On the theoretical-numerical study of the ITER Upper Port Plug structure hydraulic behaviour under steady state and draining and drying transient conditions

The ITER diagnostic Upper Port Plug (UPP) is a water-cooled stainless steel structure aimed to integrate within vacuum vessel the plasma diagnostic systems, shielding them from neutron and photon irradiation. Due to the very intense heat loads expected, a proper cooling circuit has been designed to ensure an adequate UPP cooling with an acceptable thermal rise and an unduly high pumping power and to perform its draining and drying procedure by injection of pressurized nitrogen.A theoretical research activity has been launched at the Department of Nuclear Engineering of the University of Palermo aiming to investigate the hydraulic behaviour of the UPP Trapezoid Section cooling circuit under steady state conditions and during its draining and drying transient procedure. The research activity has been performed following a theoretical-computational approach and adopting the RELAP5 thermal-hydraulic system code.The Trapezoid Section cooling circuit characteristic functions have been derived under steady state conditions at various coolant temperatures for both the coolant flow paths at the present under consideration for this circuit. The distributions of coolant mass flow rates along the channels of the cooling circuit have been calculated too. Results show that the flow path characterized by right plate inlet has improved hydraulic performances.The transient behaviour of the Trapezoid Section cooling circuit has been investigated during the draining and drying operational transient procedure, considering realistic operative scenarios, for both the coolant flow paths at the present under consideration for the cooling circuit. In particular, it has been found out that the recently proposed flow path seems to allow the complete draining of the Trapezoid Section circuit, eliminating the need for the drying procedure.     

On the theoretical-numerical study of the HEXCALIBER mock-up thermo-mechanical behaviour

Within the framework of the R&D activities promoted by European Fusion Development Agreement on the Helium-Cooled Pebble Bed Test Blanket Module to be irradiated in ITER, ENEA Brasimone and the Department of Nuclear Engineering of the University of Palermo (DIN) performed intense research activities on the modelling of the thermo-mechanical behaviour of both beryllium and lithiated ceramics pebble beds, that are envisaged to be used, respectively, as neutron multiplier and tritium breeder. In particular, the DIN developed a thermo-mechanical constitutive model for these pebble beds to be validated against the HEXCALIBER mock-up test campaign, carried out at the ENEA HE-FUS3 facility.The paper presents the main results of the mock-up experimental tests and of their numerical simulations performed adopting the finite element method, which allowed the DIN constitutive model to be assessed and validated.     

Experimental study on gas-injection enhanced circulation performed with the CIRCE facility

This paper describes the results of an experimental campaign concerning the possibility of achieving a steady state circulation by gas-injection in a pool containing lead–bismuth eutectic (LBE) as working fluid. The activity was aimed at gaining information about the basic mechanisms of the gas injection enhanced circulation intended as a pumping system for a liquid metal cooled reactor. In particular, the paper is focused on the experimental work performed in the CIRCE large-scale facility, installed at the ENEA Brasimone Centre for studying the fluid-dynamic and operating behaviour of ADS reactor plants cooled by LBE. The gas enhanced circulation tests were carried out for different LBE temperatures (from 200 to 320 °C), under isothermal conditions and with a wide range of argon injected flow rates (from 0.5 to 7.0 Nl/s). The gas is injected from the bottom of the riser, by means of an appropriate nozzle, and the liquid metal flow rate is measured by a Venturi-Nozzle flow meter installed in the single phase part of the test section. The obtained results allowed formulating a characteristic curve of the system and evaluating the void fraction distribution along the riser path by means differential pressure measurements, which play an important role to generating the driving force for the circulation.     

Transient thermal hydraulic modeling and analysis of ITER divertor plate system

A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m² plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.     

