Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518385
I. Clarkson, J. O'toole, R. Watson
The ITER vacuum vessel will be the largest such structure yet designed with a height of 14 m and an outer diameter of 26 m. The vessel must provide a high quality vacuum, high electrical resistivity, and operate at high temperature. The vessel must provide for bakeout, nuclear shielding, support of in-vessel components and access to these. Significant electromagnetic forces act on the vessel especially during a plasma disruption. The vessel is designed as a double walled toroidal shell with poloidal stiffening rings. Construction cost is reduced by fabricating the shell from a series of single curvature plates, 2-4 cm thick, that are fully welded to form a faceted structure. Material selection must consider fabricability, structural properties at temperature and over the life of the machine, and the desire for low activation. Interaction with the selected coolant, especially if it is liquid metal is a consideration. Stress relief operations and the ability to remotely cut and re-weld the vessel are important considerations. Step by step fabrication and assembly sequences were developed and illustrated using computer solid modeling techniques. Final assembly of the vessel at the ITER site considers overall sequence of machine assembly. Final vessel sector weld joint location options include mid TF coil, mid port and just to the side of the ports, which would allow factory fabrication of the more demanding port joint region. Final assembly operations demand that the weight of the vessel be kept low so that the modules can be moved into position for final welding. Nuclear shielding design plays a significant role. The design features solid built-in shield blocks, in difficult to access areas, and bulk shielding using insulated metallic balls, which can be added and removed after the vessel is fully in place. An important part of the design is provision for direction of coolant flow, ensuring adequate thermal control to all regions of the vessel. Port to shell joints consider shielding installation, coolant flow and ease of fabrication. Support of the vessel and the in-vessel components must provide for the thermal expansion experienced while protecting against seismic events. The vessel provides containment for tritium and is important to the overall safety of the facility. Postulated abnormal events must be considered in the design and safety analysis. A set of fabrication development and construction verification mock-up articles and their evaluation is planned prior to the completion of the detail design phase. ITER is in the earliest stages of the design process and today's decisions will form the basis of the detailed design, fabrication and operation.
{"title":"Construction considerations for the ITER vacuum vessel","authors":"I. Clarkson, J. O'toole, R. Watson","doi":"10.1109/FUSION.1993.518385","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518385","url":null,"abstract":"The ITER vacuum vessel will be the largest such structure yet designed with a height of 14 m and an outer diameter of 26 m. The vessel must provide a high quality vacuum, high electrical resistivity, and operate at high temperature. The vessel must provide for bakeout, nuclear shielding, support of in-vessel components and access to these. Significant electromagnetic forces act on the vessel especially during a plasma disruption. The vessel is designed as a double walled toroidal shell with poloidal stiffening rings. Construction cost is reduced by fabricating the shell from a series of single curvature plates, 2-4 cm thick, that are fully welded to form a faceted structure. Material selection must consider fabricability, structural properties at temperature and over the life of the machine, and the desire for low activation. Interaction with the selected coolant, especially if it is liquid metal is a consideration. Stress relief operations and the ability to remotely cut and re-weld the vessel are important considerations. Step by step fabrication and assembly sequences were developed and illustrated using computer solid modeling techniques. Final assembly of the vessel at the ITER site considers overall sequence of machine assembly. Final vessel sector weld joint location options include mid TF coil, mid port and just to the side of the ports, which would allow factory fabrication of the more demanding port joint region. Final assembly operations demand that the weight of the vessel be kept low so that the modules can be moved into position for final welding. Nuclear shielding design plays a significant role. The design features solid built-in shield blocks, in difficult to access areas, and bulk shielding using insulated metallic balls, which can be added and removed after the vessel is fully in place. An important part of the design is provision for direction of coolant flow, ensuring adequate thermal control to all regions of the vessel. Port to shell joints consider shielding installation, coolant flow and ease of fabrication. Support of the vessel and the in-vessel components must provide for the thermal expansion experienced while protecting against seismic events. The vessel provides containment for tritium and is important to the overall safety of the facility. Postulated abnormal events must be considered in the design and safety analysis. A set of fabrication development and construction verification mock-up articles and their evaluation is planned prior to the completion of the detail design phase. ITER is in the earliest stages of the design process and today's decisions will form the basis of the detailed design, fabrication and operation.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"5 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133027716","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518434
M. Stuart, M. J. Knolls, J. Kwan, A. Rawlins, T. Stevens, R. Wells
A surface production D/sup +/ ion source for fusion research is in the process of being developed at Lawrence Berkeley Laboratory. Significant modifications to prototype surface conversion source are described. A "converter" consisting of a water cooled molybdenum substrate with a 1 mm barium surface layer has been developed. The source has been adapted from the filament to the RF discharge. The design and fabrication of the barium converter, the development of RF antennas, and the design of diagnostic tools for evaluating source performance are discussed.
