Structural analyses for equatorial and lower ports of the ITER vacuum vessel

To investigate the structural integrity of the ITER vacuum vessel (VV) and ports, the structural analyses of the regular equatorial and the lower remote handling (RH) ports have been performed. The advanced design of the equatorial regular port adopting a pure friction type flange has been recommended as a reference design by the ITER International Organization. The structural integrity of the equatorial port flange, sealing unit, and connecting duct has been reviewed by conducting nonlinear finite element analyses. The advanced design of the regular equatorial port flange with proper pretension is acceptable in the structural design point of view.From the local analyses for a connecting duct and a sealing unit, it has been found that the stresses are less than the allowable values.The structural analyses of the lower RH port have been also performed to verify the capability for supporting the VV. Since high local stress occurs at the gusset and supporting block, the case study for the lower port has been conducted to mitigate the stress concentration and to modify the component design. The strength of the lower RH port structures can be improved by the design modification of poloidal and toroidal gusset.     

Thermo-hydraulic analysis of the generic equatorial port plug design

The port-based ITER diagnostic systems are housed primarily in two locations, the equatorial and upper port plugs. The port plug structure provides confinement function, maintains ultra-high vacuum quality and the first confinement barrier for radioactive materials at the ports. The port plug structure design, from the ITER International Organisation (IO), is cooled and heated by pressurized water which flows through a series of gun-drilled water channels and water pipes. The cooling function is required to remove nuclear heating due to radiation during operation of ITER, while the heating function is intended to heat up uniformly the machine during baking condition. The work presented provides coupled thermo-hydraulic analysis and optimization of a Generic Equatorial Port Plug (GEPP) structure cooling and heating system. The optimization performed includes positioning, minimization of number and size of gun drilled channels, complying with the flow and functional requirements during operating and baking conditions.     

CFD analysis of the ITER first wall 06 panel. Part II: Thermal-hydraulics

The computational fluid dynamics (CFD) analysis of the FW06 panel of the ITER shielding blanket is presented in two companion papers. In this Part II we concentrate on the thermal-hydraulics of the water coolant, driven by the nuclear volumetric and plasma surface heat loads discussed in Part I. Both the detailed steady state analysis of a single cooling channel and the coarse transient analysis of the whole panel are considered. The compatibility of the hot spots with the maximum recommended temperatures for the different materials is confirmed. The heat transfer coefficient between coolant and walls is obtained post-processing the results of the simulation and compared with the results of available correlations, which may be used for simpler analyses: in the fully developed flow regions of the cooling pipes, it turns out to be well approximated by the Sieder–Tate correlation. The operation margin with respect to the critical heat flux is also computed and turns out to be sufficiently large compared with the design limit.     

Preliminary thermal-hydraulic and safety analysis of China DFLL-TBM system

China has proposed the dual-functional lithium-lead (DFLL) tritium breeding blanket concept for testing in ITER as a test blanket module (TBM), to demonstrate the technologies of tritium self-sufficiency, high-grade heat extraction and efficient electricity production which are needed for DEMO and fusion power plant. Safety assessment of the TBM and its auxiliary system should be conducted to deal with ITER safety issues directly caused by the TBM system failure during the design process. In this work, three potential initial events (PIEs) – in-vessel loss of helium (He) coolant and ex-vessel loss of He coolant and loss of flow without scram (LOFWS) – were analyzed for the TBM system with a modified version of the RELAP5/MOD3 code containing liquid lithium-lead eutectic (LiPb). The code also comprised an empirical expression for MHD pressure drop relevant to three-dimensional (3D) effect, the Lubarsky–Kaufman convective heat transfer correlation for LiPb flow and the Gnielinski convective heat transfer correlation for He flow. Since both LiPb and He serve as TBM coolants, the LiPb and He ancillary cooling systems were modeled to investigate the thermal-hydraulic characteristic of the TBM system and its influence on ITER safety under those accident conditions. The TBM components and the coolants flow within the TBM were simulated with one-dimensional heat structures and their associated hydrodynamic components. ITER enclosures including vacuum vessel (VV), port cell and TCWS vault were also covered in the model for accident analyses. Through this best estimate approach, the calculation indicated that the current design of DFLL-TBM and its auxiliary system meets the thermal-hydraulic and safety requirements from ITER.     