Analysis, verification, and benchmarking of the transient thermal hydraulic ITERTHA code for the design of ITER divertor

An analytical study for the International Thermonuclear Experimental Reactor Thermal Hydraulic Analysis code (ITERTHA) is carried out for a copper divertor with a 5 mm tungsten tile. The influence of the incident heat flux, swirl-tape insertion in cooling channels as well as the coolant flow velocity on the divertor thermal response is analyzed and discussed. The ITERTHA code results are verified by the commercial finite element code, COSMOS. The heat transfer coefficients at the nodes located on the cooling channel-wall are determined outside COSMOS code by the same methodology used in ITERTHA. A good agreement is achieved under different incident heat fluxes. The ITERTHA code is also benchmarked against the thermal-hydraulic calculation of the outer divertor of the Fusion Ignition Research Experiment, FIRE for an incident heat flux of 20 MW/m² and coolant flow velocity of 10 m/s in a cooling channel of 8 mm diameter with swirl-tape inserts of 2 ratio and 1.5 mm thickness. The results show excellent agreement for both steady and transient states and prove the successful implementation of both the hydraulic and heated diameters of the swirl-tape channels in the used heat transfer correlations.     

HETS performances in He cooled power plant divertor

In the frame of the activities aimed to evaluate the performances of a He cooled divertor in the future fusion power plant, the High Efficiency Thermal Shield (HETS) concept has been proposed. This concept relies on an abrupt change of momentum of the fluid in order to increase the turbulence in the gas, and therefore the heat transfer.The requirements for the power plant divertor are, not only to sustain the required thermal flux (10 MW/m²), but He to have such properties as to be used directly in the energy production cycle and limit the required pumping power to no more than 10% of the divertor thermal power. Analytical studies developed by ENEA and UKAEA showed that the HETS concept can satisfy all the basic requirements.Experimental validations are required for the pressure drop estimates in HETS. Preliminary experiments have been performed using air at room temperature and showed values of pressure drop in line with the parameters used in the calculations. Studies aimed to optimize the shape of the channel, to further reduce the pressure drop are in progress.     

Preliminary studies on PbO reduction in liquid Pb-Bi eutectic by flowing hydrogen

In the framework of the Italian research program TRASCO for ADS, the device ORE has been designed and operated at the ENEA C.R. Brasimone in order to study the kinetics of lead oxide reduction by flowing hydrogen, diluted at 3% by volume in argon. From the measurement of water generated by the reaction between hydrogen and lead oxide, the rate of PbO reduction was determined and a first estimation of the activation energy was found. The experimental results so far achieved in terms of reaction rate, activation energy and kinetic constant, as well as the experimental set-up and the methodology, are here summarised and discussed.     

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).     

Mechanical properties of martensitic steels after exposure to flowing liquid metals

This paper describes the tensile test results of martensitic steels EUROFER 97 and T91 after exposure to flowing Pb-17Li and Pb-55.5Bi alloy, respectively. The corrosion tests were performed in LIFUS II loop (Pb-17Li) and LECOR loop (Pb-55.5Bi) located in the ENEA Brasimone Center. Tensile test were carried in SYNTECH test machine under Ar atmosphere with an extension rate of 2 mm/min, and test temperature for EUROFER and T91 steels were at 480 and 400 °C respectively. The exposure of EUROFER steel to the liquid Pb-17Li did not affect its mechanical properties, while the ductility of T91 steel was deteriorated after exposed to flowing liquid Pb-55.5Bi.     

Testing and qualification of CIRCE instrumentation based on bubble tubes

The paper is focused on the development and the qualification of the instrumentation suitable for the level and the differential pressure measurements in the experimental facility CIRCE. This large-scale facility is installed at the ENEA Brasimone Centre for studying the fluid-dynamic and operating behavior of ADS reactor plants. This a rather challenging objective since the facility adopts the molten lead bismuth eutectic (LBE) alloy as a coolant and, at present, instrumentation qualified for operating under such conditions is scarce or does not exist. Bubble tubes have been installed in CIRCE to transfer pressure signals from the LBE to differential pressure cells operating with gas at room temperature. The bubble tube is a simple measuring device, but its use in LBE must be carefully assessed. Therefore, preliminary tests of bubble tubes in representative conditions have been carried out in a smaller test section. Experimental tests were performed at several temperatures, with LBE in stagnant conditions. The results obtained in these tests, aiming at checking the performance of the bubble tubes adopted in measuring pressure, differential pressure and level in the CIRCE facility, are discussed here. The obtained information will allow to calibrate the related measuring systems and to verify the accuracy and repeatability of the measurements, as a function of the injected gas flow rate, the tube diameter and the geometry of the tube exit section.     