{"title":"Modifications of a barium surface conversion source for operation with RF discharge","authors":"M. Stuart, M. J. Knolls, J. Kwan, A. Rawlins, T. Stevens, R. Wells","doi":"10.1109/FUSION.1993.518434","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518434","url":null,"abstract":"A surface production D/sup +/ ion source for fusion research is in the process of being developed at Lawrence Berkeley Laboratory. Significant modifications to prototype surface conversion source are described. A \"converter\" consisting of a water cooled molybdenum substrate with a 1 mm barium surface layer has been developed. The source has been adapted from the filament to the RF discharge. The design and fabrication of the barium converter, the development of RF antennas, and the design of diagnostic tools for evaluating source performance are discussed.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"77 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133524716","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518420
P. Titus, F. Wong
Global analysis of the out-of-plane behavior is presented. The sensitivity of the global behavior of the tokamak to out-of-plane support structures is described.
给出了平面外行为的全局分析。描述了托卡马克整体性能对面外支撑结构的敏感性。
{"title":"ITER EDA out-of-plane structural design and analysis","authors":"P. Titus, F. Wong","doi":"10.1109/FUSION.1993.518420","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518420","url":null,"abstract":"Global analysis of the out-of-plane behavior is presented. The sensitivity of the global behavior of the tokamak to out-of-plane support structures is described.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"2005 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133440095","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518513
A. Wright, J.C. Allen, W. Cary, T. E. Harris
The first of four Varian 500 kW 110 GHz gyrotrons (VGT-8011) to be used in the new 2 MW 110 GHz electron cyclotron heating system being developed for the DIII-D tokamak was put into test at General Atomics within the last year. This gyrotron has been used to demonstrate the overall system efficiency and to validate the design of individual transmission line components. The first plasma heating observed with a 110 GHz was consistent with the power expected for the greater than 85% transmission efficiency of HE/sub 1.1/ power. A comparison of the General Atomics' TE/sub 15,2/ to HE/sub 1.1/ mode converter with the Vlasov-type mode converter designed by the University of Wisconsin showed similar conversion efficiencies. The overall ECH system performance during the first year of testing will also be discussed.
