Design codes for gas cooled reactor components

International Nuclear Information System (INIS)

1990-12-01

High-temperature gas-cooled reactor (HTGR) plants have been under development for about 30 years and experimental and prototype plants have been operated. The main line of development has been electricity generation based on the steam cycle. In addition the potential for high primary coolant temperature has resulted in research and development programmes for advanced applications including the direct cycle gas turbine and process heat applications. In order to compare results of the design techniques of various countries for high temperature reactor components, the IAEA established a Co-ordinated Research Programme (CRP) on Design Codes for Gas-Cooled Reactor Components. The Federal Republic of Germany, Japan, Switzerland and the USSR participated in this Co-ordinated Research Programme. Within the frame of this CRP a benchmark problem was established for the design of the hot steam header of the steam generator of an HTGR for electricity generation. This report presents the results of that effort. The publication also contains 5 reports presented by the participants. A separate abstract was prepared for each of these reports. Refs, figs and tabs

Balance of Plant System Analysis and Component Design of Turbo-Machinery for High Temperature Gas Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

Ballinger, Ronald G. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Wang, Chun Yun [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Kadak, Andrew [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Todreas, Neil [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Mirick, Bradley [Concepts, Northern Engineering and Research, Woburn, MA (United States); Demetri, Eli [Concepts, Northern Engineering and Research, Woburn, MA (United States); Koronowski, Martin [Concepts, Northern Engineering and Research, Woburn, MA (United States)

2004-08-30

power conversion system have been verified with an industry-standard general thermal-fluid code Flownet. With respect to the dynamic model, bypass valve control and inventory control have been used as the primary control methods for the power conversion system. By performing simulation using the dynamic model with the designed control scheme, the combination of bypass and inventory control was optimized to assure system stability within design temperature and pressure limits. Bypass control allows for rapid control system response while inventory control allows for ultimate steady state operation at part power very near the optimum operating point for the system. Load transients simulations show that the indirect, three-shaft arrangement gas turbine power conversion system is stable and controllable. For the indirect cycle the intermediate heat exchanger (IHX) is the interface between the reactor and the turbomachinery systems. As a part of the design effort the IHX was identified as the key component in the system. Two technologies, printed circuit and compact plate-fin, were investigated that have the promise of meeting the design requirements for the system. The reference design incorporates the possibility of using either technology although the compact plate-fin design was chosen for subsequent analysis. The thermal design and parametric analysis with an IHX and recuperator using the plate-fin configuration have been performed. As a three-shaft arrangement, the turbo-shaft sets consist of a pair of turbine/compressor sets (high pressure and low pressure turbines with same-shaft compressor) and a power turbine coupled with a synchronous generator. The turbines and compressors are all axial type and the shaft configuration is horizontal. The core outlet/inlet temperatures are 900/520 C, and the optimum pressure ratio in the power conversion cycle is 2.9. The design achieves a plant net efficiency of approximately 48%.

Balance of Plant System Analysis and Component Design of Turbo-Machinery for High Temperature Gas Reactor Systems

International Nuclear Information System (INIS)

Ballinger, Ronald G.; Chunyun Wang; Kadak, Andrew; Todreas, Neil

2004-01-01

power conversion system have been verified with an industry-standard general thermal-fluid code Flownet. With respect to the dynamic model, bypass valve control and inventory control have been used as the primary control methods for the power conversion system. By performing simulation using the dynamic model with the designed control scheme, the combination of bypass and inventory control was optimized to assure system stability within design temperature and pressure limits. Bypass control allows for rapid control system response while inventory control allows for ultimate steady state operation at part power very near the optimum operating point for the system. Load transients simulations show that the indirect, three-shaft arrangement gas turbine power conversion system is stable and controllable. For the indirect cycle the intermediate heat exchanger (IHX) is the interface between the reactor and the turbomachinery systems. As a part of the design effort the IHX was identified as the key component in the system. Two technologies, printed circuit and compact plate-fin, were investigated that have the promise of meeting the design requirements for the system. The reference design incorporates the possibility of using either technology although the compact plate-fin design was chosen for subsequent analysis. The thermal design and parametric analysis with an IHX and recuperator using the plate-fin configuration have been performed. As a three-shaft arrangement, the turbo-shaft sets consist of a pair of turbine/compressor sets (high pressure and low pressure turbines with same-shaft compressor) and a power turbine coupled with a synchronous generator. The turbines and compressors are all axial type and the shaft configuration is horizontal. The core outlet/inlet temperatures are 900/520 C, and the optimum pressure ratio in the power conversion cycle is 2.9. The design achieves a plant net efficiency of approximately 48%

HTGR nuclear heat source component design and experience

International Nuclear Information System (INIS)

Peinado, C.O.; Wunderlich, R.G.; Simon, W.A.

1982-05-01

The high-temperature gas-cooled reactor (HTGR) nuclear heat source components have been under design and development since the mid-1950's. Two power plants have been designed, constructed, and operated: the Peach Bottom Atomic Power Station and the Fort St. Vrain Nuclear Generating Station. Recently, development has focused on the primary system components for a 2240-MW(t) steam cycle HTGR capable of generating about 900 MW(e) electric power or alternately producing high-grade steam and cogenerating electric power. These components include the steam generators, core auxiliary heat exchangers, primary and auxiliary circulators, reactor internals, and thermal barrier system. A discussion of the design and operating experience of these components is included

Research on Evaluation Methodology for High Temperature Components and Technical Issues

International Nuclear Information System (INIS)

Kim, Y.J.; Han, S.B.

2007-03-01

The research on evaluation methodology for high temperature components and technical issues includes the comparison of evaluation technology of Very High Temperature Reactors(VHTRs) with that of present commercial reactors, the review of Hot Gas Duct(HGD) insulation designs, the analysis of the codes related to VHTR component construction and the analysis of technical issues on application of present codes to HGD construction. Codes to assure the integrity of the VHTR components are not fully prepared yet in any country. To understand the evaluation technology of the VHTR-related codes, key requirements of ASME B and PV Code Section III, Subsection NB and NH were compared. Six kinds of HGD designs were reviewed and compared. A reference which analyzed seven kinds of present component codes were reviewed and the limitations of them were summarized. Especially it was found that the selection of materials is limited, material property data are not enough, and design analysis methodology is not fully specified

Draft of standard for graphite core components in high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Shibata, Taiju; Sawa, Kazuhiro; Eto, Motokuni; Kunimoto, Eiji; Shiozawa, Shusaku; Oku, Tatsuo; Maruyama, Tadashi

2010-01-01

For the design of the graphite components in the High Temperature Engineering Test Reactor (HTTR), the graphite structural design code for the HTTR etc. were applied. However, general standard systems for the High Temperature Gas-cooled Reactor (HTGR) have not been established yet. The authors had studied on the technical issues which is necessary for the establishment of a general standard system for the graphite components in the HTGR. The results of the study were documented and discussed at a 'Special committee on research on preparation for codes for graphite components in HTGR' at Atomic Energy Society of Japan (AESJ). As a result, 'Draft of Standard for Graphite Core Components in High Temperature Gas-cooled Reactor.' was established. In the draft standard, the graphite components are classified three categories (A, B and C) in the standpoints of safety functions and possibility of replacement. For the components in the each class, design standard, material and product standards, and in-service inspection and maintenance standard are determined. As an appendix of the design standard, the graphical expressions of material property data of 1G-110 graphite as a function of fast neutron fluence are expressed. The graphical expressions were determined through the interpolation and extrapolation of the irradiated data. (author)

Design rule for fatigue of welded joints in elevated-temperature nuclear components

International Nuclear Information System (INIS)

O'Connor, D.G.; Corum, J.M.

1986-01-01

Elevated-temperature weldment fatigue failures have occurred in several operating liquid-metal reactor plants. Yet, ASME Code Case N-47, which governs the design of such plants in the United States, does not currently address the Code Subgroup on Elevated Temperature Design recently proposed a fatigue strength reduction factor for austenitic and ferritic steel weldments. The factor is based on a variety of weld metal and weldment fatigue data generated in the United States, Europe, and Japan. This paper describes the factor and its bases, and it presents the results of confirmatory fatigue tests conducted at Oak Ridge National Laboratory on 316 stainless steel tubes with axial and circumferential welds of 16-8-2 filler metal. These test results confirm the suitability of the design factor, and they support the premise that the metallurgical notch effect produced by yield strength variations across a weldment is largely responsible for the observed elevated-temperature fatigue strength reduction

Investigations into High Temperature Components and Packaging

Energy Technology Data Exchange (ETDEWEB)

Marlino, L.D.; Seiber, L.E.; Scudiere, M.B.; M.S. Chinthavali, M.S.; McCluskey, F.P.

2007-12-31

The purpose of this report is to document the work that was performed at the Oak Ridge National Laboratory (ORNL) in support of the development of high temperature power electronics and components with monies remaining from the Semikron High Temperature Inverter Project managed by the National Energy Technology Laboratory (NETL). High temperature electronic components are needed to allow inverters to operate in more extreme operating conditions as required in advanced traction drive applications. The trend to try to eliminate secondary cooling loops and utilize the internal combustion (IC) cooling system, which operates with approximately 105 C water/ethylene glycol coolant at the output of the radiator, is necessary to further reduce vehicle costs and weight. The activity documented in this report includes development and testing of high temperature components, activities in support of high temperature testing, an assessment of several component packaging methods, and how elevated operating temperatures would impact their reliability. This report is organized with testing of new high temperature capacitors in Section 2 and testing of new 150 C junction temperature trench insulated gate bipolar transistor (IGBTs) in Section 3. Section 4 addresses some operational OPAL-GT information, which was necessary for developing module level tests. Section 5 summarizes calibration of equipment needed for the high temperature testing. Section 6 details some additional work that was funded on silicon carbide (SiC) device testing for high temperature use, and Section 7 is the complete text of a report funded from this effort summarizing packaging methods and their reliability issues for use in high temperature power electronics. Components were tested to evaluate the performance characteristics of the component at different operating temperatures. The temperature of the component is determined by the ambient temperature (i.e., temperature surrounding the device) plus the

Design Procedure of Graphite Components by ASME HTR Codes

Energy Technology Data Exchange (ETDEWEB)

Kang, Ji-Ho; Jo, Chang Keun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

In this study, the ASME B and PV Code, Subsection HH, Subpart A, design procedure for graphite components of HTRs was reviewed and the differences from metal materials were remarked. The Korean VHTR has a prismatic core which is made of multiple graphite blocks, reflectors, and core supports. One of the design issues is the assessment of the structural integrity of the graphite components because the graphite is brittle and shows quite different behaviors from metals in high temperature environment. The American Society of Mechanical Engineers (ASME) issued the latest edition of the code for the high temperature reactors (HTR) in 2015. In this study, the ASME B and PV Code, Subsection HH, Subpart A, Graphite Materials was reviewed and the special features were remarked. Due the brittleness of graphites, the damage-tolerant design procedures different from the conventional metals were adopted based on semi-probabilistic approaches. The unique additional classification, SRC, is allotted to the graphite components and the full 3-D FEM or equivalent stress analysis method is required. In specific conditions, the oxidation and viscoelasticity analysis of material are required. The fatigue damage rule has not been established yet.

Design Procedure of Graphite Components by ASME HTR Codes

International Nuclear Information System (INIS)

Kang, Ji-Ho; Jo, Chang Keun

2016-01-01

In this study, the ASME B and PV Code, Subsection HH, Subpart A, design procedure for graphite components of HTRs was reviewed and the differences from metal materials were remarked. The Korean VHTR has a prismatic core which is made of multiple graphite blocks, reflectors, and core supports. One of the design issues is the assessment of the structural integrity of the graphite components because the graphite is brittle and shows quite different behaviors from metals in high temperature environment. The American Society of Mechanical Engineers (ASME) issued the latest edition of the code for the high temperature reactors (HTR) in 2015. In this study, the ASME B and PV Code, Subsection HH, Subpart A, Graphite Materials was reviewed and the special features were remarked. Due the brittleness of graphites, the damage-tolerant design procedures different from the conventional metals were adopted based on semi-probabilistic approaches. The unique additional classification, SRC, is allotted to the graphite components and the full 3-D FEM or equivalent stress analysis method is required. In specific conditions, the oxidation and viscoelasticity analysis of material are required. The fatigue damage rule has not been established yet

Investigation on structural integrity of graphite component during high temperature 950degC continuous operation of HTTR

International Nuclear Information System (INIS)

Sumita, Junya; Shimazaki, Yosuke; Shibata, Taiju

2014-01-01

Graphite material is used for internal structures in high temperature gas-cooled reactor. The core components and graphite core support structures are so designed as to maintain the structural integrity to keep core cooling capability. To confirm that the core components and graphite core support structures satisfy the design requirements, the temperatures of the reactor internals are measured during the reactor operation. Surveillance test of graphite specimens and in-service inspection using TV camera are planned in conjunction with the refueling. This paper describes the evaluation results of the integrity of the core components and graphite core support structures during the high temperature 950degC continuous operation, a high temperature continuous operation with reactor outlet temperature of 950degC for 50 days, in high temperature engineering test reactor. The design requirements of the core components and graphite core support structures were satisfied during the high temperature 950degC continuous operation. The dimensional change of graphite which directly influences the temperature of coolant was estimated considering the temperature profiles of fuel block. The magnitude of irradiation-induced dimensional change considering temperature profiles was about 1.2 times larger than that under constant irradiation temperature of 1000degC. In addition, the programs of surveillance test and ISI using TV camera were introduced. (author)

Study of the degradation of power generation combustion components at elevated temperature

International Nuclear Information System (INIS)

Castrejon, J.; Serna, S.; Wong-Moreno, A.; Fragiel, A.; Lopez-Lopez, D.

2006-01-01

Elevated temperature combustion of fuel oil that contains large amounts of vanadium, asphaltenes and mostly sulfur, presents a major challenge for materials selection and design of combustion components for the electric power generation. The combustion system, which consists of air nozzles and air swirlers, plays a key role in the performance of electric power plants. Air nozzles and air swirlers, which were operated for one year in a 350 MW boiler, were analyzed, presenting accelerated degradation. The particular features of corrosion behavior of these components made by stainless steels: 304, 446 and HH, are presented. The results obtained after optical, metallographic, and microprobe analysis revealed that the components flame contact at very high operating temperature promoted all materials degradation mechanisms. Under this scenario, it is very difficult to find a material resistant to such accelerated wastage conditions. So, the solution of the problem must be oriented to re-design and improve the efficiency of the flame contact with these components

Study of the degradation of power generation combustion components at elevated temperature

Energy Technology Data Exchange (ETDEWEB)

Castrejon, J. [Centro de Investigacion en Ingenieria y Ciencias Aplicadas-UAEM, Av. Universidad 1001, C.P. 62209, Cuernavaca, Mor., Mexico (Mexico); Serna, S. [Centro de Investigacion en Ingenieria y Ciencias Aplicadas-UAEM, Av. Universidad 1001, C.P. 62209, Cuernavaca, Mor., Mexico (Mexico)]. E-mail: aserna@uaem.mx; Wong-Moreno, A. [Instituto Mexicano del Petroleo, Eje Central No. 152, Col. San. Bartolo Atepehuacan, C.P. 07730, Mexico, DF (Mexico); Fragiel, A. [Centro de Ciencias de la Materia Condensada-UNAM, Km 7 Carretera Tijuana-Ensenada, C.P. 22800, Ensenada, Baja California (Mexico); Lopez-Lopez, D. [Instituto Mexicano del Petroleo, Eje Central No. 152, Col. San. Bartolo Atepehuacan, C.P. 07730, Mexico, DF (Mexico)

2006-01-15

Elevated temperature combustion of fuel oil that contains large amounts of vanadium, asphaltenes and mostly sulfur, presents a major challenge for materials selection and design of combustion components for the electric power generation. The combustion system, which consists of air nozzles and air swirlers, plays a key role in the performance of electric power plants. Air nozzles and air swirlers, which were operated for one year in a 350 MW boiler, were analyzed, presenting accelerated degradation. The particular features of corrosion behavior of these components made by stainless steels: 304, 446 and HH, are presented. The results obtained after optical, metallographic, and microprobe analysis revealed that the components flame contact at very high operating temperature promoted all materials degradation mechanisms. Under this scenario, it is very difficult to find a material resistant to such accelerated wastage conditions. So, the solution of the problem must be oriented to re-design and improve the efficiency of the flame contact with these components.