Neutronic analysis of iter cryopump system

The ITER Vacuum Vessel has upper, equatorial and lower port structures. The bottom ports are dedicated to the divertor replacement (five ports) and to vacuum pumping by means of cryopumps (four ports). The latest cryopump port design is more complex as it has a pump with a direct view of the vessel (upper cryopump) and a second pump at the end of a branch port (lower cryopump).3D neutronic analyses have been performed in order to study the radiation conditions in and around the port system. In detail, nuclear heating on the cryopump has been calculated updating previous analysis performed in 2003 [L. Petrizzi, ITER CTA Detailed Neutronic Analyses, Final Report on contract EFDA/01-633 ENEA ref NE-VV-R-001 April 2003. Also included in Nuclear Analyis Report NAR ITER ref document G 73 DDD 2W 0.2 (v2.0) March 2006]. Calculations have been performed by means of MCNP 5 Monte Carlo code supplied with FENDL 2.1 library. In this work a new 40° model of ITER has been used in which full details of the cryopump system and remote handling ports have been included as well as the updated divertor components.The paper will present the neutronics results. They consist of nuclear heating on cryopump components; a map of dpa and helium production is provided as well.Gamma doses after shutdown have been calculated around the port flange to have an idea of the possible dose to which the eventual operator will be subject and to plan adequately manual operations.The cryopump is located at a distance of almost 5 m from the mouth of the divertor port and it is 3 m long. Calculations of such deep penetration problem are very challenging require special variance reduction techniques with Monte Carlo codes in order to use in an efficient way the computer resources. These will be described.     

Design and construction of Alborz tokamak vacuum vessel system

The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. At the heart of the tokamak is the vacuum vessel and limiter which collectively are referred to as the vacuum vessel system. As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. The VV systems need upper and lower vertical ports, horizontal ports and oblique ports for diagnostics, vacuum pumping, gas puffing, and maintenance accesses. A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. Basic structure analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. Stresses at general part of the VV body are lower than the structure material allowable stress (117 MPa) and this analysis show that the maximum stresses occur near the gravity support, and is about 98 MPa.     

Evaluation of electromagnetic loads on various design options of the ITER diagnostic upper port plug during plasma disruptions

Electromagnetic (EM) loads due to eddy current and halo current during plasma disruptions are evaluated for the ITER diagnostic upper port plug. To reduce strong EM loads acting on the port plug fixed to the vacuum vessel like a cantilever beam, three design options have been considered: removal of the diagnostic first wall, slitting of the diagnostic shield module and recess of the port plug. The main focus of the present study is to examine the efficacy of these options in terms of EM loads on the upper port plug. It is found that making slits is more effective than removing the first wall. It is also shown that the upper port plug needs to be recessed to reduce the EM load induced by halo current.     

Engineering of diagnostic equatorial port plugs for ITER

Within the ITER vacuum vessel, there are a significant number of diagnostics, measuring items such as plasma density, temperature and impurities; and providing a visible image of the ITER plasma. Since reliable diagnostic measurements are critical to the successful operation of ITER, robust structural design of the diagnostic supports, or port plugs, is also essential. The port plugs are substantial steel structures, mounted in both the equatorial and upper ports on the vacuum vessel. They not only support the diagnostics, but also provide functions of baking, cooling, and neutron shielding.Significant progress has been made in the mechanical design of the port plugs, culminating in the proposal of a new conceptual design, which uses the lid of the port plug as a structural member. This allows the port plug's mass to be more efficiently distributed, providing additional space for diagnostics, and better neutron shielding. A critical aspect of the design has been to provide a suitable interface between the lid and body of the structure which will support all of the structural loads which may be applied to the port plug. The lid also allows easy access to the diagnostic components when maintenance is required.Analyses have been carried out in support of the proposed changes. Structural analysis indicates that the wall thickness of the port plug could be reduced from 130 mm to 40 mm. Thermal analysis has demonstrated that the cooling and baking requirement for the port plug structure is less challenging than originally thought, and hence could be carried out in a simpler fashion. Neutronics analysis has led to a better understanding of the impact of different shielding materials and cavities through the contents of the port plug, and show that it may be possible to reduce the shielding thickness from 2000 mm to 1000 mm. Further electromagnetic analysis has been carried out demonstrating that modelling the effect of plasma movement will not affect the resultant loads by more than 20%, and that the originally defined port plug loads were probably conservative.     