{"title":"Performance of the DIII-D 110 GHz ECH system during the first year of operations and testing","authors":"A. Wright, J.C. Allen, W. Cary, T. E. Harris","doi":"10.1109/FUSION.1993.518513","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518513","url":null,"abstract":"The first of four Varian 500 kW 110 GHz gyrotrons (VGT-8011) to be used in the new 2 MW 110 GHz electron cyclotron heating system being developed for the DIII-D tokamak was put into test at General Atomics within the last year. This gyrotron has been used to demonstrate the overall system efficiency and to validate the design of individual transmission line components. The first plasma heating observed with a 110 GHz was consistent with the power expected for the greater than 85% transmission efficiency of HE/sub 1.1/ power. A comparison of the General Atomics' TE/sub 15,2/ to HE/sub 1.1/ mode converter with the Vlasov-type mode converter designed by the University of Wisconsin showed similar conversion efficiencies. The overall ECH system performance during the first year of testing will also be discussed.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"122 51","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"131912668","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518473
S. Ramakrishnan, O. N. Bowen, T. O'connor, J. Edwards, N. Fromm, R. Hatcher, R. Newman, G. Rossi, T. Stevenson, A. von Halle
The Tokamak Physics Experiment (TPX) will utilize to the maximum extent the existing Tokamak Fusion Test Reactor (TFTR) equipment and facilities. This is particularly true for the TFTR Neutral Beam (NB) system. Most of the NB hardware, plant facilities, auxiliary sub-systems, power systems, service infrastructure, and control systems can be used as is. The major changes in the NB hardware are driven by the new operating duty cycle. The TFTR Neutral Beam was designed for operation of the sources for 2 seconds every 150 seconds. The TPX requires operation for 1000 seconds every 4500 seconds. During the Conceptual Design Phase of TPX every component of the TFTR NB Electrical Power System was analyzed to verify whether the equipment can meet the new operational requirements with or without modifications. The Power System converts 13.8 kV prime power to controlled pulsed power required at the NB sources. The major equipment involved are circuit breakers, auto and rectifier transformers, surge suppression components, power tetrodes, HV Decks, and HVDC power transmission to sources. Thermal models were developed for the power transformers to simulate the new operational requirements. Heat runs were conducted for the power tetrodes to verify capability. Other components were analyzed to verify their thermal limitations. This paper describes the details of the evaluation and redesign of the electrical power system components to meet the TPX operational requirements.
{"title":"Neutral beam power system for TPX","authors":"S. Ramakrishnan, O. N. Bowen, T. O'connor, J. Edwards, N. Fromm, R. Hatcher, R. Newman, G. Rossi, T. Stevenson, A. von Halle","doi":"10.1109/FUSION.1993.518473","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518473","url":null,"abstract":"The Tokamak Physics Experiment (TPX) will utilize to the maximum extent the existing Tokamak Fusion Test Reactor (TFTR) equipment and facilities. This is particularly true for the TFTR Neutral Beam (NB) system. Most of the NB hardware, plant facilities, auxiliary sub-systems, power systems, service infrastructure, and control systems can be used as is. The major changes in the NB hardware are driven by the new operating duty cycle. The TFTR Neutral Beam was designed for operation of the sources for 2 seconds every 150 seconds. The TPX requires operation for 1000 seconds every 4500 seconds. During the Conceptual Design Phase of TPX every component of the TFTR NB Electrical Power System was analyzed to verify whether the equipment can meet the new operational requirements with or without modifications. The Power System converts 13.8 kV prime power to controlled pulsed power required at the NB sources. The major equipment involved are circuit breakers, auto and rectifier transformers, surge suppression components, power tetrodes, HV Decks, and HVDC power transmission to sources. Thermal models were developed for the power transformers to simulate the new operational requirements. Heat runs were conducted for the power tetrodes to verify capability. Other components were analyzed to verify their thermal limitations. This paper describes the details of the evaluation and redesign of the electrical power system components to meet the TPX operational requirements.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"117024481","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518339
C.T. Yeaw