Principle design and data of graphite components

International Nuclear Information System (INIS)

Ishihara, Masahiro; Sumita, Junya; Shibata, Taiju; Iyoku, Tatsuo; Oku, Tatsuo

2004-01-01

The High Temperature Engineering Test Reactor (HTTR) constructed by Japan Atomic Energy Research Institute (JAERI) is a graphite-moderated and helium-gas-cooled reactor with prismatic fuel elements of hexagonal blocks. The reactor internal structures of the HTTR are mainly made up of graphite components. As well known, the graphite is a brittle material and there were no available design criteria for brittle materials. Therefore, JAERI had to develop the design criteria taking account of the brittle fracture behavior. In this paper, concept and key specification of the developed graphite design criteria is described, and also an outline of the quality control specified in the design criteria is mentioned

Construction of a 21-Component Layered Mixture Experiment Design

International Nuclear Information System (INIS)

Piepel, Gregory F.; Cooley, Scott K.; Jones, Bradley

2004-01-01

This paper describes the solution to a unique and challenging mixture experiment design problem involving: (1) 19 and 21 components for two different parts of the design, (2) many single-component and multi-component constraints, (3) augmentation of existing data, (4) a layered design developed in stages, and (5) a no-candidate-point optimal design approach. The problem involved studying the liquidus temperature of spinel crystals as a function of nuclear waste glass composition. The statistical objective was to develop an experimental design by augmenting existing glasses with new nonradioactive and radioactive glasses chosen to cover the designated nonradioactive and radioactive experimental regions. The existing 144 glasses were expressed as 19-component nonradioactive compositions and then augmented with 40 new nonradioactive glasses. These included 8 glasses on the outer layer of the region, 27 glasses on an inner layer, 2 replicate glasses at the centroid, and one replicate each of three existing glasses. Then, the 144 + 40 = 184 glasses were expressed as 21-component radioactive compositions and augmented with 5 radioactive glasses. A D-optimal design algorithm was used to select the new outer layer, inner layer, and radioactive glasses. Several statistical software packages can generate D-optimal experimental designs, but nearly all require a set of candidate points (e.g., vertices) from which to select design points. The large number of components (19 or 21) and many constraints made it impossible to generate the huge number of vertices and other typical candidate points. JMP(R) was used to select design points without candidate points. JMP uses a coordinate-exchange algorithm modified for mixture experiments, which is discussed in the paper

An approach to development of structural design criteria for highly irradiated core components

International Nuclear Information System (INIS)

Nelson, D.V.

1980-01-01

The advent of the fast breeder reactor presents novel challenges in structural design and materials engineering. For instance, the core components of these reactors experience high energy neutron irradiation at elevated temperature, which causes significant time-dependent changes in material behaviour, such as a progressive loss of ductility. New structural design criteria are needed to extend elevated temperature design-by-analysis to account for these changes. Alloys best able to cope with the demands of the core operating environment are being explored and their structural behaviour characterized. The purpose of this paper is to illustrate an approach used in the development of core component structural design criteria. To do this, several design rules, plus brief rationale, from draft RDT Standards F9-7, -8 and -9 will be presented. These recently completed standards ('Structural Design Guidelines for Breeder Reactor Core Components') were prepared for the U.S. Department of Energy and represent a consensus among most organizations participating in the U.S. breeder program. (author)

Structural analysis for elevated temperature design of the LMFBR

International Nuclear Information System (INIS)

Griffin, D.S.

1976-02-01

In the structural design of LMFBR components for elevated temperature service it is necessary to take account of the time-dependent, creep behavior of materials. The accommodation of creep to assure design reliability has required (1) development of new design limits and criteria, (2) development of more detailed representations of material behavior, and (3) application of the most advanced analysis techniques. These developments are summarized and examples are given to illustrate the current state of technology in elevated temperature design

High temperature structural design and R and Ds for heat transport system components of FBR 'Monju'

International Nuclear Information System (INIS)

Sumikawa, Masaharu; Nakagawa, Yukio; Fukuda, Yoshio; Sukegawa, Masayuki; Ishizaki, Tairo.

1980-01-01

The machines and equipments of cooling system for the fast breeder prototype reactor ''Monju'' are operated in creep temperature region, and the upper limit temperature to apply the domestic structural design standard for nuclear machines and equipment is exceeded, therefore the guideline for high temperature structural design is being drawn up, reflecting the results of recent research and development, by the Power Reactor and Nuclear Fuel Development Corp. and others. In order to obtain the basic data for the purpose, the tests on the high temperature characteristics of main structural members and structural elements were carried out, and eight kinds of the inelastic structural analysis program ''HI-EPIC'' series were developed, thus the fundamental technologies of structural desigh in non-linear region were established. Also in the non-linear region, enormous physical quantities must be evaluated, and in the design method based on real elastic analysis, many design diagrams must be employed, therefore for the purpose of improving the reliability of evaluation, the automatic evaluation program ''HI-TEP'' was developed, and preparation has been made for the design of actual machines. The high temperature structural design in ''Monju'', the development of inelastic structural analysis program and high temperature structural analysis evaluation program, and the development of high temperature structures and materials are described. (Kako, I.)

Transient temperature response of in-vessel components due to pulsed operation in tokamak fusion experimental reactor (FER)

International Nuclear Information System (INIS)

Minato, Akio; Tone, Tatsuzo

1985-12-01

A transient temperature response of the in-vessel components (first wall, blanket, divertor/limiter and shielding) surrounding plasma in Tokamak Fusion Experimental Reactor (FER) has been analysed. Transient heat load during start up/shut down and pulsed operation cycles causes the transient temperature response in those components. The fatigue lifetime of those components significantly depends upon the resulting cyclic thermal stress. The burn time affects the temperature control in the solid breeder (Li 2 O) and also affects the thermo-mechanical design of the blanket and shielding which are constructed with thick structure. In this report, results of the transient temperature response obtained by the heat transfer and conduction analyses for various pulsed operation scenarios (start up, shut down, burn and dwell times) have been investigated in view of thermo-mechanical design of the in-vessel components. (author)

Fundamental principles for a nuclear design and structural analysis code for HTR components operating at temperatures above 8000C

International Nuclear Information System (INIS)

Nickel, H.; Schubert, F.

1985-01-01

With reference to the special characteristics of an HTR plant for the supply of nuclear process heat, the investigation of the fundamental principles to form the basis for a high temperature nuclear structural design code has been described. As examples, preliminary design values are proposed for the creep rupture and fatigue behaviour. The linear damage accumulation rule is for practical reasons proposed for the determination of service life, and the difficulties in using this rule are discussed. Finally, using the data obtained in structural analysis, the main areas of investigation which will lead to improvements in the utilization of the materials are discussed. Based on the current information, the working group ''Design Code'' believes that a service life of 70000 h for the heat-exchanging components operating at above 800 0 C can be. (orig.)

Integrated Design Software Predicts the Creep Life of Monolithic Ceramic Components

Science.gov (United States)

1996-01-01

Significant improvements in propulsion and power generation for the next century will require revolutionary advances in high-temperature materials and structural design. Advanced ceramics are candidate materials for these elevated-temperature applications. As design protocols emerge for these material systems, designers must be aware of several innate features, including the degrading ability of ceramics to carry sustained load. Usually, time-dependent failure in ceramics occurs because of two different, delayedfailure mechanisms: slow crack growth and creep rupture. Slow crack growth initiates at a preexisting flaw and continues until a critical crack length is reached, causing catastrophic failure. Creep rupture, on the other hand, occurs because of bulk damage in the material: void nucleation and coalescence that eventually leads to macrocracks which then propagate to failure. Successful application of advanced ceramics depends on proper characterization of material behavior and the use of an appropriate design methodology. The life of a ceramic component can be predicted with the NASA Lewis Research Center's Ceramics Analysis and Reliability Evaluation of Structures (CARES) integrated design programs. CARES/CREEP determines the expected life of a component under creep conditions, and CARES/LIFE predicts the component life due to fast fracture and subcritical crack growth. The previously developed CARES/LIFE program has been used in numerous industrial and Government applications.

Recent UK research and the development of high temperature design methods

International Nuclear Information System (INIS)

Rose, R.T.; Tomkins, B.; Townley, C.H.A.

1987-01-01

The paper outlines recent research and development activities on high temperature design methods and criteria for high temperature components as utilized by liquid metal cooled fast breeder reactors. (orig.)

Research Activities on Development of Piping Design Methodology of High Temperature Reactors

Energy Technology Data Exchange (ETDEWEB)

Huh, Nam-Su [Seoul National Univ. of Science and Technology, Seoul(Korea, Republic of); Won, Min-Gu [Sungkyukwan Univ., Suwon (Korea, Republic of); Oh, Young-Jin [KEPCO Engineering and Construction Co. Inc., Gimcheon (Korea, Republic of); Lee, Hyeog-Yeon; Kim, Yoo-Gon [Korea Atomic Energy Research Institute, Daejeon(Korea, Republic of)

2016-10-15

A SFR is operated at high temperature and low pressure compared with commercial pressurized water reactor (PWR), and such an operating condition leads to time-dependent damages such as creep rupture, excessive creep deformation, creep-fatigue interaction and creep crack growth. Thus, high temperature design and structural integrity assessment methodology should be developed considering such failure mechanisms. In terms of design of mechanical components of SFR, ASME B and PV Code, Sec. III, Div. 5 and RCC-MRx provide high temperature design and assessment procedures for nuclear structural components operated at high temperature, and a Leak-Before-Break (LBB) assessment procedure for high temperature piping is also provided in RCC-MRx, A16. Three web-based evaluation programs based on the current high temperature codes were developed for structural components of high temperature reactors. Moreover, for the detailed LBB analyses of high temperature piping, new engineering methods for predicting creep C*-integral and creep COD rate based either on GE/EPRI or on reference stress concepts were proposed. Finally, the numerical methods based on Garofalo's model and RCC-MRx have been developed, and they have been implemented into ABAQUS. The predictions based on both models were compared with the experimental results, and it has been revealed that the predictions from Garafalo's model gave somewhat successful results to describe the deformation behavior of Gr. 91 at elevated temperatures.

The scope of additive manufacturing in cryogenics, component design, and applications

Science.gov (United States)

Stautner, W.; Vanapalli, S.; Weiss, K.-P.; Chen, R.; Amm, K.; Budesheim, E.; Ricci, J.

2017-12-01

Additive manufacturing techniques using composites or metals are rapidly gaining momentum in cryogenic applications. Small or large, complex structural components are now no longer limited to mere design studies but can now move into the production stream thanks to new machines on the market that allow for light-weight, cost optimized designs with short turnaround times. The potential for cost reductions from bulk materials machined to tight tolerances has become obvious. Furthermore, additive manufacturing opens doors and design space for cryogenic components that to date did not exist or were not possible in the past, using bulk materials along with elaborate and expensive machining processes, e.g. micromachining. The cryogenic engineer now faces the challenge to design toward those new additive manufacturing capabilities. Additionally, re-thinking designs toward cost optimization and fast implementation also requires detailed knowledge of mechanical and thermal properties at cryogenic temperatures. In the following we compile the information available to date and show a possible roadmap for additive manufacturing applications of parts and components typically used in cryogenic engineering designs.

High temperature, high pressure gas loop - the Component Flow Test Loop (CFTL)

International Nuclear Information System (INIS)

Gat, U.; Sanders, J.P.; Young, H.C.

1984-01-01

The high-pressure, high-temperature, gas-circulating Component Flow Test Loop located at Oak Ridge National Laboratory was designed and constructed utilizing Section III of the ASME Boiler and Pressure Vessel Code. The quality assurance program for operating and testing is also based on applicable ASME standards. Power to a total of 5 MW is available to the test section, and an air-cooled heat exchanger rated at 4.4 MW serves as heat sink. The three gas-bearing, completely enclosed gas circulators provide a maximum flow of 0.47 m 3 /s at pressures to 10.7 MPa. The control system allows for fast transients in pressure, power, temperature, and flow; it also supports prolonged unattended steady-state operation. The data acquisition system can access and process 10,000 data points per second. High-temperature gas-cooled reactor components are being tested

Improved temperature regulation of APS linac RF components

International Nuclear Information System (INIS)

Dortwegt, R.

1998-01-01

The temperature of the APS S-Band linac's high-power rf components is regulated by water from individual closed-loop deionized (DI) water systems. The rf components are all made of oxygen-free high-conductivity copper and respond quickly to temperature changes. The SLED cavities are especially temperature-sensitive and cause beam energy instabilities when the temperature is not well regulated. Temperature regulation better than ± 0.1 F is required to achieve good energy stability. Improvements in the closed-loop water systems have enabled them to achieve a regulation of ± 0.05 F over long periods. Regulation philosophy and equipment are discussed and numerical results are presented

Study of elevated temperature design standard against thermal loads

International Nuclear Information System (INIS)

Kasahara, Naoto; Asayama, Tai; Morishita, Masaki

2001-01-01

Elevated temperature components must be designed against both pressure and thermal loads. In the case of sodium circuits of fast breeder reactors, a restriction from the pressure load becomes small because of the high boiling point of sodium. Design approaches for thermal loads (displacement-controlled) are compared with those against pressure loads (load-controlled). Considering differences between those two approaches, a concept of the elevated temperature design standard that takes the nature of thermal loads fully into account is proposed. This concept is a basis of load evaluation techniques and an inelastic analysis guide, that are being developed. Finally, problems and plans to realize the above concept are discussed. (author)

Concepts on high temperature design analysis for SNR 300

International Nuclear Information System (INIS)

Bieniussa, K.; Zolti, E.

1976-01-01

The paper briefly describes the evolution, the present situation and the next activities on the design of high temperature components of the DEBENELUX prototype fast breeder reactor SNR-300 with particular regard to the design criteria. Elastic structural analyses are performed for the basic design of the components and are supplied by the manufacturer. In agreement with the Safety Experts simplified and/or detailed inelastic analyses of the critical areas are supplied by the prime contractor of the plant. The elastic computations are evaluated on the basis of a set of design rules derived from ASME Code Case Interpretation 1331-4 but with more conservative limits, and the inelastic ones on the basis of the ASME Code Case Interpretation 1592

Mechanical design of machine components

CERN Document Server

Ugural, Ansel C

2015-01-01

Mechanical Design of Machine Components, Second Edition strikes a balance between theory and application, and prepares students for more advanced study or professional practice. It outlines the basic concepts in the design and analysis of machine elements using traditional methods, based on the principles of mechanics of materials. The text combines the theory needed to gain insight into mechanics with numerical methods in design. It presents real-world engineering applications, and reveals the link between basic mechanics and the specific design of machine components and machines. Divided into three parts, this revised text presents basic background topics, deals with failure prevention in a variety of machine elements and covers applications in design of machine components as well as entire machines. Optional sections treating special and advanced topics are also included.Key Features of the Second Edition:Incorporates material that has been completely updated with new chapters, problems, practical examples...

Component design challenges for the ground-based SP-100 nuclear assembly test

International Nuclear Information System (INIS)

Markley, R.A.; Disney, R.K.; Brown, G.B.

1989-01-01

The SP-100 ground engineering system (GES) program involves a ground test of the nuclear subsystems to demonstrate their design. The GES nuclear assembly test (NAT) will be performed in a simulated space environment within a vessel maintained at ultrahigh vacuum. The NAT employs a radiation shielding system that is comprised of both prototypical and nonprototypical shield subsystems to attenuate the reactor radiation leakage and also nonprototypical heat transport subsystems to remove the heat generated by the reactor. The reactor is cooled by liquid lithium, which will operate at temperatures prototypical of the flight system. In designing the components for these systems, a number of design challenges were encountered in meeting the operational requirements of the simulated space environment (and where necessary, prototypical requirements) while also accommodating the restrictions of a ground-based test facility with its limited available space. This paper presents a discussion of the design challenges associated with the radiation shield subsystem components and key components of the heat transport systems

Flow boiling heat transfer coefficients at cryogenic temperatures for multi-component refrigerant mixtures of nitrogen-hydrocarbons

Science.gov (United States)

Ardhapurkar, P. M.; Sridharan, Arunkumar; Atrey, M. D.

2014-01-01

The recuperative heat exchanger governs the overall performance of the mixed refrigerant Joule-Thomson cryocooler. In these heat exchangers, the non-azeotropic refrigerant mixture of nitrogen-hydrocarbons undergoes boiling and condensation simultaneously at cryogenic temperature. Hence, the design of such heat exchanger is crucial. However, due to lack of empirical correlations to predict two-phase heat transfer coefficients of multi-component mixtures at low temperature, the design of such heat exchanger is difficult.

Preconceptual design of hyfire. A fusion driven high temperature electrolysis plant

International Nuclear Information System (INIS)

Varljen, T.C.; Chi, J.W.H.; Karbowski, J.S.

1983-01-01

Brookhaven National Laboratory has been engaged in a scoping study to investigate the potential merits of coupling a fusion reactor with a high temperature blanket to a high temperature electrolysis (HTE) process to produce hydrogen and oxygen. Westinghouse is assisting this study in the areas of systems design integration, plasma engineering, balance of plant design and electrolyzer technology. The aim of the work done in the past year has been to focus on a reference design point for the plant, which has been designated HYFIRE. In prior work, the STARFIRE commercial tokamak fusion reactor was directly used as the fusion driver. This report describes a new design obtained by scaling the basic STARFIRE design to permit the achievement of a blanket power of 6000 MWt. The high temperature blanket design employs a thermally insulated refractory oxide region which provides high temperature (>1000 deg. C) steam at moderate pressures to high temperature electrolysis units. The electrolysis process selected is based on the high temperature, solid electrolyte fuel cell technology developed by Westinghouse. An initial process design and plant layout has been completed; component cost and plant economics studies are now underway to develop estimates of hydrogen production costs and to determine the sensitivity of this cost to changes in major design parameters. (author)

Engineering design of a high-temperature superconductor current lead

International Nuclear Information System (INIS)

Niemann, R.C.; Cha, Y.S.; Hull, J.R.; Daugherty, M.A.; Buckles, W.E.

