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Toroid field coil shear key installation study, DOE task No. 22

Description: Concepts for fitting and installation of the scissor keys, triangular keys, and truss keys in the ITER Toroidal Field (TF) Coil Assembly were developed and evaluated. In addition, the process of remote removal and replacement of a failed TF coil was considered. Two concepts were addressed: central solenoid installed last (Naka Option 1) and central solenoid installed first (Naka Option 2). In addition, a third concept was developed which utilized the favorable features of both concepts. A time line for installation was estimated for the Naka Option 1 concept.
Date: January 9, 1995
Creator: Jones, C.E.; Meier, R.W. & Yuen, J.L.
Partner: UNT Libraries Government Documents Department

Linear and Nonlinear Study of Fast Particle Excitation of Alfvén Eigemodes

Description: Recent new results concerning toroidicity-induced Alfvén eigenmode (TAE) linear stability and nonlinear amplitude saturation and associated fast ion transport are presented for tokamaks, such as the National Spherical Torus Experiment (NSTX) and the International Thermonuclear Experimental Reactor (ITER), using the numerical codes HINST, NOVA-K, and ORBIT.
Date: November 1, 1998
Creator: Berk, H.L.; Chen, Y.; Cheng, C.Z.; Fu, G.Y.; Gorelenkov, N.N; Gorelenkova, M.V. et al.
Partner: UNT Libraries Government Documents Department

ITER risk workshop facilitator guide

Description: The goal of planning risk management is to make everyone involved in a program aware that risk should be a consideration in the design, development, and fielding of a system. Risk planning is a tool to assess and mitigate events that might adversely impact the program. Therefore, risk management increases the probability/likelihood of program success and can help to avoid program crisis management and improve problem solving by managing risk early in the acquisition cycle.
Date: January 1, 2009
Creator: Medina, Patricia
Partner: UNT Libraries Government Documents Department

ITER risk workshop participant guide

Description: The goal of planning risk management is to make everyone involved in a program aware that risk should be a consideration in the design, development, and fielding of a system. Risk planning is a tool to assess and mitigate events that might adversely impact the program. Therefore, risk management increases the probability/likelihood of program success and can help to avoid program crisis management and improve problem solving by managing risk early in the acquisition cycle.
Date: January 1, 2009
Creator: Medina, Patricia
Partner: UNT Libraries Government Documents Department

Tritium Experience in Large Tokamaks: Application to ITER

Description: Recent experience with the use of tritium fuel in the Tokamak Fusion Test Reactor and the Joint European Torus, together with progress in developing the technical design of the International Thermonuclear Experimental Reactor has expanded the technical knowledge base for tritium issues in fusion. This paper reports on an IEA workshop that brought together scientists and engineers to share experience and expertise on all fusion-related tritium issues. Extensive discussion periods were devoted to exploring outstanding issues and identifying potential R{ampersand}D avenues to address them. This paper summarizes the presentations, discussions, and recommendations.
Date: May 1, 1998
Creator: Skinner, C.H.; Gentile, C.; Hosea, J.; Mueller, D; Gentile, C.; Federici, G. et al.
Partner: UNT Libraries Government Documents Department

Comparison of impurities and time-dependent behavior for the ITER divertor

Description: This a the second part-of an ongoing project to model the divertor plasma for ITER. The UEDGE 2-D edge transport code is used to study the effect of impurities and tilted divertor plates to make a radiative divertor that can prevent excessive heat loads and adequately pump helium produced by fusion reactions in the core. The impurities are modeled using individual charge states with the local concentrations being determined by transport or as a fixed fraction of the hydrogenic ion density. For the multi-species model, helium, beryllium, carbon, and neon impurities are considered separately, together with the majority hydrogenic species, and a comparison is made of impurity spatial distribution and the power radiated at low impurity levels. At moderate to high impurity levels, typically only time-dependent solutions are found which are studied here for neon using both impurity models.
Date: February 25, 1997
Creator: Rensink, M.E.; Rognlien, T.D. & Hua, D.D.
Partner: UNT Libraries Government Documents Department

Development and test of the ITER SC conductor joints

Description: Joints for the ITER superconducting Central Solenoid should perform in rapidly varying magnetic field with low losses and low DC resistance. This paper describes the design of the ITER joint and presents its assembly process. Two joints were built and tested at the PTF facility at MIT. Test results are presented; losses in transverse and parallel field and the DC performance are discussed. The developed joint demonstrates sufficient margin for baseline ITRR operating scenarios.
Date: August 5, 1998
Creator: Gung, C. Y.; Jayakumar, R.; Manahan, R.; Martovetsky, N.; Michael, P.; Minervini, J. et al.
Partner: UNT Libraries Government Documents Department