The deleterious effects of the D-T fusion radiation environment upon the stability of the cable-in-conduit conductor (CICC) magnet coils have been both qualitatively and quantitatively investigated. Until now, no systematic and accurate analysis of the fluence dependence of the stability of these coils has been performed, and designs have been primarily concerned with the stability of the coils at start-up. The analysis presented here shows that stability as a function of fluence (reactor operating time) degrades much more quickly than previously anticipated. This rapid degradation of coil stability has potentially profound design ramifications. The basis for the present analysis has been a code called MagRad, specifically developed for the purpose of predicting the stability of a fusion magnet coil as a function of fluence, given the coil geometry, flow parameters, and initial materials characteristics. Radiation has significant effects upon some of the basic materials parameters of the coils, such as the stabilizer resistivity and the critical temperature and upper critical field of the superconductor. The code, CICC, developed by R.L. Wong, together with the Dresner formulation for the limiting current, have been incorporated as reliable predictors of the stability of the coil at start-up, which is used as input for MagRad. Most recent data is used with respect to radiation effects upon the materials properties of the coil. Significantly, inappropriate assumptions used in the semi-analytical form which predicts upper critical field as a function of fluence (which has hitherto been widely accepted and used in stability codes) have been corrected in this present study, and a new and much improved empirical form which represents a fit to the data is presented. That the new form is more suitable than the previous one can be clearly seen in that while the previous form gives a peak upper critical field, B/sub c20/, for binary Nb/sub 3/Sn of about 63 T at a fast neutron fluence of about 25/spl times/10/sup 18/ n/cm/sup 2/, the new form mirrors the data which gives a peak B/sub c20/ of about 25 T at a fast neutron fluence of about 4/spl times/10/sup 18/ n/cm/sup 2/ (at zero fluence B/sub c20/ is about 24 T). Additionally, these inappropriate assumptions are discussed in a qualitative manner, and correction is given to the underlying theory. In its primary functional capacity MagRad has been used to analyze the stability of a possible International Thermonuclear Experimental Reactor (ITER) Engineering Design Activity (EDA) coil design, as a function of both fluence and superconducting material.
本文从定性和定量两方面研究了D-T聚变辐射环境对导管内电缆导体(CICC)磁体线圈稳定性的有害影响。到目前为止，还没有对这些线圈稳定性的影响进行系统和准确的分析，并且设计主要关注线圈在启动时的稳定性。本文的分析表明，稳定性作为通量(反应器运行时间)的函数，其退化速度比先前预期的要快得多。这种线圈稳定性的快速退化具有潜在的深远的设计影响。本分析的基础是一个名为MagRad的代码，该代码是专门为预测熔合磁体线圈的稳定性作为通量的函数而开发的，给定线圈的几何形状、流动参数和初始材料特性。辐射对线圈的一些基本材料参数，如稳定器电阻率、超导体临界温度和上临界场等有显著影响。由R.L. Wong开发的代码CICC与Dresner的极限电流公式一起，已被纳入线圈在启动时稳定性的可靠预测指标，这被用作MagRad的输入。最新的数据是关于辐射对线圈材料特性的影响。值得注意的是，在半解析形式中使用的不适当的假设(迄今为止已被广泛接受并在稳定性规范中使用)已在本研究中得到纠正，并提出了一种新的和改进的经验形式，它代表了与数据的拟合。新比前一个更合适的形式中可以清楚地看到,虽然前面的形式给出了峰上临界磁场,B /子甜,二进制Nb /订阅3 / Sn 63 T的快中子积分通量约25 / spl次/ 10 /一口18 / n /厘米2 /一同晚餐,新形式反映的数据给出了B /子甜/峰值约25 T的快中子积分通量/ 10 / 4 / spl倍一口18 / n /厘米/一口2 /(零影响B /子甜/ 24 T)。此外,以定性的方式讨论了这些不适当的假设，并对基础理论进行了修正。在其主要功能能力中，MagRad已被用于分析可能的国际热核实验反应堆(ITER)工程设计活动(EDA)线圈设计的稳定性，作为通量和超导材料的函数。
{"title":"Investigation of the radiation effects upon the stability of fusion magnet coils","authors":"C.T. Yeaw","doi":"10.1109/FUSION.1993.518339","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518339","url":null,"abstract":"The deleterious effects of the D-T fusion radiation environment upon the stability of the cable-in-conduit conductor (CICC) magnet coils have been both qualitatively and quantitatively investigated. Until now, no systematic and accurate analysis of the fluence dependence of the stability of these coils has been performed, and designs have been primarily concerned with the stability of the coils at start-up. The analysis presented here shows that stability as a function of fluence (reactor operating time) degrades much more quickly than previously anticipated. This rapid degradation of coil stability has potentially profound design ramifications. The basis for the present analysis has been a code called MagRad, specifically developed for the purpose of predicting the stability of a fusion magnet coil as a function of fluence, given the coil geometry, flow parameters, and initial materials characteristics. Radiation has significant effects upon some of the basic materials parameters of the coils, such as the stabilizer resistivity and the critical temperature and upper critical field of the superconductor. The code, CICC, developed by R.L. Wong, together with the Dresner formulation for the limiting current, have been incorporated as reliable