1993-01-01

As part of the US Department of Energy's Superconductivity Pilot Center Program, Argonne National Laboratory and Superconductivity, Inc., are developing high-temperature superconductor (HTS) current leads suitable for application to superconducting magnetic energy storage systems. The principal objective of the development program is to design, construct, and evaluate the performance of HTS current leads suitable for near-term applications. Supporting objectives are to (1) develop performance criteria; (2) develop a detailed design; (3) analyze performance; (4) gain manufacturing experience in the areas of materials and components procurement, fabrication and assembly, quality assurance, and cost; (5) measure performance of critical components and the overall assembly; (6) identify design uncertainties and develop a program for their study; and (7) develop application-acceptance criteria

Engineering design of a high-temperature superconductor current lead

Science.gov (United States)

Niemann, R. C.; Cha, Y. S.; Hull, J. R.; Daugherty, M. A.; Buckles, W. E.

As part of the US Department of Energy's Superconductivity Pilot Center Program, Argonne National Laboratory and Superconductivity, Inc., are developing high-temperature superconductor (HTS) current leads suitable for application to superconducting magnetic energy storage systems. The principal objective of the development program is to design, construct, and evaluate the performance of HTS current leads suitable for near-term applications. Supporting objectives are to (1) develop performance criteria; (2) develop a detailed design; (3) analyze performance; (4) gain manufacturing experience in the areas of materials and components procurement, fabrication and assembly, quality assurance, and cost; (5) measure performance of critical components and the overall assembly; (6) identify design uncertainties and develop a program for their study; and (7) develop application-acceptance criteria.

Radiation and temperature effects on electronic components investigated under the CSTI high capacity power project

International Nuclear Information System (INIS)

Schwarze, G.E.; Niedra, J.M.; Frasca, A.J.; Wieserman, W.R.

1993-01-01

The effects of nuclear radiation and high temperature environments must be fully known and understood for the electronic components and materials used in both the Power Conditioning and Control subsystem and the reactor Instrumentation and Control subsystem of future high capacity nuclear space power systems. This knowledge is required by the designer of these subsystems in order to develop highly reliable, long-life power systems for future NASA missions. A review and summary of the experimental results obtained for the electronic components and materials investigated under the power management element of the CSTI high capacity power project will be presented in this paper: (1) Neutron, gamma ray, and temperature effects on power semiconductor switches, (2) Temperature and frequency effects on soft magnetic materials; and (3) Temperature effects on rare-earth permanent magnets

Radiation and temperature effects on electronic components investigated under the CSTI High Capacity Power Project

International Nuclear Information System (INIS)

Shwarze, G.E.; Wieserman, W.R.

1994-01-01

The effects of nuclear radiation and high temperature environments must be fully known and understood for the electronic components and materials used in both the Power Conditioning and Control subsystem and the reactor Instrumentation and Control subsystem of future high capacity nuclear space power systems. This knowledge is required by the designer of these subsystems in order to develop highly reliable, long-life power systems for future NASA missions. A review and summary of the experimental results obtained for the electronic components and materials investigated under the power management element of the CSTI high capacity power project will be presented in this paper: (1) Neutron, gamma ray, and temperature effects on power semiconductor switches, (2) Temperature and frequency effects on soft magnetic materials; and (3) Temperature effects on rare earth permanent magnets

High temperature pipeline design

Energy Technology Data Exchange (ETDEWEB)

Greenslade, J.G. [Colt Engineering, Calgary, AB (Canada). Pipelines Dept.; Nixon, J.F. [Nixon Geotech Ltd., Calgary, AB (Canada); Dyck, D.W. [Stress Tech Engineering Inc., Calgary, AB (Canada)

2004-07-01

It is impractical to transport bitumen and heavy oil by pipelines at ambient temperature unless diluents are added to reduce the viscosity. A diluted bitumen pipeline is commonly referred to as a dilbit pipeline. The diluent routinely used is natural gas condensate. Since natural gas condensate is limited in supply, it must be recovered and reused at high cost. This paper presented an alternative to the use of diluent to reduce the viscosity of heavy oil or bitumen. The following two basic design issues for a hot bitumen (hotbit) pipeline were presented: (1) modelling the restart problem, and, (2) establishing the maximum practical operating temperature. The transient behaviour during restart of a high temperature pipeline carrying viscous fluids was modelled using the concept of flow capacity. Although the design conditions were hypothetical, they could be encountered in the Athabasca oilsands. It was shown that environmental disturbances occur when the fluid is cooled during shut down because the ground temperature near the pipeline rises. This can change growing conditions, even near deeply buried insulated pipelines. Axial thermal loads also constrain the design and operation of a buried pipeline as higher operating temperatures are considered. As such, strain based design provides the opportunity to design for higher operating temperature than allowable stress based design methods. Expansion loops can partially relieve the thermal stress at a given temperature. As the design temperature increase, there is a point at which above grade pipelines become attractive options, although the materials and welding procedures must be suitable for low temperature service. 3 refs., 1 tab., 10 figs.

High-Temperature Air-Cooled Power Electronics Thermal Design: Annual Progress Report

Energy Technology Data Exchange (ETDEWEB)

Waye, Scot [National Renewable Energy Lab. (NREL), Golden, CO (United States)

2016-08-01

Power electronics that use high-temperature devices pose a challenge for thermal management. With the devices running at higher temperatures and having a smaller footprint, the heat fluxes increase from previous power electronic designs. This project overview presents an approach to examine and design thermal management strategies through cooling technologies to keep devices within temperature limits, dissipate the heat generated by the devices and protect electrical interconnects and other components for inverter, converter, and charger applications. This analysis, validation, and demonstration intends to take a multi-scale approach over the device, module, and system levels to reduce size, weight, and cost.

Degradation of Solar Array Components in a Combined UV/VUV High Temperature Test Environment

Directory of Open Access Journals (Sweden)

Nömayr Christel

2017-01-01

A design verification test under UV/VUV conditions of sun exposed materials and technologies on component level is presented which forms part of the overall verification and qualification of the solar array design of the MTM and MPO. The test concentrates on the self-contamination aspects and the resulting performance losses of the solar array under high intensity and elevated temperature environment representative for the photovoltaic assembly (PVA.

Mechanical design philosophy for the graphite components of the core structure of an HTGR

International Nuclear Information System (INIS)

Bodmann, E.

1987-01-01

Parallel to the layout and design of the graphite components for THTRs and the succeeding high temperature reactor projects, the design methods for graphite components have been improved over the years. The aim of this works is to develop the design methods which take into account both the particular properties of graphite and the particular functions of the components. Because of the close relation ship between materials and design codes, this development work has progressed with the development, testing and qualification of German reactor graphite. In this paper, the experience in this field of Hochtemperatur Reaktorbau GmbH and the results of the work and approach to the design problems are reported. The example of a HTR 500 design for a 550 MWe power station is taken up, and the core structure is explained. The graphite components are divided into three classes according to the stress limits. The loading of these components is reviewed. The aim of the design is not the complete avoidance of failure, but to avoid the failure of a single component from leading to a disadvantageous consequence which is not allowable. The classification of loading events, Weibull statistics and maximum allowable stress, the formation of the permissible stress, the assessment of stress due to multiaxial loading and so on are described. (Kako, I.)

RCC-MRx: Design and construction rules for mechanical components in high-temperature structures, experimental reactors and fusion reactors

International Nuclear Information System (INIS)

2015-01-01

The RCC-MRx code was developed for sodium-cooled fast reactors (SFR), research reactors (RR) and fusion reactors (FR-ITER). It provides the rules for designing and building mechanical components involved in areas subject to significant creep and/or significant irradiation. In particular, it incorporates an extensive range of materials (aluminum and zirconium alloys in response to the need for transparency to neutrons), sizing rules for thin shells and box structures, and new modern welding processes: electron beam, laser beam, diffusion and brazing. The RCC-MR code was used to design and build the prototype Fast Breeder Reactor (PFBR) developed by IGCAR in India and the ITER Vacuum Vessel. The RCC-Mx code is being used in the current construction of the RJH experimental reactor (Jules Horowitz reactor). The RCC-MRx code is serving as a reference for the design of the ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration), for the design of the primary circuit in MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) and the design of the target station of the ESS project (European Spallation Source). Contents of the 2015 edition of the RCC-MRx code: Section I General provisions; Section II Additional requirements and special provisions; Section III Rules for nuclear installation mechanical components: Volume I: Design and construction rules: Volume A (RA): General provisions and entrance keys, Volume B (RB): Class 1 components and supports, Volume C (RC): Class 2 components and supports, Volume D (RD): Class 3 components and supports, Volume K (RK): Examination, handling or drive mechanisms, Volume L (RL): Irradiation devices, Volume Z (Ai): Technical appendices; Volume II: Materials; Volume III: Examinations methods; Volume IV: Welding; Volume V: Manufacturing operations; Volume VI: Probationary phase rules

Requirements on the mechanical design of reactor systems operating at elevated temperature

International Nuclear Information System (INIS)

Schulz, H.; Glahn, M.

1979-01-01

The paper presents the contemporary status of the requirements on the mechanical design and analysis developed during the licensing procedure of reactor systems operating at elevated temperature. General requirements for the design at elevated temperature are reviewed. The main proposal is to point out some limit strain criteria which are not included in present design guidelines and codes. The developed strain criteria are used to limit the component deformations in case of power excursions like the Bethe-Tait accident. It is also applicable for loads arising from other faulted conditions. (orig.)

Analysis and evaluation system for elevated temperature design of pressure vessels

International Nuclear Information System (INIS)

Hayakawa, Teiji; Sayawaki, Masaaki; Nishitani, Masahiro; Mii, Tatsuo; Murasawa, Kanji

1977-01-01

In pressure vessel technology, intensive efforts have recently been made to develop the elevated temperature design methods. Much of the impetus of these efforts has been provided mainly by the results of the Liquid Metal Fast Breeder Reactor (LMFBR) and more recently, of the High Temperature Gas-cooled Reactor (HTGR) Programs. The pressure vessels and associated components in these new type nuclear power plants must operate for long periods at elevated temperature where creep effects are significant and then must be designed by rigorous analysis for high reliability and safety. To carry out such an elevated temperature designing, numbers of highly developed analysis and evaluation techniques, which are so complicated as to be impossible by manual work, are indispensable. Under these circumstances, the authors have made the following approaches in the study: (1) Study into basic concepts and the associated techniques in elevated temperature design. (2) Systematization (Analysis System) of the procedure for loads and stress analyses. (3) Development of post-processor, ''POST-1592'', for strength evaluation based on ASME Code Case 1592-7. By linking the POST-1592 together with the Analysis System, an analysis and evaluation system is developed for an elevated temperature design of pressure vessels. Consequently, designing of elevated temperature vessels by detailed analysis and evaluation has easily and effectively become feasible by applying this software system. (auth.)

Components of the LWR primary circuit. Pt. 2. Design, construction and calculation. Draft

International Nuclear Information System (INIS)

1995-01-01

This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673 K (400 deg C). The primary circuit as the pressure containment of the reactor coolant comprises: Reactor pressure vessel (without internals), steam generator (primary loop), pressurizer, reactor coolant pump housing, interconnecting pipings between the components mentioned above and appropriate various valve and instrument casings, pipings branding from the above components and interconnecting pipings, including the appropriate instrument casings, up to and including the first isolating valve, pressure shielding of control rod drives. (orig.) [de

Current Status of the Elevated Temperature Structure Design Codes for VHTR

International Nuclear Information System (INIS)

Kim, Jong-Bum; Kim, Seok-Hoon; Park, Keun-Bae; Lee, Won-Jae

2006-01-01

An elevated temperature structure design and analysis is one of the key issues in the VHTR (Very High Temperature Reactor) project to achieve an economic production of hydrogen which will be an essential energy source for the near future. Since the operating temperature of a VHTR is above 850 .deg. C, the existing code and standards are insufficient for a high temperature structure design. Thus the issues concerning a material selection and behaviors are being studied for the main structural components of a VHTR in leading countries such as US, France, UK, and Japan. In this study, the current status of the ASME code, French RCC-MR, UK R5, and Japanese code were investigated and the necessary R and D items were discussed

Surface temperature measurement of plasma facing components in tokamaks

International Nuclear Information System (INIS)

Amiel, Stephane

2014-01-01

During this PhD, the challenges on the non-intrusive surface temperature measurements of metallic plasma facing components in tokamaks are reported. Indeed, a precise material emissivity value is needed for classical infrared methods and the environment contribution has to be known particularly for low emissivities materials. Although methods have been developed to overcome these issues, they have been implemented solely for dedicated experiments. In any case, none of these methods are suitable for surface temperature measurement in tokamaks.The active pyrometry introduced in this study allows surface temperature measurements independently of reflected flux and emissivities using pulsed and modulated photothermal effect. This method has been validated in laboratory on metallic materials with reflected fluxes for pulsed and modulated modes. This experimental validation is coupled with a surface temperature variation induced by photothermal effect and temporal signal evolvement modelling in order to optimize both the heating source characteristics and the data acquisition and treatment. The experimental results have been used to determine the application range in temperature and detection wavelengths. In this context, the design of an active pyrometry system on tokamak has been completed, based on a bicolor camera for a thermography application in metallic (or low emissivity) environment.The active pyrometry method introduced in this study is a complementary technique of classical infrared methods used for thermography in tokamak environment which allows performing local and 2D surface temperature measurements independently of reflected fluxes and emissivities. (author) [fr

Designing for elevated temperature

International Nuclear Information System (INIS)

Boer, G.A. de

1982-01-01

The reasons for the application of higher process temperatures are explained. The properties of stainless steel are compared with those of other materials such as molybdenum. Factors influencing the choice of the material such as availability of material data at high temperature, controllability, and strength of heat-affected zone are discussed. The process of designing a structure for safe and economic high-temperature application is outlined: design-by-analysis in contrast to the design-by-rule which is general practice for low-temperature applications. The rules laid down in the ASME Pressure Vessel Code Case N47 are explained as well as the procedure for inelastic stress calculations. (author)

A simplified approach for evaluating secondary stresses in elevated temperature design

International Nuclear Information System (INIS)

Becht, C.

1983-01-01

Control of secondary stresses is important for long-term reliability of components, particularly at elevated temperatures where substantial creep damage can occur and result in cracking. When secondary stresses are considered in the design of elevated temperature components, these are often addressed by the criteria contained in Nuclear Code Case N-47 for use with elastic or inelastic analysis. The elastic rules are very conservative as they bound a large range of complex phenomena; because of this conservatism, only components in relatively mild services can be designed in accordance with these rules. The inelastic rules, although more accurate, require complex and costly nonlinear analysis. Elevated temperature shakedown is a recognized phenomenon that has been considered in developing Code rules and simplified methods. This paper develops and examines the implications of using a criteria which specifically limits stresses to the shakedown regime. Creep, fatigue, and strain accumulation are considered. The effect of elastic follow-up on the conservatism of the criteria is quantified by means of a simplified method. The level of conservatism is found to fall between the elastic and inelastic rules of N-47 and, in fact, the incentives for performing complex inelastic analyses appear to be low except in the low cycle regime. The criteria has immediate applicability to non-code components such as vessel internals in the chemical, petroleum, and synfuels industry. It is suggested that such a criteria be considered in future code rule development

Problems of the Starting and Operating of Hydraulic Components and Systems in Low Ambient Temperature (Part IV

Directory of Open Access Journals (Sweden)

Jasiński Ryszard

2017-09-01

Full Text Available Designers of hydraulically driven machines and devices are obliged to ensure during design process their high service life with taking into account their operational conditions. Some of the machines may be started in low ambient temperature and even in thermal shock conditions (due to delivering hot working medium to cold components. In order to put such devices into operation appropriate investigations, including experimental ones - usually very expensive and time-consuming, are carried out. For this reason numerical calculations can be used to determine serviceability of a hydraulic component or system operating in thermal shock conditions. Application of numerical calculation methods is much less expensive in comparison to experimental ones. This paper presents a numerical calculation method which makes it possible to solve issues of heat exchange in elements of investigated hydraulic components by using finite elements method. For performing the simulations the following data are necessary: ambient temperature, oil temperature, heat transfer coefficient between oil and surfaces of elements, as well as areas of surfaces being in contact with oil. By means of computer simulation method values of clearance between cooperating elements as well as ranges of parameters of correct and incorrect operation of hydraulic components have been determined. In this paper results of computer simulation of some experimentally tested hydraulic components such as axial piston pump and proportional spool valve, are presented. The computer simulation results were compared with the experimental ones and high conformity was obtained.

Comparison of elevated temperature design codes of ASME Subsection NH and RCC-MRx

Energy Technology Data Exchange (ETDEWEB)

Lee, Hyeong-Yeon, E-mail: hylee@kaeri.re.kr

2016-11-15

Highlights: • Comparison of elevated temperature design (ETD) codes was made. • Material properties and evaluation procedures were compared. • Two heat-resistant materials of Grade 91 steel and austenitic stainless steel 316 are the target materials in the present study. • Application of the ETD codes to Generation IV reactor components and a comparison of the conservatism was conducted. - Abstract: The elevated temperature design (ETD) codes are used for the design evaluation of Generation IV (Gen IV) reactor systems such as sodium-cooled fast reactor (SFR), lead-cooled fast reactor (LFR), and very high temperature reactor (VHTR). In the present study, ETD code comparisons were made in terms of the material properties and design evaluation procedures for the recent versions of the two major ETD codes, ASME Section III Subsection NH and RCC-MRx. Conservatism in the design evaluation procedures was quantified and compared based on the evaluation results for SFR components as per the two ETD codes. The target materials are austenitic stainless steel 316 and Mod.9Cr-1Mo steel, which are the major two materials in a Gen IV SFR. The differences in the design evaluation procedures as well as the material properties in the two ETD codes are highlighted.