Physics research needs for ITER

Description: Design of ITER entails the application of physics design tools that have been validated against the world-wide data base of fusion research. In many cases, these tools do not yet exist and must be developed as part of the ITER physics program. ITER`s considerable increases in power and size demand significant extrapolations from the current data base; in several cases, new physical effects are projected to dominate the behavior of the ITER plasma. This paper focuses on those design tools and data that have been identified by the ITER team and are not yet available; these needs serve as the basis for the ITER Physics Research Needs, which have been developed jointly by the ITER Physics Expert Groups and the ITER design team. Development of the tools and the supporting data base is an on-going activity that constitutes a significant opportunity for contributions to the ITER program by fusion research programs world-wide.
Date: July 1, 1995
Creator: Sauthoff, N.R.
Partner: UNT Libraries Government Documents Department

Design, fabrication, and testing of a helium-cooled module for the ITER divertor

Description: The International Thermonuclear Reactor (ITER) will have a single-null divertor with total power flow of 200 MW and a peak heat flux of about 5 MW/m{sup 2}. The reference coolant for the divertor is water. However, helium is a viable alternative and offers advantages from safety considerations, such as excellent radiation stability and chemical inertness. In order to prove the feasibility of helium cooling at ITER relevant heat flux conditions, General Atomics designed, fabricated, and tested a helium-cooled divertor module. The module was made from dispersion strengthened copper, with a heat flux surface 25 mm wide and 80 mm long, designed for twice the ITER divertor heat flux. Different techniques were examined to enhance the heat transfer, which in turn reduced the flow and pumping power required to cool the module. It was concluded that an extended surface was the most practical solution. An optimization study was performed to find the best extended surface parameters. The optimum extended surface geometry consisted of fins: 10 mm high, 0.4 mm thick with a 1 mm pitch. It was estimated to require a pumping power of 150 W to remove 20 kW of power. This is more than an order of magnitude reduction in pumping power requirement, compared to smooth surface. The module was fabricated by electric discharge machining (EDM) process. The testing was carried out at SNLA during August 1993. The testing confirmed the design calculations. The peak heat flux during the test was 10 MW/m{sup 2} applied over a surface area of 20 cm{sup 2}. The pumping power calculated from flow rate and pressure drop measurement was about 160 W, which was less than 1% of the power removed. It is planned to test the module to higher temperature limits and higher heat fluxes during coming months. As a result of ...
Date: August 1, 1994
Creator: Baxi, C.B.; Smith, J.P. & Youchison, D.
Partner: UNT Libraries Government Documents Department

Bolometry for divertor characterization and control

Description: Operation of the divertor will provide one of the greatest challenges for ITER. Up to 400 MW of power is expected to be produced in the core plasma which must then be handled by plasma facing components. Power flowing across the separatrix and into the scrape-off-layer (SOL) can lead to a heat flux in the divertor of 30 MW/m{sup 2} if nothing is done to dissipate the power. This peak heat flux must be reduced to 5 MW/m{sup 2} for an acceptable engineering design. The current plan is to use impurity radiation and other atomic processes from intrinsic or injected impurities to spread out the power onto the first wall and divertor chamber walls. It is estimated that 300 MW of radiation in the divertor and SOL will be necessary to achieve this solution. Measurement of the magnitude and distribution of this radiated power with bolometry will be important for understanding and controlling the nER divertor. Present experiments have shown intense regions of radiation both in the divertor near the separatrix and in the X-point region. The task of a divertor bolometer system will be to measure the distribution and magnitude of this radiation. First, radiation measurements can be used for machine protection. Intense divertor radiation will heat plasma facing surfaces that are not in direct view of temperature monitors. Measurement of the radiation distribution will provide information about the power flux to these components. Secondly, a bolometer diagnostic is a basic tool for divertor characterization and understanding. Radiation measurements are important for power accounting, as a cross check for other power diagnostics, and gross characterisation of the plasma behavior. A divertor bolometer system can provide a 2-D measurement of the radiation profile for comparison with theory and modeling. Finally a bolometer system can provide realtime signals for control of ...
Date: October 1, 1995
Creator: Leonard, A.W.; Goetz, J.; Fuchs, C.; Marashek, M.; Mast, F. & Reichle, R.
Partner: UNT Libraries Government Documents Department