Experimental and computational investigation of LBE–water interaction in LIFUS 5 facility

The interaction between heavy liquid metal, such as LBE, and pressurized water has been analyzed at ENEA Brasimone Research Centre, in order to investigate the evolution of this phenomenon over a wide range of conditions. The study includes an experimental campaign on LIFUS 5 facility and a numerical simulation activity performed with SIMMER III code.The first test of the experimental program was carried out injecting water at 7 MPa and 235 °C in a reaction vessel containing LBE at 350 °C. A pressurization up to 8 MPa was observed in the test section during the short term (about 2 s) of the transient.In the post-test analysis performed with SIMMER III code, two different geometrical models were developed in order to reproduce in the best manner the experimental results and, therefore, to confirm the code's capabilities of reproducing the phenomenology of the LBE–water interaction.The data calculated through both models agreed in a good way with the experimental results, despite the necessary simplifications adopted in the models due to the 2D features of the code.     

西安脉冲堆反应性引入事故瞬态安全特性分析

为研究西安脉冲堆（XAPR）在意外引入反应性且停堆系统失效事故下的瞬态安全特性,本文基于XAPR的结构和运行特点,建立了适用于XAPR的瞬态热工水力分析模型,并开发了用于XAPR安全特性分析的瞬态热工水力程序TSAC-XAPR。利用TSAC-XAPR程序对反应性引入事故进行模拟计算,结果表明：当XAPR在额定功率范围内运行时,发生反应性引入事故后,堆芯能依靠自身的固有反馈机制使脉冲堆重新达到稳定运行状态;当运行功率过高尤其是超过临界值时,反应性引入事故将导致脉冲堆关键热工水力参数发生振荡,无法再次达到稳态。此外,不同反应性引入方式将影响堆芯参数在反应性引入过程中的变化趋势,但并不影响其最终稳态值。     

Transient Safety Characteristic Analysis of Xi’an Pulsed Reactor during Reactivity Insertion Accident

In order to study the transient safety characteristics of Xi’an Pulsed Reactor (XAPR) when unexpected reactivity insertion accident happened and shutdown system failed, the main mathematical models were established based on the specific core structure and operation conditions of XAPR. Meanwhile, a transient thermal-hydraulic code called TSAC-XAPR was developed to analyze the safety characteristics of XAPR. The TSAC-XAPR code was then used to simulate the reactivity insertion accident of XAPR. The calculation results indicate that when XAPR operating under rated power, reactor can reach a new steady state for reactivity insertion accident, depending on its inherent feedback mechanism. When XAPR operating under high power, especially above the critical power, key thermal-hydraulic parameters of reactor will tend to oscillate and can’t reach a steady state again for reactivity insertion accident. Besides, it is also found that different reactivity insertion modes will only affect the variation trend during the phase of reactivity insertion instead of the final value at steady state.     