Development of guidelines for inelastic analysis in design of fast reactor components

International Nuclear Information System (INIS)

Nakamura, Kyotada; Kasahara, Naoto; Morishita, Masaki; Shibamoto, Hiroshi; Inoue, Kazuhiko; Nakayama, Yasunari

2008-01-01

The interim guidelines for the application of inelastic analysis to design of fast reactor components were developed. These guidelines are referred from 'Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)'. The basic policies of the guidelines are more rational predictions compared with elastic analysis approach and a guarantee of conservative results for design conditions. The guidelines recommend two kinds of constitutive equations to estimate strains conservatively. They also provide the methods for modeling load histories and estimating fatigue and creep damage based on the results of inelastic analysis. The guidelines were applied to typical design examples and their results were summarized as exemplars to support users

Discussion on Design Transients of Pebble-bed High Temperature Gas-cooled Reactor

International Nuclear Information System (INIS)

Wang Yan; Li Fu; Zheng Yanhua

2014-01-01

In order to assure high quality for the components and their supports in the reactor coolant system, etc., some thermal-hydraulic transient conditions will be selected and researched for equipment design evaluation to satisfy the requirements ASME code, which are based on the conservative estimates of the magnitude and frequency of the temperature and pressure transients resulting from various operating conditions in the plant. In the mature design on pressurized water reactor, five conditions are considered. For the developing advanced pebble-bed high temperature gas-cooled reactor(HTGR), its design and operation has much difference with other reactors, so the transients of the pebble-bed high temperature gas-cooled reactor have distinctive characteristics. In this paper, the possible design transients of the pebble-bed HTGR will be discussed, and the frequency of design transients for equipment fatigue analysis and stress analysis due to cyclic stresses is also studied. The results will provide support for the design and construct of the pebble-bed HTGR. (author)

Summary of the experimental multi-purpose very high temperature gas cooled reactor design

International Nuclear Information System (INIS)

1984-12-01

The report presents the design of Multi-purpose Very High Temperature Gas Cooled Reactor (the Experimental VHTR) based on the second stage of detailed design which was completed on March 1984, in the from of ''An application of reactor construction permit Appendix 8''. The Experimental VHTR is designed to satisfy with the design specification for the reactor thermal output 50 MW and reactor outlet temperature 950 0 C. The adequacy of the design is also checked by the safety analysis. The planning of plant system and safety is summarized such as safety design requirements and conformance with them, seismic design and plant arrangement. Concerning with the system of the Experimental VHTR the design basis, design data and components are described in the order. (author)

Nonlinear structural analysis methods and their application to elevated temperature design: A US perspective

International Nuclear Information System (INIS)

Dhalla, A.K.

1989-01-01

Technological advances over the last two decades have been assimilated into the routine design of Liquid Metal Reactor (LMR) structural components operating at elevated temperatures. The mature elevated temperature design technology is based upon: (a) an extensive material data base, (b) recent advances in nonlinear computational methods, and (c) conservative design criteria based upon past successful and reliable operating experiences with petrochemical and nonnuclear power plants. This survey paper provides a US perspective on the role of nonlinear analysis methods used in the design of LMR plants. The simplified and detailed nonlinear analysis methods and the level of computational effort required to qualify structural components for safe and reliable long-term operation are discussed. The paper also illustrates how a detailed nonlinear analysis can be used to resolve technical licensing issues, to understand complex nonlinear structural behavior, to identify predominant failure modes, and to guide future experimental programs

Design of high temperature Engineering Test Reactor (HTTR)

International Nuclear Information System (INIS)

Saito, Shinzo; Tanaka, Toshiyuki; Sudo, Yukio

1994-09-01

Construction of High Temperature Engineering Test Reactor (HTTR) is now underway to establish and upgrade basic technologies for HTGRs and to conduct innovative basic research at high temperatures. The HTTR is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 850degC for rated operation and 950degC for high temperature test operation. It is planned to conduct various irradiation tests for fuels and materials, safety demonstration tests and nuclear heat application tests. JAERI received construction permit of HTTR reactor facility in February 1990 after 22 months of safety review. This report summarizes evaluation of nuclear and thermal-hydraulic characteristics, design outline of major systems and components, and also includes relating R and D result and safety evaluation. Criteria for judgment, selection of postulated events, major analytical conditions for anticipated operational occurrences and accidents, computer codes used in safety analysis and evaluation of each event are presented in the safety evaluation. (author)

Apparatus and method for temperature mapping a turbine component in a high temperature combustion environment

Science.gov (United States)

Baleine, Erwan; Sheldon, Danny M

2014-06-10

Method and system for calibrating a thermal radiance map of a turbine component in a combustion environment. At least one spot (18) of material is disposed on a surface of the component. An infrared (IR) imager (14) is arranged so that the spot is within a field of view of the imager to acquire imaging data of the spot. A processor (30) is configured to process the imaging data to generate a sequence of images as a temperature of the combustion environment is increased. A monitor (42, 44) may be coupled to the processor to monitor the sequence of images of to determine an occurrence of a physical change of the spot as the temperature is increased. A calibration module (46) may be configured to assign a first temperature value to the surface of the turbine component when the occurrence of the physical change of the spot is determined.

Design issues and implications for the structural integrity and lifetime of fusion power plant components

International Nuclear Information System (INIS)

Karditas, P.J.

1996-05-01

This review discusses, with example calculations, the criteria, and imposed constraints and limitations, for the design of fusion components and assesses the implications for successful design and power plant operation. The various loading conditions encountered during the operation of a tokamak lead to structural damage and possible failure by such mechanisms as yielding, thermal creep rupture and fatigue due to thermal cycling, plastic strain cycling (ratcheting), crack growth-propagation and radiation induced swelling and creep. Of all the possible damage mechanisms, fatigue, creep and their combination are the most important in the structural design and lifetime of fusion power plant components operating under steady or load varying conditions. Also, the effect of neutron damage inflicted onto the structural materials and the degradation of key properties is of major concern in the design and lifetime prediction of components. Structures are classified by, and will be restricted by existing or future design codes relevant to medium and high temperature power plant environments. The ways in which existing design codes might be used in present and near future design activities, and the implications, are discussed; the desirability of an early start towards the development of fusion-specific design codes is emphasised. (UK)

Composition-Based Prediction of Temperature-Dependent Thermophysical Food Properties: Reevaluating Component Groups and Prediction Models.

Science.gov (United States)

Phinney, David Martin; Frelka, John C; Heldman, Dennis Ray

2017-01-01

Prediction of temperature-dependent thermophysical properties (thermal conductivity, density, specific heat, and thermal diffusivity) is an important component of process design for food manufacturing. Current models for prediction of thermophysical properties of foods are based on the composition, specifically, fat, carbohydrate, protein, fiber, water, and ash contents, all of which change with temperature. The objectives of this investigation were to reevaluate and improve the prediction expressions for thermophysical properties. Previously published data were analyzed over the temperature range from 10 to 150 °C. These data were analyzed to create a series of relationships between the thermophysical properties and temperature for each food component, as well as to identify the dependence of the thermophysical properties on more specific structural properties of the fats, carbohydrates, and proteins. Results from this investigation revealed that the relationships between the thermophysical properties of the major constituents of foods and temperature can be statistically described by linear expressions, in contrast to the current polynomial models. Links between variability in thermophysical properties and structural properties were observed. Relationships for several thermophysical properties based on more specific constituents have been identified. Distinctions between simple sugars (fructose, glucose, and lactose) and complex carbohydrates (starch, pectin, and cellulose) have been proposed. The relationships between the thermophysical properties and proteins revealed a potential correlation with the molecular weight of the protein. The significance of relating variability in constituent thermophysical properties with structural properties--such as molecular mass--could significantly improve composition-based prediction models and, consequently, the effectiveness of process design. © 2016 Institute of Food Technologists®.

Effect of Temperature and Hose Genotype on Components of ...

African Journals Online (AJOL)

Effect of Temperature and Hose Genotype on Components of Resistance to Groundnut Rust. P Subrahmanyam, PV Subba Rao, PM Reddy, D McDonald. Abstract. The effects of temperature on incubation period, infection frequency, lesion diameter, leaf area damage, pustule rupture, and sporulation were quantified for six ...

HTR-E project. High-temperature components and systems

International Nuclear Information System (INIS)

Breuil, E.; Exner, R.

2002-01-01

The HTR-E European project (four years project) is proposed for the 5th Framework Programme and concerns the technical developments needed for the innovative components of a modern HTR with a direct cycle. These components have been selected with reference to the present projects (GT-MHR, PBMR): (1) the helium turbine, the recuperator heat exchanger, the electro-magnetic bearings and the helium rotating seal; (2) the tribology. Sliding innovative components in helium environment are particularly concerned. (3) the helium purification system. Recommendations on impurities contents have to be provided in accordance with the materials proposed for the innovative components. The main outcomes expected from the HTR-E project are the design recommendations and identification of further R and D needs for these components. This will be based: (1) on experience feedback from European past helium test loops and reactors; (2) on design studies, thermal-hydraulic and structural analyses; (3) and on experimental tests

Thermohydraulic design of saturated temperature capsule for IASCC irradiation test

International Nuclear Information System (INIS)

Ide, Hiroshi; Matsui, Yoshinori; Itabashi, Yukio

2002-10-01

An advanced water chemistry controlled irradiation research device is being developed in JAERI, to perform irradiation tests for irradiation assisted stress corrosion cracking (IASCC) research concerned with aging of LWR. This device enables the irradiation tests under the water chemistry condition and the temperature, which simulate the conditions for BWR core internals. The advanced water chemistry controlled irradiation research device is composed of saturated temperature capsule inserted into the JMTR core and the water chemistry control unit installed in the reactor building. Regarding the saturated temperature capsule, the Thermohydraulic design of capsule structure was done, aimed at controlling the specimen's temperature, feeding water velocity on specimen's surface to the environment of BWR nearer. As the result of adopting the new capsule structure based on the design study, it was found out that feeding water velocity at the surface of specimen's is increased to about 10 times as much as before, and nuclear heat generated in the capsule components can be removed safely even in the abnormal event such as the case of loss of feeding water. (author)

Thermohydraulic design of saturated temperature capsule for IASCC irradiation test

Energy Technology Data Exchange (ETDEWEB)

Ide, Hiroshi; Matsui, Yoshinori; Itabashi, Yukio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

2002-10-01

An advanced water chemistry controlled irradiation research device is being developed in JAERI, to perform irradiation tests for irradiation assisted stress corrosion cracking (IASCC) research concerned with aging of LWR. This device enables the irradiation tests under the water chemistry condition and the temperature, which simulate the conditions for BWR core internals. The advanced water chemistry controlled irradiation research device is composed of saturated temperature capsule inserted into the JMTR core and the water chemistry control unit installed in the reactor building. Regarding the saturated temperature capsule, the Thermohydraulic design of capsule structure was done, aimed at controlling the specimen's temperature, feeding water velocity on specimen's surface to the environment of BWR nearer. As the result of adopting the new capsule structure based on the design study, it was found out that feeding water velocity at the surface of specimen's is increased to about 10 times as much as before, and nuclear heat generated in the capsule components can be removed safely even in the abnormal event such as the case of loss of feeding water. (author)

Temperature dependence of relaxation times in proton components of fatty acids

International Nuclear Information System (INIS)

Kuroda, Kagayaki; Iwabuchi, Taku; Saito, Kensuke; Obara, Makoto; Honda, Masatoshi; Imai, Yutaka

2011-01-01

We examined the temperature dependence of relaxation times in proton components of fatty acids in various samples in vitro at 11 tesla as a standard calibration data for quantitative temperature imaging of fat. The spin-lattice relaxation time, T 1 , of both the methylene (CH 2 ) chain and terminal methyl (CH 3 ) was linearly related to temperature (r>0.98, P 2 signal for calibration and observed the signal with 18% of CH 3 to estimate temperature. These findings suggested that separating the fatty acid components would significantly improve accuracy in quantitative thermometry for fat. Use of the T 1 of CH 2 seems promising in terms of reliability and reproducibility in measuring temperature of fat. (author)

Mechanical components design for PWR - control rod drive mechanism

International Nuclear Information System (INIS)

Leme, Francisco Louzano; Mattar Neto, Miguel

2002-01-01

The Control Rod Drive Mechanism (CRDM) is usually - a high precision - equipment incorporating mechanical and electrical components designed to move the control rods. The 'control rods' refer to all rods or assemblies that are moved to assess the performance of the reactor. The CRDM here presented is the Nut and Lead Screw type. This type is basically a power screw type magnetically coupled to a slow speed reluctance electric motor that provides a means of axially positioning the movable fuel assemblies in the reactor core for purpose of controlling core reactivity. A helically threaded lead screw assembly, comprising one element of power screw, is attached to a movable fuel assemblies. The CRDM usually has closer and more consistent contact with environment peculiar to the reactor than has only other machinery component. This environment includes not only the radiation field of the reactor, but also the temperature, pressure and chemical properties associated with the material used as the coolant for reactor fuel. Specific and special materials are needed because of the above mentioned application. Due to the importance of the above described CRDM functions, this paper will also consider the nuclear functions and their safety classes as well as the CRDM nuclear design criteria. (author)

High-temperature stability of laser-joined silicon carbide components

Energy Technology Data Exchange (ETDEWEB)

Herrmann, Marion, E-mail: marion.herrmann@tu-dresden.de; Lippmann, Wolfgang; Hurtado, Antonio

2013-11-15

Silicon carbide is recommended for applications in energy technology due to its good high-temperature corrosion resistance, mechanical durability, and abrasion resistance. The prerequisite for use is often the availability of suitable technologies for joining or sealing the components. A laser-induced process using fillers and local heating of the components represents a possible low-cost option. Investigations in which yttrium aluminosilicate glass was used for laser-induced brazing of SiC components of varying geometry are presented. A four-point bending strength of 112 MPa was found for these joints. In burst tests, laser-joined components were found to withstand internal pressures of up to 54 MPa. Helium leak tests yielded leak rates of less than 10{sup –8} mbar l s{sup −1}, even after 300 h at 900 °C. In contrast, the assemblies showed an increased leak rate after annealing at 1050 °C. The short process time of the laser technique – in the range of a few seconds to a few minutes – results in high temperature gradients and transients. SEM analysis showed that the filler in the seam predominantly solidifies in a glassy state. Crystallization occurred during later thermal loading of the joined components, with chemical equilibrium being established. Differences in seam structures yielded from different cooling rates in the laser process could not be equalized by annealing. The results demonstrated the long-term stability of laser-brazed SiC assemblies to temperatures in the range of glass transformation (900 °C) of the yttrium aluminosilicate filler. In technological investigations, the suitability of the laser joining technique for sealing of SiC components with a geometry approximating that of a fuel element sleeve pin (pin) in a gas-cooled fast reactor was proven.

Fundamental conceptual design of the experimental multi-purpose high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Shimokawa, Junichi; Yasuno, Takehiko; Yasukawa, Shigeru; Mitake, Susumu; Miyamoto, Yoshiaki

1975-06-01

The fundamental conceptual design of the experimental multi-purpose very high-temperature gas-cooled reactor (experimental VHTR of thermal output 50 MW with reactor outlet-gas temperature 1,000 0 C) has been carried out to provide the operation modes of the system consisting of the reactor and the heat-utilization system, including characteristics and performance of the components and safety of the plant system. For the heat-utilization system of the plant, heat distribution, temperature condition, cooling system constitution, and the containment facility are specified. For the operation of plant, testing capability of the reactor and controlability of the system are taken into consideration. Detail design is made of the fuel element, reactor core, reactivity control and pressure vessel, and also the heat exchanger, steam reformer, steam generator, helium circulator, helium-gas turbine, and helium-gas purification, fuel handling, and engineered safety systems. Emphasis is placed on providing the increase of the reactor outlet-gas temperature. Fuel element design is directed to the prismatic graphite blocks of hexagonal cross-section accommodating the hollow or tubular fuel pins sheathed in graphite sleeve. The reactor core is composed of 73 fuel columns in 7 stages, concerning the reference design MK-II. Orificing is made in the upper portion of core; one orifice for every 7 fuel columns. Average core power density is 2.5 watts/cm 3 . Fuel temperature is kept below 1,300 0 C in rated power. The main components, i.e. pressure vessel, reformer, gas turbine and intermediate heat exchanger are designed in detail; the IHX is of a double-shell and helically-wound tube coils, the reformer is of a byonet tube type, and the turbine-compressor unit is of an axial flow type (turbine in 6 stages and compressor in 16 stages). (auth.)

Design manual. [High temperature heat pump for heat recovery system

Energy Technology Data Exchange (ETDEWEB)

Burch, T.E.; Chancellor, P.D.; Dyer, D.F.; Maples, G.

