The upgraded laser in vessel viewing system (IVVS) for ITER

A prototype of a laser in vessel viewing and ranging system was developed at ENEA laboratories in Frascati, it uses the amplitude modulated laser radar concept and it is conceived to withstand the severe ITER conditions. The in vessel viewing system (IVVS) probe has been designed and built to perform sub-millimetric three-dimensional images inside ITER; it is based on an intrinsically radiation resistant concept and architecture. A first characterization of the IVVS probe under room conditions was made at Frascati Labs on a full-scale mock-up of the ITER first wall panel (FWP) and of the divertor vertical target. The first characterization demonstrated that an upgrade of the IVVS performances was necessary especially to cope with divertor surfaces made of carbon material, which is highly absorbent from optical point of view.The paper describes the new developments of IVVS prototype to increase range measurement performances that was obtained increasing the modulation frequency, the optical efficiency of the probe optics and the laser power. A new test campaign has been carried out on the upgraded IVVS and a compact characteristic curve describing its performances has been found both in mathematical and graphical form. As far as viewing is concerned, the system has confirmed the sub-millimetric viewing resolution, reaching in the worst cases ～1 mm of resolution. The image quality was excellent in almost all the cases. The range measurement performance of IVVS system has been strongly upgraded reducing the standard deviation of range measure of a factor varying from 6 to 12. The increased performance allows measuring surface shapes and erosion on first wall tiles and divertor also for inclination angles completely outside the previous IVVS characteristics.     

Containment vessel design and practice

This paper attempts to review the state of the art of methods of analysis and design of the concrete containment vessels required for BWR and PWR. A step-by-step critical appraisal of the existing work is given. Elastic, inelastic and cracking conditions under extreme loads are fully discussed. Problems associated with these structures are highlighted. A three-dimensional finite element analysis given by the author is included to cater for service, overload and dynamic cracking of such structures. Missile impact and seismic effects are included in this work. The second analysis is known as the limit state analysis, which is given to design such vessels for any kind of load.Two existing vessels in reinforced and pre-stressed concrete are examined. Substantial calculations are given in order to assess their behaviour. Both these original analyses were developed by the author and are given in the appendices. They have been fully tested on four other vesels and two models. Due to limitations of space, some of these details could not be revealed. A brief explanation is given regarding the computer programs supporting the above analysis.     

Thermal-fluid assessment of the design options for reactor vessel cooling in a prismatic core VHTR

The design of the reactor pressure vessel is an important issue in the VHTR design due to its high operating temperature. The extensive experience base in Light Water Reactor makes SA508/533 steel emerge as a strong candidate for the VHTR reactor vessel but requires maintaining the vessel temperature below the ASME code limit. To meet the temperature requirement, three types of vessel cooling options for a prismatic core VHTR are considered: an internal vessel cooling, an external vessel cooling, and an internal insulation. The performances of the vessel cooling options are evaluated by using a system thermo-fluid analysis code and a commercial computational fluid dynamics code during normal operation and accidents. The results suggested that the internal vessel cooling with the modified inlet flow path will be a promising option. The external cooling option does not ensure an effective cooling of the RPV. The insulation option provides an effective reduction of the RPV temperature in the normal and accident conditions but reduces the fuel safety margin during the accidents, requiring careful consideration before the implementation.     

Reactor pressure vessel design and practice

This paper attempts to review the state of the art of methods of analysis and design of the concrete reactor pressure vessels and their components. Existing vessels have been examined for elastic, inelastic and cracking conditions. The results obtained from the analyses given in the appendices are well collaborated with those available from the published experimental tests on models and from site monitoring. The text is provided with an up-to-date comprehensive bibliography. The analytical derivations given in the appendices are based on the authors own research.     

ITER Upper Port Plug handling cask system assessment and design proposals

The current design of the ITER cask for Upper Port Plugs has been evaluated. Careful reduction of the number of mechanical degrees of freedom is an opportunity to relax the tolerances in the design, resulting in cost reduction and reliability increase. A new kinematical design for the tractor module has a higher stiffness to weight ratio, reduces actuator forces by a factor four and minimizes cross-talk between lift and rotation motion. Non-cantilevered handling is recommended to reduce wheel loads on the tractor by a factor six and to simplify guidance. At the system level the tubular guide (TG) is proposed, a semi-permanent 3.5 m long tube which is an extension of the Upper Port. Cask docking is simplified and the risk of the cask tilting is prevented. Redesigning the system concept is recommended and the TG looks promising. Since a system level redesign impacts the external interfaces, overall feasibility has to be investigated.     