超临界压力水冷包层方案第一壁的热与应力分析

针对超临界水冷包层中第一壁的运行工况,利用数值计算软件ANSYS中CFX和Workbench两个模块对第一壁结构中的固体域和流体域进行数值分析研究。对比矩形管道和圆形管道内传热及热应力分布发现,矩形管道四个角域强化了壁面流体和主流流体的动量和热量的交换,使传热性能优于圆形管道,而四个角域的存在也造成了该处的应力集中,使结构材料的最大应力明显高于圆形管道。进一步研究冷却剂流向和冷却管道几何结构参数对第一壁结构温度场和应力场的影响发现,在ITER运行工况下,冷却剂流向影响很小,增大冷却管道直径和减小冷却管道最小壁厚均能改善第一壁结构材料中的最高温度,而这两个几何结构参数对第一壁应力的影响较为复杂。     

Transient analyses for accelerator driven system PDS-XADS using the extended SIMMER-III code

Transient analyses for Preliminary Design Studies of an Experimental Accelerator Driven System (PDS-XADS) were performed with the reactor safety analysis code SIMMER-III, which was originally developed for the safety assessment of sodium-cooled fast reactors and recently extended by the authors so as to describe the XADS specifics such as subcritical core, strong external neutron source and lead–bismuth–eutectic (LBE) coolant. As transient scenarios, the following cases were analyzed in accordance with the PDS-XADS program: spurious beam trip (BT), unprotected beam overpower (UBOP), unprotected transient overpower (UTOP), unprotected loss of flow (ULOF) and unprotected blockage (UBL) in a single fuel assembly. In addition, to cover some core-melt situations and investigate the potential for recriticalities, so-called snap-shot analyses with ad hoc postulated severe blockage conditions were also investigated.The simulation results for BT and UBOP showed that immediate fuel damage might not take place under short-time beam interruption or a 100% increase of the external neutron source. Concerning UTOP, it was found that a reactivity jump of 1 $ would not lead to damage of the fuel and the cladding. The ULOF simulation showed that the remaining natural convection of the coolant would prevent the cladding from disruptions. In the simulation of UBL in a single fuel assembly, it was shown that no cladding failure might be expected, due to the radial heat transfer and the coolant flow in the hexcan gap. Under an artificial suppression of the radial heat transfer for this UBL case, a pin failure occurred in the simulation but subsequent fuel sweep-out into the upper plenum region would bring a reactivity reduction and no power excursion. The severe accident simulations starting from postulated blockage above an already disrupted core showed that a severe recriticality could be avoided by the fuel sweep-out into the dummy-assembly or hexcan gap regions.The present simulation results showed that the current PDS-XADS design has a remarkable resistance against severe transient scenarios even in core-degradation conditions.     

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.     

Analysis and optimization of the manifolds of the HCCB blanket for CFETR

Thermal-hydraulic performance is a challenging issue in helium-cooled ceramic breeder (HCCB) blanket design due to the extremely complicated working environment and the strict limits of materials temperature. The heat loads deposited on the HCCB blanket comprises of severe surface heat flux from plasma and the volumetric nuclear heat from neutron irradiation, which can be exhausted by the built-in cooling channels of the components. High pressure helium with 8 MPa, distributed from the coolant manifolds is employed as coolant in the blanket. The design and optimization of the manifolds configuration was performed to guarantee the accurate flow control of helium coolant. The flow distribution in the coolant manifolds was investigated based on the structural improvement of manifolds aiming at overall uniform mass flow rates and better flow streamline distribution without obvious vortexes. The peak temperature of different functional materials in the blanket under normal operating condition is below allowable material limits. It is found that the components in the current blanket module could be cooled effectively under the intense thermal loads due to the updated design and optimization analysis of manifolds.     

Safety design of Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR)

In Pb–Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb–Bi coolant above the core, and Pb–Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb–Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits.     