1980-01-01

The design and performance of a waste heat recovery system which utilizes a high temperature heat pump and which is intended for use in those industries incorporating indirect drying processes are described. It is estimated that use of this heat recovery system in the paper, pulp, and textile industries in the US could save 3.9 x 10/sup 14/ Btu/yr. Information is included on over all and component design for the heat pump system, comparison of prime movers for powering the compressor, control equipment, and system economics. (LCL)

Robust design of microelectronics assemblies against mechanical shock, temperature and moisture effects of temperature, moisture and mechanical driving forces

CERN Document Server

Wong, E-H

2015-01-01

Robust Design of Microelectronics Assemblies Against Mechanical Shock, Temperature and Moisture discusses how the reliability of packaging components is a prime concern to electronics manufacturers. The text presents a thorough review of this important field of research, providing users with a practical guide that discusses theoretical aspects, experimental results, and modeling techniques. The authors use their extensive experience to produce detailed chapters covering temperature, moisture, and mechanical shock induced failure, adhesive interconnects, and viscoelasticity. Useful progr

Construction of a 21-Component Layered Mixture Experiment Design Using a New Mixture Coordinate-Exchange Algorithm

International Nuclear Information System (INIS)

Piepel, Gregory F.; Cooley, Scott K.; Jones, Bradley

2005-01-01

This paper describes the solution to a unique and challenging mixture experiment design problem involving: (1) 19 and 21 components for two different parts of the design, (2) many single-component and multi-component constraints, (3) augmentation of existing data, (4) a layered design developed in stages, and (5) a no-candidate-point optimal design approach. The problem involved studying the liquidus temperature of spinel crystals as a function of nuclear waste glass composition. The statistical objective was to develop an experimental design by augmenting existing glasses with new nonradioactive and radioactive glasses chosen to cover the designated nonradioactive and radioactive experimental regions. The existing 144 glasses were expressed as 19-component nonradioactive compositions and then augmented with 40 new nonradioactive glasses. These included 8 glasses on the outer layer of the region, 27 glasses on an inner layer, 2 replicate glasses at the centroid, and one replicate each of three existing glasses. Then, the 144 + 40 = 184 glasses were expressed as 21-component radioactive compositions, and augmented with 5 radioactive glasses. A D-optimal design algorithm was used to select the new outer layer, inner layer, and radioactive glasses. Several statistical software packages can generate D-optimal experimental designs, but nearly all of them require a set of candidate points (e.g., vertices) from which to select design points. The large number of components (19 or 21) and many constraints made it impossible to generate the huge number of vertices and other typical candidate points. JMP was used to select design points without candidate points. JMP uses a coordinate-exchange algorithm modified for mixture experiments, which is discussed and illustrated in the paper

Design standard issues for ITER in-vessel components

International Nuclear Information System (INIS)

Majumdar, S.

1994-01-01

Unique requirements that must be addressed by a structural design code for the ITER have been summarized. Existing codes such as ASME Section III, or the French RCC-MR were developed primarily for fission reactor out-of-core components and are not directly applicable to the ITER. They may be used either as a guide for developing a design code for the ITER or as interim standards. However, new rules will be needed for handling the irradiation-induced embrittlement problems faced by the ITER blanket components. Design standards developed in the past for the design of fission reactor core components in the United States can be used as guides in this area

Process weakness assessment by profiling all incoming design components

Science.gov (United States)

Zhuang, Linda; Cai, MengFeng; Zhu, Annie; Zhang, Yifan; Sweis, Jason; Lai, Ya-Chieh

2017-03-01

Foundries normally receive a large number of designs from different customers every day. It is desired to automatically profile each incoming design to quantify certain metrics like 1) the number of polygons per GDS layers 2) what kind of electrical components the design contains 3) what the dimensions of each electrical component are 4) how frequently any size of components have been used and their physical locations. This paper will present a novel method of how to generate a complete profile of components for any particular design. The component checking flow need to be completed within hours so it will have very little impact on the tape-out time. A pre-layer checking method is also run to group commonly used layers for different electrical components and then employ different layout profiling flows. The foundry does this design chip analysis in order to find potentially weak devices due to their size or special size requirements for particular electrical components. The foundry can then take pre-emptive action to avoid yield loss or make an unnecessary mask for new incoming products before fab processing starts.

Mechanical testing - designers need: a view at component design and operations stages

International Nuclear Information System (INIS)

Shrivastava, S.K.

2007-01-01

Mechanical design of any component requires knowledge of values of various material properties which designer(s) make(s) use in designing the component. In design of nuclear power plant components, it assumes even greater importance in view of degree of precision and accuracy with which the values of various properties are required. This is in turn demands, high accuracy in testing machines and measuring methods. In this paper, attempt has been made to bring out that even from conventional tension test, how designer today looks for availability of engineering stress-strain diagram preferably through digitally acquired data points during the test from which he can derive values of Ramberg-Osgood parameters for use in fracture mechanics based analysis. Attempt has been also made to provide account of some of important fracture mechanics related tests which have been evolved in last two decades and designers need for evolution of simple test techniques to measure many more fracture mechanics related parameters as well as cater to constraints such as shape and size of material available from the components. Nuclear power plant has been primarily kept in view and ASME. Section III NB, ASME Section XI and relevant ASTM Standards have been taken as standard references. Further pressure retaining materials of pressure vessels/Reactor Pressure Vessels have been kept in view. (author)

Implementation of constitutive equations for creep damage mechanics into the ABAQUS finite element code - some practical cases in high temperature component design and life assessment

International Nuclear Information System (INIS)

Segle, P.; Samuelson, L.Aa.; Andersson, Peder; Moberg, F.

1996-01-01

Constitutive equations for creep damage mechanics are implemented into the finite element program ABAQUS using a user supplied subroutine, UMAT. A modified Kachanov-Rabotnov constitutive equation which accounts for inhomogeneity in creep damage is used. With a user defined material a number of bench mark tests are analyzed for verification. In the cases where analytical solutions exist, the numerical results agree very well. In other cases, the creep damage evolution response appear to be realistic in comparison with laboratory creep tests. The appropriateness of using the creep damage mechanics concept in design and life assessment of high temperature components is demonstrated. 18 refs

An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components

Energy Technology Data Exchange (ETDEWEB)

Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

2009-11-01

This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

Evaluation of internal boiler components and gases using a high-temperature infrared (IR) lens

Science.gov (United States)

Hammaker, Robert G.; Colsher, Richard J.; Miles, Jonathan J.; Madding, Robert P.

1996-03-01

Fuel accounts for an average of seventy percent of the yearly operational and maintenance costs of all the fossil stations in the United States. This amounts to 30 billion dollars spent for fuel each year. In addition, federal and state environmental codes have been enforcing stricter regulations that demand cleaner environments, such as the reduction of nitrogen oxides (NOx), which are a by-product of the fossil fuel flame. If the burn of the flame inside a boiler could be optimized, the usage of fuel and the amounts of pollution produced would be significantly reduced, and many of the common boiler tube failures can be avoided. This would result in a major dollar savings to the utility industry, and would provide a cleaner environment. Accomplishing these goals will require a major effort from the designers and operators that manufacture, operate, and maintain the fossil stations. Over the past few years re-designed burners have been installed in many boilers to help control the temperatures and shape of the flame for better performance and NOx reduction. However, the measurement of the processes and components inside the furnace, that could assist in determining the desired conditions, can at times be very difficult due to the hostile hot environment. In an attempt to resolve these problems, the EPRI M&D Center and a core group of EPRI member utilities have undertaken a two-year project with various optical manufacturers, IR manufacturers, and IR specialists, to fully develop an optical lens that will withstand the high furnace temperatures. The purpose of the lens is to explore the possibilities of making accurate high temperature measurements of the furnace processes and components in an ever-changing harsh environment. This paper provides an introduction to EPRI's internal boiler investigation using an IR high temperature lens (HTL). The paper describes the objectives, approach, benefits, and project progress.

Design and Verification of Fault-Tolerant Components

DEFF Research Database (Denmark)

Zhang, Miaomiao; Liu, Zhiming; Ravn, Anders Peter

2009-01-01

We present a systematic approach to design and verification of fault-tolerant components with real-time properties as found in embedded systems. A state machine model of the correct component is augmented with internal transitions that represent hypothesized faults. Also, constraints...... to model and check this design. Model checking uses concrete parameters, so we extend the result with parametric analysis using abstractions of the automata in a rigorous verification....... relatively detailed such that they can serve directly as blueprints for engineering, and yet be amenable to exhaustive verication. The approach is illustrated with a design of a triple modular fault-tolerant system that is a real case we received from our collaborators in the aerospace field. We use UPPAAL...

Evaluation test of high temperature strain gages used in a stethoscope for OGL-1 components in an elevated temperature service

Energy Technology Data Exchange (ETDEWEB)

Sato, Toshimi (Kyowa Electronic Inst. Co. Ltd. (Japan)); Tanaka, Isao; Komori, Yoshihiro; Suzuki; Toshiaki

1982-08-01

The stethoscope for OGL-1 components in a elevated temperature service (SOCETS) is a measuring system of evaluation integrity of structures for high temperature pipings during operations of Japan Material Testing Reactor. This paper is described about the results on fundamental performance on high temperature strain gages. From their test results that have been based on correlation of temperature-timestrain factors, it became clear that two weldable strain gages and a capacitance strain gage were available for strain measurements of OGL-1 components.

Design, Analysis and R&D of the EAST In-Vessel Components

Science.gov (United States)

Yao, Damao; Bao, Liman; Li, Jiangang; Song, Yuntao; Chen, Wenge; Du, Shijun; Hu, Qingsheng; Wei, Jing; Xie, Han; Liu, Xufeng; Cao, Lei; Zhou, Zibo; Chen, Junling; Mao, Xinqiao; Wang, Shengming; Zhu, Ning; Weng, Peide; Wan, Yuanxi

2008-06-01

In-vessel components are important parts of the EAST superconducting tokamak. They include the plasma facing components, passive plates, cryo-pumps, in-vessel coils, etc. The structural design, analysis and related R&D have been completed. The divertor is designed in an up-down symmetric configuration to accommodate both double null and single null plasma operation. Passive plates are used for plasma movement control. In-vessel coils are used for the active control of plasma vertical movements. Each cryo-pump can provide an approximately 45 m3/s pumping rate at a pressure of 10-1 Pa for particle exhaust. Analysis shows that, when a plasma current of 1 MA disrupts in 3 ms, the EM loads caused by the eddy current and the halo current in a vertical displacement event (VDE) will not generate an unacceptable stress on the divertor structure. The bolted divertor thermal structure with an active cooling system can sustain a load of 2 MW/m2 up to a 60 s operation if the plasma facing surface temperature is limited to 1500 °C. Thermal testing and structural optimization testing were conducted to demonstrate the analysis results.

Design rules for high temperature plant - the implications of recent research in relation to current practice

International Nuclear Information System (INIS)

Townley, C.H.A.

1977-01-01

An historical summary is presented of design rules for high temperature plant, leading to the rules applicable to high temperature reactors, particularly the liquid metal fast breeder reactor. Special attention is given to creep rupture properties of ferritic and austenitic materials used for the construction of components such as boilers and pressure vessels. (author)

Evaluation test of high temperature strain gages used in a stethoscope for OGL-1 components in an elevated temperature service

International Nuclear Information System (INIS)

Sato, Toshimi; Tanaka, Isao; Komori, Yoshihiro; Suzuki; Toshiaki.

1982-01-01

The stethoscope for OGL-1 components in a elevated temperature service (SOCETS) is a measuring system of evaluation integrity of structures for high temperature pipings during operations of Japan Material Testing Reactor. This paper is described about the results on fundamental performance on high temperature strain gages. From their test results that have been based on correlation of temperature-timestrain factors, it became clear that two weldable strain gages and a capacitance strain gage were available for strain measurements of OGL-1 components. (author)

High-temperature, high-pressure bonding of nested tubular metallic components

International Nuclear Information System (INIS)

Quinby, T.C.

1980-01-01

This invention is a tool for effecting high-temperature, high compression bonding between the confronting faces of nested, tubular, metallic components. In a typical application, the tool is used to produce tubular target assemblies for irradiation in nuclear reactors or particle accelerators, the target assembly comprising a uranium foil and an aluminum-alloy substrate. The tool preferably is composed throughout of graphite. It comprises a tubular restraining member in which a mechanically expandable tubular core is mounted to form an annulus with the member. The components to be bonded are mounted in nested relation in the annulus. The expandable core is formed of individually movable, axially elongated segments whose outer faces cooperatively define a cylindrical pressing surface and whose inner faces cooperatively define two opposed, inwardly tapered, axial bores. Tapered rams extend respectively into the bores. The loaded tool is mounted in a conventional hot-press provided with evacuation means, heaters for maintaining its interior at bonding temperature, and hydraulic cylinders for maintaining a selected inwardly directed pressure on the tapered rams. With the hotpress evacuated and the loaded tool at the desired temperature, the cylinders are actuated to apply the selected pressure to the rams. The rams in turn expand the segmented core to maintain the nested components in compression against the restraining member. These conditions are maintained until the confronting faces of the nested components are joined in a continuous, uniform bond characterized by high thermal conductivity

High-temperature, high-pressure bonding of nested tubular metallic components

Science.gov (United States)

Quinby, T.C.

A tool is described for effecting high-temperature, high-compression bonding between the confronting faces of nested, tubular, metallic components. In a typical application, the tool is used to produce tubular target assemblies for irradiation in nuclear reactors or particle accelerators. The target assembly comprising a uranum foil and an aluninum-alloy substrate. The tool is composed of graphite. It comprises a tubular restraining member in which a mechanically expandable tubular core is mounted to form an annulus. The components to be bonded are mounted in nested relation in the annulus. The expandable core is formed of individually movable, axially elongated segments whose outer faces cooperatively define a cylindrical pressing surface and whose inner faces cooperatively define two opposed, inwardly tapered, axial bores. Tapered rams extend into the bores. The loaded tool is mounted in a conventional hot-press provided with evacuation means, heaters for maintaining its interior at bonding temperature, and hydraulic cylinders for maintaining a selected inwardly directed pressure on the tapered rams. With the hot-press evacuated and the loaded tool at the desired temperature, the cylinders are actuated to apply the selected pressure to the rams. The rams in turn expand the segmented core to maintain the nested components in compression against the restraining member. These conditions are maintained until the confronting faces of the nested components are joined in a continuous, uniform bond characterized by high thermal conductivity.

Preliminary design of high temperature ultrasonic transducers for liquid sodium environments

Science.gov (United States)

Prowant, M. S.; Dib, G.; Qiao, H.; Good, M. S.; Larche, M. R.; Sexton, S. S.; Ramuhalli, P.

2018-04-01

Advanced reactor concepts include fast reactors (including sodium-cooled fast reactors), gas-cooled reactors, and molten-salt reactors. Common to these concepts is a higher operating temperature (when compared to light-water-cooled reactors), and the proposed use of new alloys with which there is limited operational experience. Concerns about new degradation mechanisms, such as high-temperature creep and creep fatigue, that are not encountered in the light-water fleet and longer operating cycles between refueling intervals indicate the need for condition monitoring technology. Specific needs in this context include periodic in-service inspection technology for the detection and sizing of cracking, as well as technologies for continuous monitoring of components using in situ probes. This paper will discuss research on the development and evaluation of high temperature (>550°C; >1022°F) ultrasonic probes that can be used for continuous monitoring of components. The focus of this work is on probes that are compatible with a liquid sodium-cooled reactor environment, where the core outlet temperatures can reach 550°C (1022°F). Modeling to assess sensitivity of various sensor configurations and experimental evaluation have pointed to a preferred design and concept of operations for these probes. This paper will describe these studies and ongoing work to fabricate and fully evaluate survivability and sensor performance over extended periods at operational temperatures.

Design criteria for high-temperature-affected, metallic and ceramic components, and for the prestressed concrete reactor pressure vessel of future HTR systems. Final report. Vol. 1-4

International Nuclear Information System (INIS)

1988-08-01

This work in five separate volumes reports on the elaboration of basic data for the formulation of design criteria for HTR components and is arranged into the four following subject areas : (1) safety-specific limiting conditions; (2) metallic components; (3) prestressed concrete reactor pressure vessels; (4) graphitic reactor internals. Under item 2, the mechanical and physical characteristics of the materials X20CrMoV 12 1, X10NiCrAlTi 32 20, and NiCr23Co12Mo are examined up to temperatures of 950deg C. Stress-strain rate laws are elaborated for description of the inelastic deformation behavior. The representation of the subject area reactor pressure vessels deals with four main topics: Prestressed concrete support structure, liner, vessel closures, thermal protection system. Quality-assurance classes are defined under item 4 for graphitic components and load levels for load categories. The material evaluation is discussed in detail (e.g. manufacturing monitoring from the raw material to the graphitization and manufacturing testing up to the acceptance test). In addition, the corrosion behavior and irradiation behavior of graphite is examined and rules for computation of stresses in irradiated and unirradiated graphitic components are elaborated. (MM) [de

Beamline standard component designs for the Advanced Photon Source

International Nuclear Information System (INIS)

Shu, D.; Barraza, J.; Brite, C.; Chang, J.; Sanchez, T.; Tcheskidov, V.; Kuzay, T.M.

