

Estimation of the activation of local reactor shielding concretes

Many elements in reactor shielding concrete become radioactive due to interaction with neutrons during the operation of a nuclear reactor. These radioactive elements will build up over the time due to (n,γ) reaction and contribute to the radioactive waste during decommissioning of the reactor, they will increase the dose from the inner part of the biological shielding (concrete) surrounding a nuclear reactor during maintenance works within reactor containment, and their effect should be taken in the calculations of the dose behind the concrete shield.Six mixtures of local ordinary and special shielding concretes have been investigated in this study. Their compositions were determined using X-ray diffraction and neutron activation analysis using the Syrian Miniature Neutron Source Reactor (MNSR) and consequently the dose rates.Based on the results of the analyses and comparing to the published limits in the references, it has been found that ⁶⁰Co, ¹⁵²Eu, ¹⁵⁴Eu and ¹³⁴Cs will be the most effective radioactive nuclides existing in the local concrete, but Dolomite aggregates contain the minimum limits of them and it has been found that Fe, Mg, and Ti the major elements existing in the local concrete.     

Gd burnable poison system for reactivity control of the first cycle of a PWR

The reactivity control of a PWR core may be performed by a system of burnable poison (BP) rods. In such a case, the soluble B system may be eliminated and the BP rods will be responsible for the excess reactivity provided for fuel depletion and fission products accumulation. A strong negative moderator temperature coefficient is a desirable safety feature, inherent to a poison-free moderator. The design objective of a PWR core controlled completely by a system of BP rods is achieved by utilization of Gd as the poison material and annular geometry of a BP rod. The proposed concept is tested as a retrofittable option for the current generation, as well as new PWR plants. A plausible incore fuel-management scheme is demonstrated, with planar power distribution, close to an acceptable range. The fuel-cycle penalty due to the residual poison content at EOC is relatively small.     

Effect of neutron irradiation on some properties of heat-resistant concretes

tn contrast to the structural materials of nuclear reactors, the radiation resistances of concretes used in biological shielding have not been sufficiently studied. A tendency has recently arisen for the preferential use of heat-resistant concretes in biological shielding instead of materials such as steel, cast iron, graphite, boron, etc., which are costly and in relatively short supply. In this paper we shall indicate the effect of reactor neutron irradiation on certain properties of Porland-cement and liquid-glass heat-resistant chromite concretes. The integral neutron flux used in this investigation was (2–2.4) x 10²¹ neutrons/cm² and the irradiation temperature up to 550° C.It was found experimentally that these concretes:retain quite high strength and elastic properties. The thermal conductivity and thermal expansion coefficient change very littte. It is concluded that such concretes may be recommended for use in the biological shielding of nuclear reactors.Translated from Atomnaya Énergiya, Vol. 21, No. 2, pp. 108–112, August, 1966.     

Assessment of a heat transfer correlations package for water-cooled plasma-facing components in fusion reactors

Subcooled water at high velocity has been selected by the ITER and NET design teams as the coolant for the divertor plates in these new tokamak fusion reactors. This paper presents a thermalhydraulics package intended for the thermal design analysis of the plasma-facing components in these facilities. The selection of heat transfer correlations was based on their performances against a bank of specially collected data bases. The results show that existing correlations for nuclear and process systems can be modified, with the aid of suitable data bases, to meet the interim design needs of these projects. Specific areas that require further validation and development are discussed.     

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.