Fracture assessment of a PWR pressure vessel

The Second Report of the Marshall Study Group [1] reviewed the design of a pressure vessel to identify those regions which were most sensitive to the expected loading conditions. The design and loading conditions, including the pressure test, were also reviewed. A stress analysis data base of sufficient detail was established to permit a full fracture analysis to be carried out. After reviewing currently available methods for fracture assessment the CEGB R6 method was adopted. Selected normal and upset, emergency and fault transients were considered. Initiation defect sizes for normal and upset transients were generally large. In the beltline and nozzle shell course, for example, the minimum initiating height for an extended line defect was approximately 70 mm. For emergency and fault conditions, providing upper shelf material properties apply, the height to initiate semi-elliptical defects exceeds 25 mm and cracks substantially larger than this will extend by only small amounts in a ductile manner and remain stable. If during the large LOCA upper shelf material properties cannot be maintained semi-elliptic defects 6 to 10 mm high may propagate transversely parallel to the inside surface of the vessel. This may be limited by warm prestressing.     

Optimum shape design of pressure vessel and nozzle junction

Thick spherical pressure vessels are frequently used in nuclear structural applications. The design of the junction of pipe nozzles and vessels needs careful consideration. In this paper the optimum shape design of the junction is considered. Minimization of the stress concentration factor is considered as the objective function. The junction is completely defined by a set of four curves which in turn depend upon four design variables. The stresses in the junction due to internal pressure are evaluated by finite element analysis using ring-shaped isoparametric elements. Stresses are sampled at a number of selected points along the boundary of the junction. The optimization of the nonlinear programming problem is carried out by an improved move limit method of sequential linear programming. Optimum shapes for internal pressure loading are presented for different transition lengths. The effect of axial load on the nozzle of optimum shapes is studied for a few cases. Detailed discussion of the results is presented.     

Final design of ITER vacuum vessel thermal shield

The final design of ITER vacuum vessel thermal shield (VVTS), which is planned to be procured completely by Korea, has been implemented after the procurement arrangement was signed. In this paper, the design and the supporting analysis are described for the key components of the VVTS such as joint, panel, support, and stopper. The VVTS design is revised and finalized based on the manufacturing feasibility, interface requirement and assemble feasibility. The inboard and the outboard supports of VVTS are designed in detail considering structural rigidity and assemble feasibility. The shape of in-pit joint, which is installed every 40° sector in toroidal direction for compensation of possible misalignment during sector assembly, is determined. Three types of joints are developed in accordance with their locations and assemble feasibilities are checked through the R&D. Stopper design is developed in order to prevent direct contact against adjacent components such as vacuum vessel and magnets. Structural rigidity of the whole VVTS is also validated by finite element analysis under various kinds of operating conditions, such as deadweight, electro-magnetic load, seismic load and load combinations.     

Structural integrity assessment of the reactor pressure vessel under the external reactor vessel cooling condition

Since the suggestion of external reactor vessel cooling (ERVC), the effects of melting and cooling on the response of structural integrity of the reactor pressure vessel (RPV) under core melting accident conditions have been investigated. To investigate the initial behavior of RPV lower head and the effects of analysis conditions on the structural integrity of RPV, the transient analysis is utilized considering the transient state. To obtain an analogy with real phenomena, the material properties were determined by combining and modifying the existing results considering phase transformation and temperature dependency. The temperature and stress analyses are performed for core melting accident by using ABAQUS. Finally, the potential for vessel damage is discussed using the Larson-Miller curve and damage rule. In addition, the results by transient analysis are compared with those by steady state analysis and the effects of analysis conditions on structural integrity are reviewed.     

Elastic-plastic fracture-mechanics assessment of a CrMoNi pressure vessel steel

This program involved the testing both of SENB- and CT-specimens and model vessels from pressure vessel steel 18CrMoNi5.4 V with the aim of studying specimen size and flaw size effects on the toughness behaviour. A comparison was made between the R-curves J-Δa of usual specimens in fracture-mechanics investigation (CT, SENB, partly 20% side-grooved) and Charpy specimens with static and dynamic loading. The crack blunting line was determined experimentally from SZW-measurements and used as a reference line for determining the crack initiation point. The R6-procedure was used to estimate the influence of flaw size in cylindrical model vessels with axial surface flaws loaded by internal pressure. A failure curve was constructed for option 2 with regard to the material deformation properties by the Ramberg-Osgood law.     

Interaction of package design development and package design safety assessment

Abstract

In his plenary presentation at PATRAM 2010, Professor Bernhard Droste of BAM (Federal Institute for Materials Research and Testing), Berlin, Germany, reviewed recent developments in package design and safety assessment.     

Status of design and procurement activities in JT-60SA

The JT-60SA experiment is one of the three projects to be undertaken in Japan as part of the Broader Approach Agreement, conducted jointly by Europe and Japan, and complementing the construction of ITER in Europe. It is a fully superconducting tokamak capable of confining break-even equivalent deuterium plasmas with equilibria covering high plasma shaping with a low aspect ratio at a maximum plasma current of I_p = 5.5 MA. In late 2007 the BA Parties, prompted by cost concerns, asked the JT-60SA Team to carry out a re-baselining effort with the purpose to fit in the original budget while aiming to retain the machine mission, performance, and experimental flexibility. Subsequently the Integrated Project Team has undertaken a machine re-optimization followed by engineering design activities aimed to reduce costs while maintaining the machine radius and plasma current. This effort led the Parties to the approval of the new design in late 2008 and hence final design and procurement activities have commenced. The paper will describe the process leading to the re-baselining, the resulting final design and technical solutions and the present status of procurement activities.     