1994-01-01

The Advanced Photon Source (APS) has initiated a design standardization and modularization activity for the APS synchrotron radiation beamline components. These standard components are included in components library, sub-components library and experimental station library. This paper briefly describes these standard components using both technical specifications and side view drawings

Effects of seawater components on radiolysis of water at elevated temperature

International Nuclear Information System (INIS)

Wada, Yoichi; Tachibana, Masahiko; Ishida, Kazushige; Ota, Nobuyuki; Shigenaka, Naoto; Inagaki, Hiromitsu; Noda, Hiroshi

2014-01-01

Effects of seawater components on radiolysis of water at elevated temperature have been studied with a radiolysis model in order to evaluate influence on integrity of materials used in an ABWR. In 2011, seawater flowed into a wide part of the nuclear power plant system of the Hamaoka Nuclear Power Station Reactor No. 5 owned by Chubu Electric Power Co., Inc. after condenser tubes broke during the plant shutdown operation. The reactor water temperature was 250°C and its maximum Cl − concentration was ca. 450 ppm when seawater was mixed with reactor water. In order to clarify effects of the sea water components on radiolysis of water at elevated temperature, a radiolysis model calculation was conducted with Hitachi's radiolysis analysis code 'SIMFONY'. For the calculation, the temperature range was set from 50 to 250°C with 50°C increments and the gamma dose rate was set at 60 Gys −1 to see the effect of gamma irradiation from fuels under shutdown conditions. Concentrations of radiolytic species were calculated for 10 5 s. Dilution ratio of seawater was changed to see the effects of concentration of seawater components. Reaction rate constants of the Cl − , Br − , HCO 3 − , and SO 4 2− systems were considered. The main radiolytic species were predicted to be hydrogen and oxygen. Hydrogen peroxide of low concentration was produced in seawater-mixed water at elevated temperatures. Compared with these main products, concentrations of radiolytic products originating from chloride ion and other seawater components were found to be rather low. The dominant product among them was ClO 3 − and its concentration was found to be below 0.01ppm at 10 5 s. Then, during the plant shutdown operation, the harmful influence from radiolytic species originating from seawater components on integrity of fuel materials must be smaller than that of chloride ion which is the main ionic species in seawater. (author)

Nuclear heat source component design considerations for HTGR process heat reactor plant concept

International Nuclear Information System (INIS)

McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

1982-05-01

The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported [via an intermediate heat exchanger (IHX)] to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

Views on seismic design standardization of structures, systems and components of nuclear facilities

International Nuclear Information System (INIS)

Reddy, G.R.

2011-01-01

Structures, Systems and Components (SSCs) of nuclear facilities have to be designed for normal operating loads such as dead weight, pressure, temperature etc., and accidental loads such as earthquakes, floods, extreme, wind air craft impact, explosions etc. Manmade accidents such as aircraft impact, explosions etc., sometimes may be considered as design basis event and sometimes taken care by providing administrative controls. This will not be possible in the case of natural events such as earthquakes, flooding, extreme winds etc. Among natural events earthquakes are considered as most devastating and need to be considered as design basis event which has certain annual frequency specified in design codes. For example nuclear power plants are designed for a seismic event has 10000 year return period. It is generally felt that design of SSCs for earthquake loads is very time consuming and expensive. Conventional seismic design approaches demands for large number of supports for systems and components. This results in large space occupation and in turn creates difficulties for maintenance and in service inspection of systems and components. In addition, complete exercise of design need to be repeated for plants being located at different sites due to different seismic demands. However, advanced seismic response control methods will help to standardize the seismic design meeting the safety and economy. These methods adopt passive, semi active and active devices, and base isolators to control the seismic response. In nuclear industry, it is advisable to go for passive devices to control the seismic responses. Ideally speaking, these methods will make the designs made for normal loads can also satisfy the seismic demand without calling for change in material, geometry, layout etc. in the SSCs. This paper explain the basic ideas of seismic response control methods, demonstrate the effectiveness of control methods through case studies and eventually give the procedure to

The Composition and Temperature Effects on the Ultra High Strength Stainless Steel Design

Science.gov (United States)

Xu, W.; Del Castillo, P. E. J. Rivera Díaz; van der Zwaag, S.

Alloy composition and heat treatment are of paramount importance to determining alloy properties. Their control is of great importance for new alloy design and industrial fabrication control. A base alloy utilizing MX carbide is designed through a theory guided computational approach coupling a genetic algorithm with optimization criteria based on thermodynamic, kinetic and mechanical principles. The combined effects of 11 alloying elements (Al, C, Co, Cr, Cu, Mo, Nb, Ni, Si, Ti and V) are investigated in terms of the composition optimization criteria: the martensite start (Ms) temperature, the suppression of undesirable phases, the Cr concentration in the matrix and the potency of the precipitation strengthening contribution. The results show the concentration sensitivities of each component and also point out new potential composition domains for further strength increase. The aging temperature effect is studied and the aging temperature industrially followed is recovered.

Design of smart sensing components for volcano monitoring

Science.gov (United States)

Xu, M.; Song, W.-Z.; Huang, R.; Peng, Y.; Shirazi, B.; LaHusen, R.; Kiely, A.; Peterson, N.; Ma, A.; Anusuya-Rangappa, L.; Miceli, M.; McBride, D.

2009-01-01

In a volcano monitoring application, various geophysical and geochemical sensors generate continuous high-fidelity data, and there is a compelling need for real-time raw data for volcano eruption prediction research. It requires the network to support network synchronized sampling, online configurable sensing and situation awareness, which pose significant challenges on sensing component design. Ideally, the resource usages shall be driven by the environment and node situations, and the data quality is optimized under resource constraints. In this paper, we present our smart sensing component design, including hybrid time synchronization, configurable sensing, and situation awareness. Both design details and evaluation results are presented to show their efficiency. Although the presented design is for a volcano monitoring application, its design philosophy and framework can also apply to other similar applications and platforms. ?? 2009 Elsevier B.V.

Waste Package Component Design Methodology Report

International Nuclear Information System (INIS)

D.C. Mecham

2004-01-01

This Executive Summary provides an overview of the methodology being used by the Yucca Mountain Project (YMP) to design waste packages and ancillary components. This summary information is intended for readers with general interest, but also provides technical readers a general framework surrounding a variety of technical details provided in the main body of the report. The purpose of this report is to document and ensure appropriate design methods are used in the design of waste packages and ancillary components (the drip shields and emplacement pallets). The methodology includes identification of necessary design inputs, justification of design assumptions, and use of appropriate analysis methods, and computational tools. This design work is subject to ''Quality Assurance Requirements and Description''. The document is primarily intended for internal use and technical guidance for a variety of design activities. It is recognized that a wide audience including project management, the U.S. Department of Energy (DOE), the U.S. Nuclear Regulatory Commission, and others are interested to various levels of detail in the design methods and therefore covers a wide range of topics at varying levels of detail. Due to the preliminary nature of the design, readers can expect to encounter varied levels of detail in the body of the report. It is expected that technical information used as input to design documents will be verified and taken from the latest versions of reference sources given herein. This revision of the methodology report has evolved with changes in the waste package, drip shield, and emplacement pallet designs over many years and may be further revised as the design is finalized. Different components and analyses are at different stages of development. Some parts of the report are detailed, while other less detailed parts are likely to undergo further refinement. The design methodology is intended to provide designs that satisfy the safety and operational

Waste Package Component Design Methodology Report

Energy Technology Data Exchange (ETDEWEB)

D.C. Mecham

2004-07-12

This Executive Summary provides an overview of the methodology being used by the Yucca Mountain Project (YMP) to design waste packages and ancillary components. This summary information is intended for readers with general interest, but also provides technical readers a general framework surrounding a variety of technical details provided in the main body of the report. The purpose of this report is to document and ensure appropriate design methods are used in the design of waste packages and ancillary components (the drip shields and emplacement pallets). The methodology includes identification of necessary design inputs, justification of design assumptions, and use of appropriate analysis methods, and computational tools. This design work is subject to ''Quality Assurance Requirements and Description''. The document is primarily intended for internal use and technical guidance for a variety of design activities. It is recognized that a wide audience including project management, the U.S. Department of Energy (DOE), the U.S. Nuclear Regulatory Commission, and others are interested to various levels of detail in the design methods and therefore covers a wide range of topics at varying levels of detail. Due to the preliminary nature of the design, readers can expect to encounter varied levels of detail in the body of the report. It is expected that technical information used as input to design documents will be verified and taken from the latest versions of reference sources given herein. This revision of the methodology report has evolved with changes in the waste package, drip shield, and emplacement pallet designs over many years and may be further revised as the design is finalized. Different components and analyses are at different stages of development. Some parts of the report are detailed, while other less detailed parts are likely to undergo further refinement. The design methodology is intended to provide designs that satisfy the safety

Stainless steel component with compressed fiber Bragg grating for high temperature sensing applications

Science.gov (United States)

Jinesh, Mathew; MacPherson, William N.; Hand, Duncan P.; Maier, Robert R. J.

2016-05-01

A smart metal component having the potential for high temperature strain sensing capability is reported. The stainless steel (SS316) structure is made by selective laser melting (SLM). A fiber Bragg grating (FBG) is embedded in to a 3D printed U-groove by high temperature brazing using a silver based alloy, achieving an axial FBG compression of 13 millistrain at room temperature. Initial results shows that the test component can be used for up to 700°C for sensing applications.

LEDA RF distribution system design and component test results

International Nuclear Information System (INIS)

Roybal, W.T.; Rees, D.E.; Borchert, H.L.; McCarthy, M.; Toole, L.

1998-01-01

The 350 MHz and 700 MHz RF distribution systems for the Low Energy Demonstration Accelerator (LEDA) have been designed and are currently being installed at Los Alamos National Laboratory. Since 350 MHz is a familiar frequency used at other accelerator facilities, most of the major high-power components were available. The 700 MHz, 1.0 MW, CW RF delivery system designed for LEDA is a new development. Therefore, high-power circulators, waterloads, phase shifters, switches, and harmonic filters had to be designed and built for this applications. The final Accelerator Production of Tritium (APT) RF distribution systems design will be based on much of the same technology as the LEDA systems and will have many of the RF components tested for LEDA incorporated into the design. Low power and high-power tests performed on various components of these LEDA systems and their results are presented here

Design of plasma facing components for the SST-1 tokamak

International Nuclear Information System (INIS)

Jacob, S.; Chenna Reddy, D.; Choudhury, P.; Khirwadkar, S.; Pragash, R.; Santra, P.; Saxena, Y.C.; Sinha, P.

2000-01-01

Steady state Superconducting Tokamak, SST-1, is a medium sized tokamak with major and minor radii of 1.10 m and 0.20 m respectively. Elongated plasma operation with double null poloidal divertor is planned with a maximum input power of 1 MW. The Plasma Facing Components (PFC) like Divertors and Baffles, Poloidal limiters and Passive stabilizers form the first material boundary around the plasma and hence receive high heat and particle fluxes. The PFC design should ensure efficient heat and particle removal during steady state tokamak operation. A closed divertor geometry is adopted to ensure high neutral pressure in the divertor region (and hence high recycling) and less impurity influx into the core plasma. A set of poloidal limiters are provided to assist break down, current ramp-up and current ramp down phases and for the protection of the in-vessel components. Two pairs of Passive stabilizers, one on the inboard and the other on the outboard side of the plasma, are provided to slow down the vertical instability growth rates of the shaped plasma column. All PFCs are actively cooled to keep the plasma facing surface temperature within the design limits. The PFCs have been shaped/profiled so that maximum steady state heat flux on the surface is less than 1 MW/m 2 . (author)

Design of sodium cooled reactor systems and components for maintainability

International Nuclear Information System (INIS)

Carr, R.W.; Charnock, H.O.; McBride, J.P.

1978-09-01

Special maintenability problems associated with the design and operation of sodium cooled reactor plants are discussed. Some examples of both good and bad design practice are introduced from the design of the FFTF plant and other plants. Subjects include design for drainage, cleaning, decontamination, access, component removal, component disassembly and reassembly, remote tooling, jigs, fixtures, and design for minimizing radiation exposure of maintenance personnel. Check lists are included

Main components of business cards design

Directory of Open Access Journals (Sweden)

Ю. В. Романенкова

2003-03-01

Full Text Available The essay is dedicated to the urgent problem of necessity of creation of professional design of business cards, that are important part of the image of modem businessman. There are classification of cards by functional principle, the functions of cards of each type were analyzed. All components of business card, variants of its composition schemes, color characteristics, principles of use of trade marks and other design elements have been allocated

Recommended practices in elevated temperature design: A compendium of breeder reactor experiences (1970-1986): An overview

International Nuclear Information System (INIS)

Wei, B.C.; Cooper, W.L. Jr.; Dhalla, A.K.

1987-09-01

Significant experiences have been accumulated in the establishment of design methods and criteria applicable to the design of Liquid Metal Fast Breeder Reactor (LMFBR) components. The Subcommittee of the Elevated Temperature Design under the Pressure Vessel Research Council (PVRC) has undertaken to collect, on an international basis, design experience gained, and the lessons learned, to provide guidelines for next generation advanced reactor designs. This paper shall present an overview and describe the highlights of the work

Evaluation of weldment creep and fatigue strength-reduction factors for elevated-temperature design

International Nuclear Information System (INIS)

Corum, J.M.

1989-01-01

New explicit weldment strength criteria in the form of creep and fatigue strength-reduction factors were recently introduced into the American Society of Mechanical Engineers Code Case N-47, which governs the design of elevated-temperature nuclear plants components in the United States. This paper provides some of the background and logic for these factors and their use, and it describes the results of a series of long-term, confirmatory, creep-rupture and fatigue tests of simple welded structures. The structures (welded plates and tubes) were made of 316 stainless steel base metal and 16-8-2 weld filler metal. Overall, the results provide further substantiation of the validity of the strength-reduction factor approach for ensuring adequate life in elevated-temperature nuclear component weldments. 16 refs., 7 figs

Monitoring of coolant temperature stratification on piping components in WWER-440 NPPs

International Nuclear Information System (INIS)

Hudcovsky, S.; Slanina, M.; Badiar, S.

2001-01-01

The presentation deals with the aims of non-standard temperature measurements installed on primary and secondary circuit in WWER-440 NPPs, explains reasons of coolant temperature stratification on the piping components. It describes methods of the measurements on pipings, range of installation of the temperature measurements in EBO and EMO units and illustrates results of measurements of coolant temperature stratification. (Authors)

Modelling temporal variance of component temperatures and directional anisotropy over vegetated canopy

Science.gov (United States)

Bian, Zunjian; du, yongming; li, hua

2016-04-01

Land surface temperature (LST) as a key variable plays an important role on hydrological, meteorology and climatological study. Thermal infrared directional anisotropy is one of essential factors to LST retrieval and application on longwave radiance estimation. Many approaches have been proposed to estimate directional brightness temperatures (DBT) over natural and urban surfaces. While less efforts focus on 3-D scene and the surface component temperatures used in DBT models are quiet difficult to acquire. Therefor a combined 3-D model of TRGM (Thermal-region Radiosity-Graphics combined Model) and energy balance method is proposed in the paper for the attempt of synchronously simulation of component temperatures and DBT in the row planted canopy. The surface thermodynamic equilibrium can be final determined by the iteration strategy of TRGM and energy balance method. The combined model was validated by the top-of-canopy DBTs using airborne observations. The results indicated that the proposed model performs well on the simulation of directional anisotropy, especially the hotspot effect. Though we find that the model overestimate the DBT with Bias of 1.2K, it can be an option as a data reference to study temporal variance of component temperatures and DBTs when field measurement is inaccessible

A Components Database Design and Implementation for Accelerators and Detectors

International Nuclear Information System (INIS)

Chan, A.; Meyer, S.

2011-01-01

Many accelerator and detector systems being fabricated for the PEP-II Accelerator and BABAR Detector needed configuration control and calibration measurements tracked for their components. Instead of building a database for each distinct system, a Components Database was designed and implemented that can encompass any type of component and any type of measurement. In this paper we describe this database design that is especially suited for the engineering and fabrication processes of the accelerator and detector environments where there are thousands of unique component types. We give examples of information stored in the Components Database, which includes accelerator configuration, calibration measurements, fabrication history, design specifications, inventory, etc. The World Wide Web interface is used to access the data, and templates are available for international collaborations to collect data off-line.

Methods for very high temperature design

International Nuclear Information System (INIS)

Blass, J.J.; Corum, J.M.; Chang, S.J.

1989-01-01

Design rules and procedures for high-temperature, gas-cooled reactor components are being formulated as an ASME Boiler and Pressure Vessel Code Case. A draft of the Case, patterned after Code Case N-47, and limited to Inconel 617 and temperatures of 982/degree/C (1800/degree/F) or less, will be completed in 1989 for consideration by relevant Code committees. The purpose of this paper is to provide a synopsis of the significant differences between the draft Case and N-47, and to provide more complete accounts of the development of allowable stress and stress rupture values and the development of isochronous stress vs strain curves, in both of which Oak Ridge National Laboratory (ORNL) played a principal role. The isochronous curves, which represent average behavior for many heats of Inconel 617, were based in part on a unified constitutive model developed at ORNL. Details are also provided of this model of inelastic deformation behavior, which does not distinguish between rate-dependent plasticity and time-dependent creep, along with comparisons between calculated and observed results of tests conducted on a typical heat of Inconel 617 by the General Electric Company for the Department of Energy. 4 refs., 15 figs., 1 tab

Design, Manufacturing and Integration of LHC Cryostat Components an Example of Collaboration between CERN and Industry

CERN Document Server

Slits, Ivo; Canetti, Marco; Colombet, Thierry; Gangini, Fabrizio; Parma, Vittorio; Tock, Jean-Philippe

2006-01-01

The components for the LHC cryostats and interconnections are supplied by European industry. The manufacturing, assembly and testing of these components in accordance with CERN technical specifications require a close collaboration and dedicated approach from the suppliers. This paper presents the different phases of design, manufacturing, testing and integration of four LHC cryostat components supplied by RIAL Vacuum (Parma, Italy), including 112 Insulation Vacuum Barriers (IVB), 482 Cold-mass Extension Tubes (CET), 121 cryostat vacuum vessel Jumper Elbows (JE) and 10800 Interconnection Sleeves (IS). The Quality Assurance Plan, which the four projects have in common, is outlined. The components are all leak-tight thin stainless steel assemblies (<10-8 mbar l/s), most of them operating at cryogenic temperature (2 K), however each having specific requirements. The particularities of each component are presented with respect to manufacturing, assembly and testing. These components are being integrated ...