Engineering design of the ITER RF systems

Description: Parallel conceptual design efforts for auxiliary heating systems on ITER are being carried out in both the electron cyclotron range of frequencies (ECRF) and ion cyclotron range of frequencies (ICRF). These systems are required to deliver a minimum of 50 MW of CW power to the plasma for the primary purpose of heating and the secondary purpose of current drive. Current designs of the two systems are presented and the primary design issues are discussed.
Date: December 31, 1994
Creator: Makowski, M.; Bosia, G.; Nagashima, T. & Remsen, D.
Partner: UNT Libraries Government Documents Department

Development and test of the ITER conductor joints

Description: Joints for the ITER superconducting Central Solenoid should perform in rapidly varying magnetic field with low losses and low DC resistance. This paper describes the design of the ITER joint and presents its assembly process. Two joints were built and tested at the PTF facility at MIT. Test results are presented, losses in transverse and parallel field and the DC performance are discussed. The developed joint demonstrates sufficient margin for baseline ITER operating scenarios.
Date: May 14, 1998
Creator: Martovetsky, N.
Partner: UNT Libraries Government Documents Department

Testing of ITER central solenoid coil insulation in an array

Description: A glass-polyimide insulation system has been proposed by the US team for use in the Central Solenoid (CS) coil of the international Thermonuclear Experimental Reactor (ITER) machine and it is planned to use this system in the CS model coil inner module. The turn insulation will consist of 2 layers of combined prepreg and Kapton. Each layer is 50% overlapped with a butt wrap of prepreg and an overwrap of S glass. The coil layers will be separated by a glass-resin composite and impregnated in a VPI process. Small scale tests on the various components of the insulation are complete. It is planned to fabricate and test the insulation in a 4 x 4 insulated CS conductor array which will include the layer insulation and be vacuum impregnated. The conductor array will be subjected to 20 thermal cycles and 100000 mechanical load cycles in a Liquid Nitrogen environment. These loads are similar to those seen in the CS coil design. The insulation will be electrically tested at several stages during mechanical testing. This paper will describe the array configuration, fabrication: process, instrumentation, testing configuration, and supporting analyses used in selecting the array and test configurations.
Date: September 29, 1995
Creator: Jayakumar, R.; Martovetsky, N.N. & Perfect, S.A.
Partner: UNT Libraries Government Documents Department

Disruptions, loads, and dynamic response of ITER

Description: Plasma disruptions and the resulting electromagnetic loads are critical to the design of the vacuum vessel and in-vessel components of the International Thermonuclear Experimental Reactor (ITER). This paper describes the status of plasma disruption simulations and related analysis, including the dynamic response of the vacuum vessel and in-vessel components, stresses and deflections in the vacuum vessel, and reaction loads in the support structures.
Date: December 31, 1995
Creator: Nelson, B.; Riemer, B.; Sayer, R.; Strickler, D.; Barabaschi, P.; Ioki, K. et al.
Partner: UNT Libraries Government Documents Department

ITER vacuum vessel fabrication plan and cost study (D 68) for the international thermonuclear experimental reactor

Description: ITER Task No. 8, Vacuum Vessel Fabrication Plan and Cost Study (D68), was initiated to assess ITER vacuum vessel fabrication, assembly, and cost. The industrial team of Raytheon Engineers & Constructors and Chicago Bridge & Iron (Raytheon/CB&I) reviewed the current vessel basis and prepared a manufacturing plan, assembly plan, and cost estimate commensurate with the present design. The guidance for the Raytheon/CB&I assessment activities was prepared by the ITER Garching Work Site. This guidance provided in the form of work descriptions, sketches, drawings, and costing guidelines for each of the presently identified vacuum vessel Work Breakdown Structure (WBS) elements was compiled in ITER Garching Joint Work Site Memo (Draft No. 9 - G 15 MD 01 94-17-05 W 1). A copy of this document is provided as Appendix 1 to this report. Additional information and clarifications required for the Raytheon/CB&I assessments were coordinated through the US Home Team (USHT) and its technical representative. Design details considered essential to the Task 8 assessments but not available from the ITER Joint Central Team (JCT) were generated by Raytheon/CB&I and documented accordingly.
Date: January 1, 1995
Partner: UNT Libraries Government Documents Department