Effects of Coolant Flow on Light Water Reactor Fuel Behaviors during Reactivity Initiated Accident

This paper describes the in-pile experimental results to study the influences of coolant flow on fuel behaviors under reactivity initiated accident (RIA) conditions performed in the Nuclear Safety Research Reactor (NSRR). A single PWR type test fuel rod was irradiated by a large neutron pulse in the NSRR to simulate a prompt power excursion of RIA's. The effects of coolant flow were studied at a coolant flow velocity of 0.3~1.8m/s and a coolant temperature of 20~90°C under the atmospheric pressure. It was found that the cooling conditions had considerable influences on fuel thermal behaviors under prompt heat-up. The increase of coolant flow velocity and subcooling enhanced heat transfer coefficient at cladding surface during film boiling, which resulted in large decrease of maximum cladding temperature and film boiling duration, and consequently in the increase of fuel failure threshold energy. The data tendencies are summarized and the influences of coolant flow are discussed with some computer analyses.     

Some Results of an Investigation and Prospects for the Development of a High-Temperature Power Reactor with a Solid Coolant

A validation of the possibility of developing and the basic advantages of a high-temperature nuclear reactor where the first-loop coolant is a solid are presented. The basic requirements for a solid coolant are formulated, a technology for fabricating spherical graphite particles by gas-phase pyrolytic deposition is developed, and three experimental batches are prepared. The experimental facilities for investigating the motion and heat transfer, including coolant flow stability, heat exchange, and durability, are described. The results of a determination of the heat-emission coefficient during the flow of the solid coolant in a 10-mm in diameter circular channel with warming-wall temperatures in the range 373–1073 K and flow velocities 0.1–0.22 m/sec in vacuum, argon, and helium are presented. The requirements for a 500-kW bench model, on which the basic parameters of the nuclear power system with a solid coolant are to be obtained, are formulated. __________ Translated from Atomnaya Energiya, Vol. 99, No. 5, pp. 358–365, November, 2005.     

Thermohydraulic analysis of high-Prandtl-number fluid in complex duct simulating first wall in fusion reactor

For fusion reactors, molten salt is one of the candidates for coolant materials. Molten salt is a high-Prandtl-number fluid; thus, it is necessary to enhance the heat transfer coefficient. It is proposed that rods are inserted into a duct to enhance the heat transfer coefficient. The flow field behind the rod in the duct is visualized to compare experimental data with simulation results. The trends and distributions in the numerical simulation are the same as those in the experiment, and furthermore, the magnitudes of the time and space scales in the numerical simulation are of the same order as those in the experiment. Thermohydraulic numerical analysis confirmed that the heat transfer coefficient is improved by inserting the rod when the fluid is a high-Prandtl-number fluid and the flow field is in the turbulent region. However, it is necessary for the rods to be arranged in the streamwise direction.     

Detailed 3-D nuclear analysis of ITER blanket modules

In ITER, the blanket modules (BM) are arranged around the plasma to provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. As a part of the BM design process, nuclear analysis is required to determine the level of nuclear heating, helium production, and radiation damage in the BM. Additionally, nuclear heating in the VV is also important for assessing the BM design. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40-degree partially homogenized ITER global model. The regions analyzed include BM01, the neutral beam injection (NB) region, and the upper port region. For BM01, the results show that He production meets the limit necessary for re-welding, and the VV heating behind BM01 is acceptable. For the NBI region, the VV nuclear heating behind the NB region exceeds the design limit by a factor of two. For the upper port region, the nuclear heating of the VV exceeds the design limit by up to 20%. The results presented in this work are being used to modify the BM design in the cases where limits are exceeded.     

Turbulence-induced Heat Transfer in PBMR Core Using LES and RANS

This paper introduces the results of numerical simulations on flow fields and relevant heat transfer in the pebble bed reactor (PBR) core. In the core, since the coolant passes a highly complicated random flow path with a high Reynolds number, an appropriate treatment of the turbulence is required. A set of simple experiments for the flow over a circular cylinder with heat transfer was conducted to finally select the large eddy simulation (LES) and k-ω model among the considering Reynolds-averaged Navier-Stokes (RANS) models for PBR application. Using these models, the PBR cores, whose geometries were simplified to the body-centered cubical (BCC) and face-centered cubical (FCC) structures, were simulated. A larger pressure drop, a more random flow field, a higher vorticity magnitude and a higher temperature at the local hot spots on the pebble surface were found in the results of the LES than in those of RANS for both geometries. In cases of the LES, the flow structures were resolved up to the grid scales. Irregular distributions of the flow and local heat transfer were found in the BCC core, while relatively regular distributions for the FCC core. The turbulent nature of the coolant flow in the pebble core evidently affected the fuel surface temperature distribution.