Ratcheting study in pressurized piping components under cyclic loading at room temperature

International Nuclear Information System (INIS)

Ravi Kiran, A.; Agrawal, M.K.; Reddy, G.R.; Vaze, K.K.; Ghosh, A.K.; Kushwaha, H.S.

2006-07-01

The nuclear power plant piping components and systems are often subjected to reversing cyclic loading conditions due to various process transients, seismic and other events. Earlier the design of piping subjected to seismic excitation was based on the principle of plastic collapse. It is believed that during such events, fatigue-ratcheting is likely mode of failure of piping components. The 1995 ASME Boiler and Pressure Vessel code, Section-III, has incorporated the reverse dynamic loading and ratcheting into the code. Experimental and analytical studies are carried out to understand this failure mechanism. The biaxial ratcheting characteristics of SA 333, Gr. 6 steel and SS 304 stainless steel at room temperature are investigated in the present work. Experiments are carried out on straight pipes subjected to internal pressure and cyclic bending load applied in a three point and four point bend test configurations. A shake table test is also carried out on a pressurized elbow by applying sinusoidal base excitation. Analytical simulation of ratcheting in the piping elements is carried out. Chaboche nonlinear kinematic hardening model is used for ratcheting simulation. (author)

Application of high temperature superconductivity to electric motor design

International Nuclear Information System (INIS)

Edmonds, J.S.; Sharma, D.K.; Jordan, H.E.; Edick, J.D.; Schiferl, R.F.

1992-01-01

This paper reports on progress made in a joint project conducted by the Electric Power Research Institute and Reliance Electric Company to study the possible application of High Temperature Super Conductors (HTSC), materials to electric motors. Specific applications are identified which can be beneficially served by motors constructed with HTSC materials. A summary is presented of the components and design issues related to HTSC motors designed for these applications. During the course of this development program, a three tier HTSC wire performance specification has evolved. The three specifications and the rationale behind these three levels of performance are explained. A description of a test motor that has been constructed to verify the electromagnetic analytical techniques of HTSC motor design is given. Finally, a DC motor with an HTSC field coil is described. Measured data with the motor running is presented showing that the motor is operating with the field winding in the superconducting state

Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

Energy Technology Data Exchange (ETDEWEB)

S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

2008-04-01

A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

International Nuclear Information System (INIS)

S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

2008-01-01

A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: (1) Identifies pre-conceptual design requirements; (2) Develops test loop equipment schematics and layout; (3) Identifies space allocations for each of the facility functions, as required; (4) Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems; (5) Identifies pre-conceptual utility and support system needs; and (6) Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs

Implication of irradiation effects on materials data for the design of near core components

International Nuclear Information System (INIS)

Dietz, W.; Breitling, H.

1995-01-01

For LWR's strict regulations exist for the consideration of irradiation in the design and surveillance of the reactor pressure vessel in the various codes (ASME, RCC-M, KTA) but less for near core components. For FBR's no firm rules exist either for the vessel nor the reactor internals. In this paper the German design practices for the loop type SNR-300 will be presented, and also some information from the surveillance programme of the KNK-reactor. Austenitic stainless steels have been mainly selected for the near core components. For some special applications Ni-alloys and a stabilized 2 1/4 Cr 1 Mo-alloy were specified. Considerations of the irradiation effects on material properties will be made for the various temperature and fluence levels around the core. The surveillance programmes will be described. Both, the consideration of irradiation effects in the elastic and inelastic analysis and the surveillance programmes had been a part of the licensing process for SNR-300. (author). 8 figs, 4 tabs

Geometric component of charge pumping current in nMOSFETs due to low-temperature irradiation

Science.gov (United States)

Witczak, S. C.; King, E. E.; Saks, N. S.; Lacoe, R. C.; Shaneyfelt, M. R.; Hash, G. L.; Hjalmarson, H. P.; Mayer, D. C.

2002-12-01

The geometric component of charge pumping current was examined in n-channel metal-oxide-silicon field effect transistors (MOSFETs) following low-temperature irradiation. In addition to the usual dependencies on channel length and gate bias transition time, the geometric component was found to increase with radiation-induced oxide-trapped charge density and decreasing temperature. A postirradiation injection of electrons into the gate oxide reduces the geometric component along with the density of oxide-trapped charge, which clearly demonstrates that the two are correlated. A fit of the injection data to a first-order model for trapping kinetics indicates that the electron trapping occurs predominantly at a single type of Coulomb-attractive trap site. The geometric component results primarily from the bulk recombination of channel electrons that fail to transport to the source or drain during the transition from inversion to accumulation. The radiation response of these transistors suggests that Coulomb scattering by oxide-trapped charge increases the bulk recombination at low temperatures by impeding electron transport. These results imply that the geometric component must be properly accounted for when charge pumping irradiated n-channel MOSFETs at low temperatures.

A Hybrid Hardware and Software Component Architecture for Embedded System Design

Science.gov (United States)

Marcondes, Hugo; Fröhlich, Antônio Augusto

Embedded systems are increasing in complexity, while several metrics such as time-to-market, reliability, safety and performance should be considered during the design of such systems. A component-based design which enables the migration of its components between hardware and software can cope to achieve such metrics. To enable that, we define hybrid hardware and software components as a development artifact that can be deployed by different combinations of hardware and software elements. In this paper, we present an architecture for developing such components in order to construct a repository of components that can migrate between the hardware and software domains to meet the design system requirements.

Plasma facing components design of KT-2 tokamak

International Nuclear Information System (INIS)

In, Sang Ryul; Yoon, Byung Joo; Song, Woo Soeb; Xu, Chao Yin

1997-04-01

The vacuum vessel of KT-2 tokamak is protected from high thermal loads by various kinds of plasma facing components (PFC): outer and inner divertors, neutral baffle, inboard limiter, poloidal limiter, movable limiter and passive plate, installed on the inner wall of the vessel. In this report the pre-engineering design of the plasma facing components, including design requirements and function, structures of PFC assemblies, configuration of cooling systems, calculations of some mechanical and hydraulic parameters, is presented. Pumping systems for the movable limiter and the divertor are also discussed briefly. (author). 49 figs

Effect of Feed Melting, Temperature History and Minor Component Addition on Spinel Crystallization in High-Level Waste Glass

International Nuclear Information System (INIS)

Izak, Pavel; Hrma, Pavel R.; Arey, Bruce W.; Plaisted, Trevor J.

2001-01-01

This study was undertaken to help design mathematical models for high-level waste (HLW) glass melter that simulate spinel behavior in molten glass. Spinel, (Fe,Ni,Mn) (Fe,Cr)2O4, is the primary solid phase that precipitates from HLW glasses containing Fe and Ni in sufficient concentrations. Spinel crystallization affects the anticipated cost and risk of HLW vitrification. To study melting reactions, we used simulated HLW feed, prepared with co-precipitated Fe, Ni, Cr, and Mn hydroxides. Feed samples were heated up at a temperature-increase rate (4C/min) close to that which the feed experiences in the HLW glass melter. The decomposition, melting, and dissolution of feed components (such as nitrates, carbonates, and silica) and the formation of intermediate crystalline phases (spinel, sodalite (Na8(AlSiO4)6(NO2)2), and Zr-containing minerals) were characterized using evolved gas analysis, volume-expansion measurement, optical microscope, scanning electron microscope, thermogravimetric analysis, differential scanning calorimetry, and X-ray diffraction. Nitrates and quartz, the major feed components, converted to a glass-forming melt by 880C. A chromium-free spinel formed in the nitrate melt starting from 520C and Sodalite, a transient product of corundum dissolution, appeared above 600C and eventually dissolved in glass. To investigate the effects of temperature history and minor components (Ru,Ag, and Cu) on the dissolution and growth of spinel crystals, samples were heated up to temperatures above liquidus temperature (TL), then subjected to different temperature histories, and analyzed. The results show that spinel mass fraction, crystals composition, and crystal size depend on the chemical and physical makeup of the feed and temperature history

FFTF Heat Transport System (HTS) component and system design

International Nuclear Information System (INIS)

Young, M.W.; Edwards, P.A.

1980-01-01

The FFTF Heat Transport Systems and Components designs have been completed and successfully tested at isothermal conditions up to 427 0 C (800 0 F). General performance has been as predicted in the design analyses. Operational flexibility and reliability have been outstanding throughout the test program. The components and systems have been demonstrated ready to support reactor powered operation testing planned later in 1980

Feature-based component model for design of embedded systems

Science.gov (United States)

Zha, Xuan Fang; Sriram, Ram D.

2004-11-01

An embedded system is a hybrid of hardware and software, which combines software's flexibility and hardware real-time performance. Embedded systems can be considered as assemblies of hardware and software components. An Open Embedded System Model (OESM) is currently being developed at NIST to provide a standard representation and exchange protocol for embedded systems and system-level design, simulation, and testing information. This paper proposes an approach to representing an embedded system feature-based model in OESM, i.e., Open Embedded System Feature Model (OESFM), addressing models of embedded system artifacts, embedded system components, embedded system features, and embedded system configuration/assembly. The approach provides an object-oriented UML (Unified Modeling Language) representation for the embedded system feature model and defines an extension to the NIST Core Product Model. The model provides a feature-based component framework allowing the designer to develop a virtual embedded system prototype through assembling virtual components. The framework not only provides a formal precise model of the embedded system prototype but also offers the possibility of designing variation of prototypes whose members are derived by changing certain virtual components with different features. A case study example is discussed to illustrate the embedded system model.

Thermal-hydraulic and thermo-mechanical design of plasma facing components for SST-1 tokamak

International Nuclear Information System (INIS)

Chaudhuri, Paritosh; Santra, P.; Chenna Reddy, D.; Parashar, S.K.S.

2014-01-01

The Plasma Facing Components (PFCs) are one of the major sub-systems of ssT-1 tokamak. PFC of ssT-1 consisting of divertors, passive stabilizers, baffles and limiters are designed to be compatible for steady state operation. The main consideration in the design of the PFC cooling is the steady state heat removal of up to 1 MW/m 2 . The PFC has been designed to withstand the peak heat fluxes and also without significant erosion such that frequent replacement of the armor is not necessary. Design considerations included 2-D steady state and transient tile temperature distribution and resulting thermal loads in PFC during baking, and cooling, coolant parameters necessary to maintain optimum thermal-hydraulic design, and tile fitting mechanism. Finite Element (FE) models using ANSYS have been developed to carry out the heat transfer and stress analyses of the PFC to understand its thermal and mechanical behaviors. The results of the calculation led to a good understanding of the coolant flow behavior and the temperature distribution in the tube wall and the different parts of the PFC. Thermal analysis of the PFC is carried out with the purpose of evaluating the thermal mechanical behavior of PFCs. The detailed thermal-hydraulic and thermo-mechanical designs of PFCs of ssT-1 are discussed in this paper. (authors)

Design methods for high temperature power plant structures

International Nuclear Information System (INIS)

Townley, C.H.A.

1984-01-01

The subject is discussed under the headings: introduction (scope of paper - reviews of design methods and design criteria currently in use for both nuclear and fossil fuelled power plant; examples chosen are (a) BS 1113, representative of design codes employed for power station boiler plant; (b) ASME Code Case N47, which is being developed for high temperature nuclear reactors, especially the liquid metal fast breeder reactor); design codes for power station boilers; Code Case N47 (design in the absence of thermal shock and thermal fatigue; design against cyclic loading at high temperature; further research in support of high temperature design methods and criteria for LMFBRs); concluding remarks. (U.K.)

Design Environment for Multifidelity and Multidisciplinary Components

Science.gov (United States)

Platt, Michael

2014-01-01

One of the greatest challenges when developing propulsion systems is predicting the interacting effects between the fluid loads, thermal loads, and structural deflection. The interactions between technical disciplines often are not fully analyzed, and the analysis in one discipline often uses a simplified representation of other disciplines as an input or boundary condition. For example, the fluid forces in an engine generate static and dynamic rotor deflection, but the forces themselves are dependent on the rotor position and its orbit. It is important to consider the interaction between the physical phenomena where the outcome of each analysis is heavily dependent on the inputs (e.g., changes in flow due to deflection, changes in deflection due to fluid forces). A rigid design process also lacks the flexibility to employ multiple levels of fidelity in the analysis of each of the components. This project developed and validated an innovative design environment that has the flexibility to simultaneously analyze multiple disciplines and multiple components with multiple levels of model fidelity. Using NASA's open-source multidisciplinary design analysis and optimization (OpenMDAO) framework, this multifaceted system will provide substantially superior capabilities to current design tools.

Design guidance for fracture-critical components at Lawrence Livermore National Laboratory

International Nuclear Information System (INIS)

Streit, R.D.

1982-01-01

Fracture is an important design consideration for components whose sudden and catastrophic failure could result in a serious accident. Elements of fracture control and fracture mechanics design methods are reviewed. Design requirements, which are based on the consequences of fracture of a given component, are subsequently developed. Five categories of consequences are defined. Category I is the lowest risk, and relatively lenient design requirements are employed. Category V has the highest potential for injury, release of hazardous material, and damage. Correspondingly, the design requirements for these components are the most stringent. Environmental, loading, and material factors that can affect fracture safety are also discussed

Design of Annular Linear Induction Pump for High Temperature Liquid Lead Transportation

Energy Technology Data Exchange (ETDEWEB)

Kwak, Jae Sik; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2014-05-15

EM(Electro Magnetic) Pump is divided into two parts, which consisted of the primary one with electromagnetic core and exciting coils, and secondary one with liquid lead flow. The main geometrical variables of the pump included core length, inner diameter and flow gap while the electromagnetic ones covered pole pitch, turns of coil, number of pole pairs, input current and input frequency. The characteristics of design variables are analyzed by electrical equivalent circuit method taking into account hydraulic head loss in the narrow annular channel of the ALIP. The design program, which was composed by using MATLAB language, was developed to draw pump design variables according to input requirements of the flow rate, developing pressure and operation temperature from the analyses. The analysis on the design of ALIP for high temperature liquid lead transportation was carried for the produce of ALIP designing program based on MATLAB. By the using of ALIP designing program, we don't have to bother about geometrical relationship between each component during detail designing process because code calculate automatically. And prediction of outputs about designing pump can be done easily before manufacturing. By running the code, we also observe and analysis change of outputs caused by changing of pump factors. It will be helpful for the research about optimization of pump outputs.

High Temperature Corrosion Problem of Boiler Components in presence of Sulfur and Alkali based Fuels

Science.gov (United States)

Ghosh, Debashis; Mitra, Swapan Kumar

2011-04-01

Material degradation and ageing is of particular concern for fossil fuel fired power plant components. New techniques/approaches have been explored in recent years for Residual Life assessment of aged components and material degradation due to different damage mechanism like creep, fatigue, corrosion and erosion etc. Apart from the creep, the high temperature corrosion problem in a fossil fuel fired boiler is a matter of great concern if the fuel contains sulfur, chlorine sodium, potassium and vanadium etc. This paper discusses the material degradation due to high temperature corrosion in different critical components of boiler like water wall, superheater and reheater tubes and also remedial measures to avoid the premature failure. This paper also high lights the Residual Life Assessment (RLA) methodology of the components based on high temperature fireside corrosion. of different critical components of boiler.

Nuclear component design ontology building based on ASME codes

International Nuclear Information System (INIS)

Bao Shiyi; Zhou Yu; He Shuyan

2005-01-01

The adoption of ontology analysis in the study of concept knowledge acquisition and representation for the nuclear component design process based on computer-supported cooperative work (CSCW) makes it possible to share and reuse numerous concept knowledge of multi-disciplinary domains. A practical ontology building method is accordingly proposed based on Protege knowledge model in combination with both top-down and bottom-up approaches together with Formal Concept Analysis (FCA). FCA exhibits its advantages in the way it helps establish and improve taxonomic hierarchy of concepts and resolve concept conflict occurred in modeling multi-disciplinary domains. With Protege-3.0 as the ontology building tool, a nuclear component design ontology based ASME codes is developed by utilizing the ontology building method. The ontology serves as the basis to realize concept knowledge sharing and reusing of nuclear component design. (authors)

Challenges in design of zirconium alloy reactor components

International Nuclear Information System (INIS)

Kakodkar, Anil; Sinha, R.K.