ITER Building Design (D230-B), Task No. 28. Final report

Description: The International Thermonuclear Experimental Reactor (ITER) Project requires a set of buildings, each with its own distinct function, to support ITER`s mission. The Joint Central Team (JCT) has identified all the buildings in the set and has placed them in an efficient arrangement on the site. The JCT has developed a conceptual layout of each individual building. The buildings have been categorized into two main groups: (1) {open_quotes}Level 1 Buildings{close_quotes} which are on the construction schedule critical path and (2) {open_quotes}Level 2 Buildings{close_quotes} which, while important, are not on the critical path. The buildings are further categorized according to construction material, that is, {open_quotes}reinforced concrete{close_quotes} or {open_quotes}steel-frame on concrete slab{close_quotes}. This Report responds to the Project`s request to perform the initial structural steel design for all the {open_quotes}steel-frame on concrete slab{close_quotes} buildings. Of the twelve (12) {open_quotes}steel-frame on concrete slab{close_quotes} buildings, four (4) are Level 1 and eight (8) are Level 2 Buildings. This Report is a deliverable for the ITER Task Assignment entitled {open_quotes}ITER Buildings Design (D230-B){close_quotes}, also designated as Task No. 28. ITER U.S. Home Team Industrial Consortium members, Raytheon Engineers & Constructors (RE&C) and Stone & Webster Engineering Corporation (SWEC), teamed to perform Task 28. This task commenced in May 1995. It was performed in accordance with the design criteria specified by the ITER-JCT, San Diego Joint Work Site.
Date: December 1, 1995
Partner: UNT Libraries Government Documents Department

Alpha-physics and measurement requirements for ITER

Description: This paper reviews alpha particle physics issues in ITER and their implications for alpha particle measurements. A comparison is made between alpha heating in ITER and NBI and ICRH heating systems in present tokamaks, and alpha particle issues in ITER are discussed in three physics areas: `single particle` alpha effects, `collective` alpha effects, and RF interactions with alpha particles. 29 refs., 4 figs., 4 tabs.
Date: December 31, 1995
Creator: Zweben, S.J.; Young, K.M.; Putvinski, S.; Petrov, M.P.; Sadler, G. & Tobita, K.
Partner: UNT Libraries Government Documents Department

Evaluation of US demo helium-cooled blanket options

Description: A He-V-Li blanket design was developed as a candidate for the U.S. fusion demonstration power plant. This paper presents an 18 MPa helium-cooled, lithium breeder, V-alloy design that can be coupled to the Brayton cycle with a gross efficiency of 46%. The critical issue of designing to high gas pressure and the compatibility between helium impurities and V-alloy are addressed.
Date: October 1, 1995
Creator: Wong, C.P.C.; McQuillan, B.W. & Schleicher, R.W.
Partner: UNT Libraries Government Documents Department

Design of fast tuning elements for the ITER ICH system

Description: The coupling between the ion cyclotron (IC) antenna and the ITER plasma (as expressed by the load resistance the antenna sees) will experience relatively fast variations due to plasma edge profile modifications. If uncompensated, these will cause an increase in the amount of power reflected back to the transmitter and ultimately a decrease in the amount of radio frequency (rf) power to the plasma caused by protective suppression of the amount of rf power generated by the transmitter. The goals of this task were to study several alternate designs for a tuning and matching (T&M) system and to recommend some research and development (R&D) tasks that could be carried out to test some of the most promising concepts. Analyses of five different T&M configurations are presented in this report. They each have different advantages and disadvantages, and the choice among them must be made depending on the requirements for the IC system. Several general conclusions emerge from our study: The use of a hybrid splitter as a passive reflected-power dump [``edge localized mode (ELM)-dump``] appears very promising; this configuration will protect the rf power sources from reflected power during changes in plasma loading due to plasma motion or profile changes (e.g., ELM- induced changes in the plasma scrape-off region) and requires no active control of the rf system. Trade-offs between simplicity of design and capability of the system must be made. Simple system designs with few components near the antenna either have high voltages over considerable distances of transmission lines, or they are not easily tuned to operate at different frequencies. Designs using frequency shifts and/or fast tuning elements can provide fast matching over a wide range of plasma loading; however, the designs studied here require components near the antenna, complicating assembly and maintenance. Capacitor-tuned resonant systems may offer a ...
Date: May 1, 1996
Creator: Swain, D.W. & Goulding, R.H.
Partner: UNT Libraries Government Documents Department

Neutron cameras for ITER

Description: Neutron cameras with horizontal and vertical views have been designed for ITER, based on systems used on JET and TFTR. The cameras consist of fan-shaped arrays of collimated flight tubes, with suitably chosen detectors situated outside the biological shield. The sight lines view the ITER plasma through slots in the shield blanket and penetrate the vacuum vessel, cryostat, and biological shield through stainless steel windows. This paper analyzes the expected performance of several neutron camera arrangements for ITER. In addition to the reference designs, the authors examine proposed compact cameras, in which neutron fluxes are inferred from {sup 16}N decay gammas in dedicated flowing water loops, and conventional cameras with fewer sight lines and more limited fields of view than in the reference designs. It is shown that the spatial sampling provided by the reference designs is sufficient to satisfy target measurement requirements and that some reduction in field of view may be permissible. The accuracy of measurements with {sup 16}N-based compact cameras is not yet established, and they fail to satisfy requirements for parameter range and time resolution by large margins.
Date: December 31, 1998
Creator: Johnson, L.C.; Barnes, C.W. & Batistoni, P.
Partner: UNT Libraries Government Documents Department