predictors of the stability of the coil at start-up, which is used as input for MagRad. Most recent data is used with respect to radiation effects upon the materials properties of the coil. Significantly, inappropriate assumptions used in the semi-analytical form which predicts upper critical field as a function of fluence (which has hitherto been widely accepted and used in stability codes) have been corrected in this present study, and a new and much improved empirical form which represents a fit to the data is presented. That the new form is more suitable than the previous one can be clearly seen in that while the previous form gives a peak upper critical field, B/sub c20/, for binary Nb/sub 3/Sn of about 63 T at a fast neutron fluence of about 25/spl times/10/sup 18/ n/cm/sup 2/, the new form mirrors the data which gives a peak B/sub c20/ of about 25 T at a fast neutron fluence of about 4/spl times/10/sup 18/ n/cm/sup 2/ (at zero fluence B/sub c20/ is about 24 T). Additionally, these inappropriate assumptions are discussed in a qualitative manner, and correction is given to the underlying theory. In its primary functional capacity MagRad has been used to analyze the stability of a possible International Thermonuclear Experimental Reactor (ITER) Engineering Design Activity (EDA) coil design, as a function of both fluence and superconducting material.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"93 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"117302041","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518333
T. Burgess, G. R. Walton, T. Meighan, B. Paul
Special tooling has been developed and used to tighten toroidal field (TF) coil casing bolts that have loosened from years of Tokamak Fusion Test Reactor (TFTR) operation. Due to their location, many of the TF casing bolts cannot be directly accessed or viewed. Their condition was first discovered during unrelated inspections in 1988. Engineering solutions were sought until 1992, when a remotely operated wrench concept was successfully demonstrated on a TF coil mockup. The concept was developed into several working tools that have successfully been applied to tighten several thousand TF casing bolts during recent scheduled outages. This effort has improved the integrity and reliability of the TF coil system in preparing for the final experimental phase of the TFTR. This paper discusses the design and application of this tooling.
{"title":"Special remote tooling developed and utilized to tighten TFTR TF coil casing bolts","authors":"T. Burgess, G. R. Walton, T. Meighan, B. Paul","doi":"10.1109/FUSION.1993.518333","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518333","url":null,"abstract":"Special tooling has been developed and used to tighten toroidal field (TF) coil casing bolts that have loosened from years of Tokamak Fusion Test Reactor (TFTR) operation. Due to their location, many of the TF casing bolts cannot be directly accessed or viewed. Their condition was first discovered during unrelated inspections in 1988. Engineering solutions were sought until 1992, when a remotely operated wrench concept was successfully demonstrated on a TF coil mockup. The concept was developed into several working tools that have successfully been applied to tighten several thousand TF casing bolts during recent scheduled outages. This effort has improved the integrity and reliability of the TF coil system in preparing for the final experimental phase of the TFTR. This paper discusses the design and application of this tooling.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"197 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116471277","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518482
S. Sandri, G. Cambi, S. Ciattaglia
The Occupational Radiation Exposure (ORE) evaluation is part of the safety analysis process that has the purpose to permit the optimisation of protection, according to the basic recommendations of ICRP. This paper presents the criteria adopted in order to evaluate the ORE during normal operation and maintenance of NET II/ITER and some results concerning system components located in tokamak and tritium buildings. Prompt radiation, activity concentration and intake situations as well as number of workers, number of events and exposure time are considered. Many systems and components, whose location in the plant can affect radiological protection during maintenance and/or surveillance, are identified together with the operations needed for each activity. Accidental conditions and equipment failures have been considered in the special maintenance activity when they are due to events with high probability of occurrence so that such events might be expected during the life of the plant. Some results are reported showing the ORE figures with reference to the main activities.