Analytical assessment of an inclad defect in a PWR vessel belt line

The French approach to the assessment of the integrity of PWR vessels requires, in particular, that existence of large margins with respect to fast fracture shall be demonstrated for all kinds of defects which can be produced during manufacturing, taken with envelope sizes. The case of defects in the cladding with one tip against the base metal interface raises several difficult problems, mainly on the effect of residual stresses in the cladding, and on the choice of a relevant criterion for the risk of initiating cleavage cracking in the base material. The behaviour of a typical defect has been computed with elastic-plastic analyses and the criteria of the local approach of fracture: the effect of residual stresses is negligible and the margins with respect to fast fracture are much larger than those indicated previously by LEFM computations with plasticity corrections. The values obtained ensure that the integrity of the vessel would not be affected if such defects had occurred during manufacturing.     

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.     

An assessment of acoustic emission for nuclear pressure vessel monitoring

Recent research has greatly improved our understanding of the basic mechanisms of deformation and fracture that generate detectable acoustic emission signals in structural steels. A critical review of the application of acoustic emission (AE) to the fabrication, proof testing and in-service monitoring of nuclear pressure vessels is presented in the light of this improved understanding. The detectability of deformation and fracture processes in pressure vessel steels is discussed, and recommendations made for improving source location accuracy and the development of quantitative source assessment techniques.

Published data suggest that AE can make an important contribution to weld fabrication monitoring, and to the detection of defects in lower toughness materials during vessel proof testing. In high toughness materials, however, the signals generated during ductile crack growth may frequently be too weak for reliable detection. The feasibility of AE for continuous monitoring has not yet been adequately demonstrated because of high background noise levels and uncertainty about AE signal strengths from the defect growth processes that occur in service. In-service leak detection by AE shows considerable promise.

It is recommended that further tests are carried out with realistic defects, and under realistic conditions of loading (including thermal shock and fatigue) and of environment.     

Design approach of the ARIES-AT power core and vacuum vessel cost assessment

The complexity of fusion power plants require the integration of many diverse and important system requirements to achieve a design approach that is viewed as a commercially viable electric plant. The ARIES-AT power core design builds upon a history of fusion power core designs that evolve along with physics and engineering advances. The baseline design point is optimized for maximum performance and minimum capital cost based upon the ARIES systems code results, along with physics and engineering analyses. The ARIES-AT power core is designed to be quick and easily maintainable to achieve high plant availability. A key element to achieve the high availability is the integration of the core elements with the design of the vacuum vessel. The vacuum vessel design is developed in more detail to assure the key assembly and maintenance features could be realized at an affordable cost.     

Overview on R&D and design activities for the ITER core charge exchange spectroscopy diagnostic system

The ITER core charge exchange recombination spectroscopy (core CXRS) diagnostic system is designed to provide experimental access to various measurement quantities in the ITER core plasma such as ion densities, temperatures and velocities. The implementation of the approved CXRS diagnostic principle on ITER faces significant challenges: First, a comparatively low CXRS signal intensity is expected, together with a high noise level due to bremsstrahlung, while the requested measurement accuracy and stability for the core CXRS system go far beyond the level commonly achieved in present-day fusion experiments. Second, the lifetime of the first mirror surface is limited due to either erosion by fast particle bombardment or deposition of impurities. Finally, the hostile technical environment on ITER imposes challenging boundary conditions for the diagnostic integration and operation, including high neutron loads, electro-magnetic loads, seismic events and a limited access for maintenance. A brief overview on the R&D and design activities for the core CXRS system is presented here, while the details are described in parallel papers.     

The HTR module pressure vessel unit, design criteria and safety philosophy

A modular-helium-cooled high temperature reactor system for the cogeneration of electricity and process heat has been developed by Siemens—Interatom.Design, manufacture and operation of the pressure vessel unit will conform to German nuclear codes and standards for LWR's, some deviations or peculiarities for their application to HTR's are unavoidable. These are for instance:

• - The main steam nozzle, through which the steam line at 530°C penetrates the steam generator pressure vessel with a nominal design temperature of 350°C.

• - The pressure test concept in which the preservice pressure test will be performed in complete accordance with the codes and standards at 1.3 times the design pressure of 70 bar using water. Afterwards, the presence of graphite structures, ceramic insulation and, of course, the pebble bed core has to be considered. Pneumatic pressure tests are performed at 1.1 times design pressure accompanied by more detailed ultrasonic examinations.

• - The position of operational material irradiation surveillance specimens has to be chosen carefully. Design postulates concerning the incrase of ΔRT_NDT will pe confirmed in a separate program.

In general, the requirements of the assured safety concept, aimed to rule out catastrophic failure of the pressure vessel unit during lifetime are fulfilled.