Physics of locked modes in ITER: Error field limits, rotation for obviation, and measurement of field errors

Description: The existing theoretical and experimental basis for predicting the levels of resonant static error field at different components m,n that stop plasma rotation and produce a locked mode is reviewed. For ITER ohmic discharges, the slow rotation of the very large plasma is predicted to incur a locked mode (and subsequent disastrous large magnetic islands) at a simultaneous weighted error field ({Sigma}{sub 1}{sup 3}w{sub m1}B{sup 2}{sub rm1}){sup {1/2}}/B{sub T} {ge} 1.9 x 10{sup -5}. Here the weights w{sub m1} are empirically determined from measurements on DIII-D to be w{sub 11} = 0. 2, w{sub 21} = 1.0, and w{sub 31} = 0. 8 and point out the relative importance of different error field components. This could be greatly obviated by application of counter injected neutral beams (which adds fluid flow to the natural ohmic electron drift). The addition of 5 MW of 1 MeV beams at 45{degrees} injection would increase the error field limit by a factor of 5; 13 MW would produce a factor of 10 improvement. Co-injection beams would also be effective but not as much as counter-injection as the co direction opposes the intrinsic rotation while the counter direction adds to it. A means for measuring individual PF and TF coil total axisymmetric field error to less than 1 in 10,000 is described. This would allow alignment of coils to mm accuracy and with correction coils make possible the very low levels of error field needed.
Date: February 1, 1997
Creator: La Haye, R.J.
Partner: UNT Libraries Government Documents Department

Analysis of losses in ITER joints in varying parallel field

Description: One of the options for a design of a Central Solenoid in ITER and other tokamak machines is pancake would modules. In this configuration joints have to be placed in maximum magnetic field with high changing rate. In this condition joints should be designed to have at least the same or larger temperature margin as that for the conductor in the same field. It is argued that joints in parallel field can be designed to meet this requirement along with reasonably low DC resistance. Losses in parallel field are calculated and design features which can suppress AC losses without increasing DC resistance are discussed. Recommendations for low loss, low DC resistance joints are made.
Date: August 8, 1996
Creator: Martovetsky, N.N.
Partner: UNT Libraries Government Documents Department

Improved fusion performance in low-q, low triangularity plasmas with negative central magnetic shear

Description: Fusion performance in DIII-D low-q single-null divertor discharges has doubled as a result of improved confinement and stability, achieved through modification of pressure and current density profiles. These discharges extend the regime of neoclassical core confinement associated with negative or weak central magnetic shear to plasmas with the low safety factor (q{sub 95}{approximately}3) and triangularity ({delta}{approximately}0.3) anticipated in future tokamaks such as ITER. Energy confinement times exceed the ITER-89P L- mode scaling law by up to a factor of 4, and are almost twice as large as in previous single-null cases with 3{le}q{sub 95}{le}4. The normalized beta [{beta}(aB/I)] reaches values as high as 4, comparable to the best previous single-null discharges. Although high triangularity allows a larger plasma current, the fusion gain in these low triangularity plasmas is similar to that of high triangularity double-null plasmas at the same plasma current. These results are encouraging for advanced performance operation in ITER and for D-T experiments in JET.
Date: July 1, 1996
Creator: Strait, E.J.; Casper, T.N. & Chu, M.S.
Partner: UNT Libraries Government Documents Department

ITER vacuum vessel design (D201 subtask 1.3 and subtask 3). Final report

Description: ITER Task No. D201, Vacuum Vessel Design (Subtask 1.3 and Subtask 3), was initiated to propose and evaluate local vacuum vessel reinforcement alternatives in proximity to the Neutral Beam, Radial Mid-Plane, Top, and Divertor Ports. These areas were reported to be highly stressed regions based on the results of preliminary stress analyses performed by the USHT (US Home Team) and the ITER Joint Central Team (JCT) at the Garching JWS (Joint Work Site). Initial design activities focused on the divertor port region which was reported to experience the highest stress intensities. Existing stress analysis models and results were reviewed with the USHT stress analysts to obtain an overall understanding of the vessel response to the various applied loads. These reviews indicated that the reported stress intensities in the divertor port region were significantly affected by the loads applied to the vessel in adjacent regions.
Date: August 1, 1996
Partner: UNT Libraries Government Documents Department