{"title":"Preliminary occupational radiation exposure evaluation related to NET II/ITER","authors":"S. Sandri, G. Cambi, S. Ciattaglia","doi":"10.1109/FUSION.1993.518482","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518482","url":null,"abstract":"The Occupational Radiation Exposure (ORE) evaluation is part of the safety analysis process that has the purpose to permit the optimisation of protection, according to the basic recommendations of ICRP. This paper presents the criteria adopted in order to evaluate the ORE during normal operation and maintenance of NET II/ITER and some results concerning system components located in tokamak and tritium buildings. Prompt radiation, activity concentration and intake situations as well as number of workers, number of events and exposure time are considered. Many systems and components, whose location in the plant can affect radiological protection during maintenance and/or surveillance, are identified together with the operations needed for each activity. Accidental conditions and equipment failures have been considered in the special maintenance activity when they are due to events with high probability of occurrence so that such events might be expected during the life of the plant. Some results are reported showing the ORE figures with reference to the main activities.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"26 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124013999","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518291
R. Daugert, G. Lemunyan
A capability for inspection of the hardware inside the Tokamak Fusion Test Reactor (TFTR) vacuum vessel is required for the Deuterium-Tritium (DT) operation phase of the TFTR program. Three remotely controlled illuminators have been installed on the TFTR vacuum vessel to provide the required illumination. The illuminators are controlled from the TFTR control room. Two new thermoelectrically cooled integrating CCD camera systems are installed in the TFTR basement on existing periscopes. Improved performance of the integrating video cameras reduced the number of originally required illuminators from six to three. The two CCD cameras allow integration of video in excess of one minute without CCD dark current becoming significant. These cameras provide a much greater dynamic range than that provided by the standard video cameras which were previously used for inspection.
{"title":"DT configuration of the TFTR illumination system","authors":"R. Daugert, G. Lemunyan","doi":"10.1109/FUSION.1993.518291","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518291","url":null,"abstract":"A capability for inspection of the hardware inside the Tokamak Fusion Test Reactor (TFTR) vacuum vessel is required for the Deuterium-Tritium (DT) operation phase of the TFTR program. Three remotely controlled illuminators have been installed on the TFTR vacuum vessel to provide the required illumination. The illuminators are controlled from the TFTR control room. Two new thermoelectrically cooled integrating CCD camera systems are installed in the TFTR basement on existing periscopes. Improved performance of the integrating video cameras reduced the number of originally required illuminators from six to three. The two CCD cameras allow integration of video in excess of one minute without CCD dark current becoming significant. These cameras provide a much greater dynamic range than that provided by the standard video cameras which were previously used for inspection.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"6 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127113423","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518431
M. Cole, B. Nelson, P. Fogarty, G. Jones, P. Goranson, Y. Gohar, S. Liew
The tokamak radiation shielding includes the neutron and gamma shielding around the torus and penetrations required to (1) limit activation of components outside the shield to levels that permit hands-on maintenance and (2) limit the nuclear heating of the superconducting coils and cold structure. The primary design drivers are space, the 350/spl deg/C bakeout temperature, and cost; therefore, different shield materials were used for different shield components and locations. The shielding is divided into three areas: (1) torus shielding around the vacuum vessel, (2) duct shielding around the vacuum pumping ducts and vertical diagnostic ducts, and (3) penetration shielding in and around the radial ports. The major shield components include water between the walls of the vacuum vessel, lead monosilicate/boron carbide tiles that are attached to the exterior of the vacuum vessel, shield plugs that fill the openings of the large radial ports, and polyethylene/lead/boron shield blocks for duct shielding. A description of the shielding configuration and the performance and operational requirements will be discussed.
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