Technological review of the HRP manufacturing process R&D activity

ENEA and Ansaldo Nucleare S.p.A. have been deeply involved in the European International Thermonuclear Experimental Reactor (ITER) R&D activities for the manufacturing of high heat flux plasma-facing components (HHFC), and in particular for the inner vertical target (IVT) of the ITER divertor.This component has to be manufactured by using both armour and structural materials whose properties are defined by ITER. Their physical properties prevent the use of standard joining techniques. The reference armour materials are tungsten and carbon/carbon fibre composite (CFC). The cooling pipe is made of copper alloy (CuCrZr-IG).During the last years ENEA and Ansaldo have jointly manufactured several actively cooled monoblock mock-ups and prototypical components of different length, geometry and materials, by using innovative processes: HRP (hot radial pressing) and PBC (pre-brazed casting).The history of the technical issues solved during the R&D phase and the improvements implemented to the assembling tools and equipments are reviewed in the paper together with the testing results.The optimization of the processes started from the successful manufacturing of both W and CFC armoured small scale mockups thermal fatigue tested in the worst ITER operating condition (20 MW/m²) through the achievement of record performances obtained from a monoblock medium scale mockup.On the base of these results ENEA-ANSALDO participated to the European programme for the qualification of the manufacturing technology to be used for the procurement of the ITER divertor IVT, according to the F4E specifications. A divertor inner vertical target prototype (400 mm total length) with three plasma facing component units, was successfully tested at ITER relevant thermal heat fluxes.Now, ANSALDO and ENEA are ready to face the challenge of the ITER inner vertical target production, transferring to an industrial production line the experience gained in the development, optimization and qualification of the PBC and HRP processes.     

Manufacturing,testing and post-test examination of ITER divertor vertical target W small scale mock-ups

ENEA is involved in the International Thermonuclear Experimental Reactor (ITER) R&D activities. During the last years ENEA has set up and widely tested a manufacturing process, named Hot Radial Pressing (HRP), suitable for the construction of high heat flux plasma-facing components, such as the divertor targets.In the frame of the EFDA contract six mock-ups were manufactured by HRP in the ENEA labs using W monoblocks supplied by the Efremov Institute in St. Petersburg, Russian Federation and IG CuCrZr tubes.According to the technical specifications the mock-ups were examined by ultrasonic technique and after their acceptance they were delivered to the Efremov Institute TSEFEY-M e-beam facility for the thermal fatigue testing. The test consisted in 3000 cycles of 15 s heating and 15 s cooling at 10 MW/m² and finally 1000 cycles at 20 MW/m².After the testing the ultrasonic non-destructive examination was repeated and the results compared with the investigation performed before the testing.A microstructure modification of the W monoblock material due to the overheating of the surfaces and the copper interlayer structure modification were observed in the high heat flux area.The leakage points of the mock-ups that did not conclude the testing were localized in the middle of the monoblock while they were expected between two monoblocks.This paper reports the manufacturing route, the thermal fatigue testing, the pre and post non destructive examination and finally the results of the destructive examination performed on the monoblock small scale mock-ups.     

物理-热工耦合对超临界水堆系统特性的影响分析

物理-热工耦合是超临界水堆系统分析的关键问题之一。以日本超临界水冷热堆Super LWR的堆芯设计为例,借助Dragon编制中子截面数据库,建立双群中子扩散方程计算模块,联系同时建立的热工计算模块,得到超临界水堆的物理-热工耦合计算模型。通过对比稳态与瞬态工况下耦合前、后的热工工况,分析物理-热工耦合条件下的超临界水堆系统热工特性。结果表明：在稳态工况下,物理-热工耦合将导致内、外组件堆芯功率峰值沿轴向发生明显偏移,使得部分节点的包壳温度升高,但包壳最高温度降低;在瞬态工况下,物理-热工耦合将导致堆芯包壳最高温度的发生位置有所改变。发生给水加热丧失瞬态后,在某一时刻,外部组件的包壳最高温度将转而超过内部组件的包壳最高温度。可见,物理-热工耦合对包壳最高温度的大小和发生位置均可能产生明显影响。计算分析可为超临界水堆瞬态及安全分析提供相应理论参考。     

Semi 2D modeling of the horizontal steam generator PGV-1000 using the RELAP5 code

Developing a reliable thermal-hydraulic model of the steam generator is an essential process in the steady state and transient analysis for the Pressurized Water Reactor type of the Nuclear Power Plants. This paper provides a semi two dimensional thermal-hydraulic model of the PGV-1000 horizontal steam generator using the RELAP5 code. Applying the qualified nodalization and the cross-flow effects are some of the advantages in the present model. The obtained results from the RELAP5 steady state analysis showed a reasonable agreement with the Bushehr NPP Final Safety Analysis Reports (FSAR).