1992-01-01

Zirconium alloy components used in core-internal assemblies of heavy water reactors have to be designed under constraints imposed by need to have minimum mass, limitations of fabrication, welding and joining techniques with this material, and unique mechanisms for degradation of the operating performance of these components. These constraints manifest as challenges for design and development when the size, shape and dimensions of the components and assemblies are unconventional or untried, or when one is aiming for maximization of service life of these components under severe operating conditions. A number of such challenges were successfully met during the development of core-internal components and assemblies of Dhruva reactor. Some of the then untried ideas which were developed and successfully implemented include use of electron beam welding, cold forming of hemispherical ends of reentrant cans, and a large variety of rolled joints of innovative designs. This experience provided the foundation for taking up and successfully completing several tasks relating to coolant channels, liquid poison channels and sparger channels for PHWRs and test sections for the in-pile loops of Dhruva reactor. For life prediction and safety assessment of coolant channels of PHWRs some analytical tools, notably, a computer code for prediction of creep limited life of coolant channels has been developed. Some of the future challenges include the development of easily replaceable coolant channels and also large diameter coolant channels for Advanced Heavy Water Reactor, and development of solutions to overcome deterioration of service life of coolant channels due to hydriding. (author). 5 refs., 13 figs., 1 tab

Design and construction of γ-rays irradiation facility for remote-handling parts and components of fusion reactor

International Nuclear Information System (INIS)

Yagi, Toshiaki; Morita, Yousuke; Seguchi, Tadao

1995-03-01

For the evaluation of radiation resistance of remote-handling system for International Thermonuclear Experimental Reactor(ITER), 'high dose-rate and high temperature (upper 350degC) γ-rays irradiation facility' was designed and constructed. In this facility, the parts and components of remote-handling system such as sensing devices, motors, optical glasses, wires and cables, etc., are tested by irradiation with 2x10 6 Roentgen/h Co-60 γ-rays at a temperature up to 350degC under various atmospheres (dry nitrogen gas, argon gas, dry air and vacuum). (author)

Context sensitivity and ambiguity in component-based systems design

Energy Technology Data Exchange (ETDEWEB)

Bespalko, S.J.; Sindt, A.

1997-10-01

Designers of components-based, real-time systems need to guarantee to correctness of soft-ware and its output. Complexity of a system, and thus the propensity for error, is best characterized by the number of states a component can encounter. In many cases, large numbers of states arise where the processing is highly dependent on context. In these cases, states are often missed, leading to errors. The following are proposals for compactly specifying system states which allow the factoring of complex components into a control module and a semantic processing module. Further, the need for methods that allow for the explicit representation of ambiguity and uncertainty in the design of components is discussed. Presented herein are examples of real-world problems which are highly context-sensitive or are inherently ambiguous.

Conceptual designs for advanced, high-temperature CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Bushby, S.J. [Atomic Energy of Canada Ltd., Corrosion and Surface Science Branch, Chalk River Laboratories, Chalk River, ON (Canada); Dimmick, G.R. [Atomic Energy of Canada Ltd., Fuel Channel Thermmalhydraulics Branch, Chalk River, ON (Canada); Duffey, R.B. [Atomic Energy of Canada Ltd., Principal Scientist, Chalk River Laboratories, Chalk River, On (Canada); Spinks, N.J. [Atomic Energy of Canada Ltd., Researcher Emeritus, Chalk River Laboratories, Chalk River, ON (Canada); Burrill, K.A. [Atomic Energy of Canada Ltd., Chalk River Laboratories, Chalk River, ON (Canada); Chan, P.S.W. [Atomic Energy of Canada Ltd., Reactor Core Physics Branch, Mississauga, ON (Canada)

2000-07-01

AECL is studying advanced reactor concepts with the aim of significant cost reduction through improved thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, also incorporates enhanced safety features, and flexible, proliferation-resistant fuel cycles, whilst retaining the fundamental design characteristics of CANDU: neutron economy, horizontal fuel channels, and a separate D{sub 2}O moderator that provides a passive heat sink. Where possible, proven, existing components and materials would be adopted, so that 'first-of-a-kind' costs and uncertainties are avoided. Three reactor concepts ranging in output from {approx}375 MW(e) to 1150 MW(e) are described. The modular design of a pressure tube reactor allows the plant size for each concept to be tailored to a given market through the addition or removal of fuel channels. Each concept uses supercritical water as the coolant at a nominal pressure of 25 MPa. Core outlet temperatures range from {approx}400degC to 625degC, resulting in substantial improvements in thermodynamic efficiencies compared to current nuclear stations. The CANDU-X Mark 1 concept is an extension of the present CANDU design. An indirect cycle is employed, but efficiency is increased due to higher coolant temperature, and changes to the secondary side; as well, the size and number of pumps and steam generators are reduced. Safety is enhanced through facilitation of thermo-siphoning of decay heat by increasing the temperature of the moderator. The CANDU-X NC concept is also based on an indirect cycle, but natural convection is used to circulate the primary coolant. This approach enhances cycle efficiency and safety, and is viable for reactors operating near the pseudo-critical temperature of water because of large changes in heat capacity and thermal expansion in that region. In the third concept (CANDUal-X), a dual cycle is employed. Supercritical water exits the core and feeds directly into a very high

High temperature fusion reactor design

International Nuclear Information System (INIS)

Harkness, S.D.; dePaz, J.F.; Gohar, M.Y.; Stevens, H.C.

1979-01-01

Fusion energy may have unique advantages over other systems as a source for high temperature process heat. A conceptual design of a blanket for a 7 m tokamak reactor has been developed that is capable of producing 1100 0 C process heat at a pressure of approximately 10 atmospheres. The design is based on the use of a falling bed of MgO spheres as the high temperature heat transfer system. By preheating the spheres with energy taken from the low temperature tritium breeding part of the blanket, 1086 MW of energy can be generated at 1100 0 C from a system that produces 3000 MW of total energy while sustaining a tritium breeding ratio of 1.07. The tritium breeding is accomplished using Li 2 O modules both in front of (6 cm thick) and behind (50 cm thick) the high temperature ducts. Steam is used as the first wall and front tritium breeding module coolant while helium is used in the rear tritium breeding region. The system produces 600 MW of net electricity for use on the grid

Enhanced Design Alternative I: Low Temperature Design

International Nuclear Information System (INIS)

MacNeil, K.

1999-01-01

The purpose of this document is to evaluate Enhanced Design Alternative (EDA) 1, the low temperature repository design concept (CRWMS M and O 1999a). This technical document will provide supporting information for Site Recommendation (SR) and License Application (LA). Preparation of this evaluation will be in accordance with the technical document preparation plan (TDPP), (CRWMS M and O 1999b). EDA 1, one of five EDAs, was evolved from evaluation of a series of design features and alternatives developed during the first phase of the License Application Design Selection (LADS) process. Low, medium, and high temperature concepts were developed from the design features and alternatives prepared during Phase 1 of the LADS effort (CRWMS M and O 1999a). EDA 1 will first be evaluated against a single Screening Criterion, outlined in CRWMS M and O 1999a, which addresses post-closure performance of the repository. The performance of the repository is defined quantitatively as the peak radiological dose rate to an average individual of a critical group at a distance of 20 km from the repository site within 10,000 years. To satisfy this criterion the peak dose rate must not exceed the anticipated regulatory level of 25 mrem/yr within 10,000 years. If the EDA meets the screening criterion, the EDA will be further evaluated against the LADS Phase 2 Evaluation Criteria contained in CRWMS M and O 1999a

The materials programme for the high-temperature gas-cooled reactor in the Federal Republic of Germany: Status of the development of high-temperature materials, integrity concept, and design codes

International Nuclear Information System (INIS)

Nickel, H.; Bodmann, E.; Seehafer, H.J.

1990-01-01

During the last 15 years, the research and development of materials for high temperature gas-cooled reactor (HTGR) applications in the Federal Republic of Germany have been concentrated on the qualification of high-temperature structural alloys. Such materials are required for heat exchanger components of advanced HTGRs supplying nuclear process heat in the temperature range between 750 deg. and 950 deg. C. The suitability of the candidate alloys for service in the HTGR has been established, and continuing research is aimed at verification of the integrity of components over the envisaged service lifetimes. The special features of the HTGR which provide a high degree of safety are the use of ceramics for the core construction and the low power density of the core. The reactor integrity concept which has been developed is based on these two characteristics. Previously, technical guidelines and design codes for nuclear plants were tailored exclusively to light water reactor systems. An extensive research project was therefore initiated which led to the formulation of the basic principles on which a high temperature design code can be based. (author)

Design and development of major balance of plant components in solid oxide fuel cell system

Energy Technology Data Exchange (ETDEWEB)

Hong, Wen-Tang; Huang, Cheng-Nan; Tan, Hsueh-I; Chao, Yu [Institute of Nuclear Energy Research Atomic Energy Council, Taoyuan County 32546 (Taiwan, Province of China); Yen, Tzu-Hsiang [Green Technology Research Institute, CPC Corporation, Chia-Yi City 60036 (Taiwan, Province of China)

2013-07-01

The balance of plant (BOP) of a Solid Oxide Fuel Cell (SOFC) system with a 2 kW stack and an electric efficiency of 40% is optimized using commercial GCTool software. The simulation results provide a detailed understanding of the optimal operating temperature, pressure and mass flow rate in all of the major BOP components, i.e., the gas distributor, the afterburner, the reformer and the heat exchanger. A series of experimental trials are performed to validate the simulation results. Overall, the results presented in this study not only indicate an appropriate set of operating conditions for the SOFC power system, but also suggest potential design improvements for several of the BOP components.

Thermomechanical fatigue life prediction of high temperature components

Energy Technology Data Exchange (ETDEWEB)

Seifert, Thomas; Hartrott, Philipp von; Riedel, Hermann; Siegele, Dieter [Fraunhofer-Inst. fuer Werkstoffmechanik (IWM), Freiburg (Germany)

2009-07-01

The aim of the work described in this paper is to provide a computational method for fatigue life prediction of high temperature components, in which the time and temperature dependent fatigue crack growth is a relevant damage mechanism. The fatigue life prediction is based on a law for microcrack growth and a fracture mechanics estimate of the cyclic crack tip opening displacement. In addition, a powerful model for nonisothermal cyclic plasticity is employed, and an efficient laboratory test procedure is proposed for the determination of the model parameters. The models are efficiently implemented into finite element programs and are used to predict the fatigue life of a cast iron exhaust manifold and a notch in the perimeter of a turbine rotor made of a ferritic/martensitic 10%-chromium steel. (orig.)

Specification of properties and design allowables for copper alloys used in HHF components of ITER

DEFF Research Database (Denmark)

Kalinin, G.M.; Fabritziev, S.A.; Singh, B.N.

2002-01-01

CrZr and CuAl25 are not yet fully characterised. The performed R&D gives a basis for the specification of physical and mechanical properties required for the design analysis in accordance with the ITER Structural Design Criteria for In-vessel Components (SDC-IC). For both CuCrZr-IG and CuAl25-IG alloys......Two types of copper alloys, precipitation hardened (PH) Cu (CuCrZr-IG) and dispersion strengthened (DS) Cu (CuAl25-IG), are proposed as heat sink materials for the high heat flux (HHF) components of ITER. However, copper alloys are not included in any national codes, and properties of both Cu......, the statistical evaluation of available experimental data has been used to calculate the temperature dependence of the average value and of the 95% confidence limit of tensile properties. The stress limits, Sm, Se, and Sd, have been estimated on the basis of available data. The procedure used for specification...

Design and safety consideration in the High-Temperature Engineering Test Reactor (HTTR)

International Nuclear Information System (INIS)

Saito, Shinzo; Tanaka, Toshiuki; Sudo, Yukio; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru

1990-01-01

The budget for construction of the High-Temperature Engineering Test Reactor (HTTR) was recently committed by the Government in Japan. The HTTR is a test reactor with thermal output of 30 MW and reactor outlet coolant temperature of 950 deg. C at high temperature test operation. The HTTR plant uses a pin-in-block design core and will be used as an experience leading to high temperature applications. Several major important safety considerations are adopted in the design of the HTTR. These are as follows: 1) A coated particle fuel must not be failed during a normal reactor operation and an anticipated operational occurrence; 2) Two independent and diverse reactor shut-down systems are provided in order to shut down the reactor safely and reliably in any condition; 3) Back-up reactor cooling systems which are safety ones are provided in order to remove residual heat of reactor in any condition; 4) Multiple barriers and countermeasures are provided to contain fission products such as a containment, pressure gradient between the primary and secondary cooling circuit and so on, though coated particle fuels contain fission products with high reliability; 5) The functions of materials used in the primary cooling circuit are separated to be pressure-resisting and heat-resisting in order to resolve material problems and maintain high reliability. The detailed design of the HTTR was completed with extensive accumulation of material data and component tests. (author)

COMPUTER AIDED THREE DIMENSIONAL DESIGN OF MOLD COMPONENTS

Directory of Open Access Journals (Sweden)

Kerim ÇETİNKAYA

2000-02-01

Full Text Available Sheet metal molding design with classical methods is formed in very long times calculates and drafts. At the molding design, selection and drafting of most of the components requires very long time because of similar repetative processes. In this study, a molding design program has been developed by using AutoLISP which has been adapted AutoCAD packet program. With this study, design of sheet metal molding, dimensioning, assemly drafting has been realized.

Components of the primary circuit of LWRs. Design, construction and calculation. Komponenten des Primaerkreises von Leichtwasserreaktoren. Auslegung, Konstruktion und Berechnung

Energy Technology Data Exchange (ETDEWEB)

1980-01-01

This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673 K (400 C). The primary circuit as the pressure containment of the reactor coolant comprises: Reactor pressure vessel (without internals), steam generator (primary loop), pressurizer, reactor coolant pump housing, interconnecting pipings between the components mentioned above and appropriate various valve and instrument casings, pipings branding from the above components and interconnecting pipings, including the appropriate instrument casings, up to and including the first isolating valve, pressure shielding of control rod drives.

Conceptual design of heat transport systems and components of PFBR-NSSS

International Nuclear Information System (INIS)

Chetal, S.C.; Bhoje, S.B.; Kale, R.D.; Rao, A.S.L.K.; Mitra, T.K.; Selvaraj, A.; Sethi, V.K.; Sundaramoorthy, T.R.; Balasubramaniyan, V.; Vaidyanathan, G.

1996-01-01

The production of electrical power from sodium cooled fast reactors in the present power scenario in India demands emphasis on plant economics consistent with safety. Number of heat transport systems/components and the design of principal heat transport components viz sodium pumps, IHX and steam generators play significant role in the plant capital cost and capacity factor. The paper discusses the basis of selection of 2 primary pumps, 4 IHX, 2 secondary loops, 2 secondary pumps and 8 steam generators for the 500 MWe Prototype Fast Breeder Reactor (PFBR), which is now in design stage. The principal design features of primary pump, IHX and steam generator have been selected based on design simplicity, ease of manufacture and utilization of established designs. The paper also describes the conceptual design of above mentioned three components. (author). 3 figs, 2 tabs

High temperature brazing of primary-system components in the nuclear field

International Nuclear Information System (INIS)

Belicic, M.; Fricker, H.W.; Iversen, K.; Leukert, W.

1981-01-01

Apart from the well-known welding procedures, high-temperature brazing is successfully applied in the manufacture of primary components in the field of nuclear reactor construction. This technique is applied in all cases where apart from sufficient resistance and high production safety importance is laid on dimensional stability without subsequent mechanical processing of the components. High-temperature brazing is therefore very important in the manufacture of fuel rod spacers or control rod guide tubes. In this context, during one brazing process many brazing seams have to be produced in extremely narrow areas and within small tolerances. As basic materials precipitation hardening alloys with a high nickel percentage, austenitic Cr-Ni-steels or the zirconium alloy Zry 4 are used. Generally applied are: boron free nickel or zirconium brazing filler metals. (orig.)

Opto-mechanical subsystem with temperature compensation through isothemal design

Science.gov (United States)

Goodwin, F. E. (Inventor)

1977-01-01

An opto-mechanical subsystem for supporting a laser structure which minimizes changes in the alignment of the laser optics in response to temperature variations is described. Both optical and mechanical structural components of the system are formed of the same material, preferably beryllium, which is selected for high mechanical strength and good thermal conducting qualities. All mechanical and optical components are mounted and assembled to provide thorough thermal coupling throughout the subsystem to prevent the development of temperature gradients.

Molecular Orientation in Two Component Vapor-Deposited Glasses: Effect of Substrate Temperature and Molecular Shape

Science.gov (United States)

Powell, Charles; Jiang, Jing; Walters, Diane; Ediger, Mark

Vapor-deposited glasses are widely investigated for use in organic electronics including the emitting layers of OLED devices. These materials, while macroscopically homogenous, have anisotropic packing and molecular orientation. By controlling this orientation, outcoupling efficiency can be increased by aligning the transition dipole moment of the light-emitting molecules parallel to the substrate. Light-emitting molecules are typically dispersed in a host matrix, as such, it is imperative to understand molecular orientation in two-component systems. In this study we examine two-component vapor-deposited films and the orientations of the constituent molecules using spectroscopic ellipsometry, UV-vis and IR spectroscopy. The role of temperature, composition and molecular shape as it effects molecular orientation is examined for mixtures of DSA-Ph in Alq3 and in TPD. Deposition temperature relative to the glass transition temperature of the two-component mixture is the primary controlling factor for molecular orientation. In mixtures of DSA-Ph in Alq3, the linear DSA-Ph has a horizontal orientation at low temperatures and slight vertical orientation maximized at 0.96Tg,mixture, analogous to one-component films.

Monthly version of HadISST sea surface temperature state-space components

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — State-Space Decomposition of Monthly version of HadISST sea surface temperature component (1-degree). See Rayner, N. A., Parker, D. E., Horton, E. B., Folland, C....

Additive Manufacturing Design Considerations for Liquid Engine Components

Science.gov (United States)

Whitten, Dave; Hissam, Andy; Baker, Kevin; Rice, Darron

2014-01-01

The Marshall Space Flight Center's Propulsion Systems Department has gained significant experience in the last year designing, building, and testing liquid engine components using additive manufacturing. The department has developed valve, duct, turbo-machinery, and combustion device components using this technology. Many valuable lessons were learned during this process. These lessons will be the focus of this presentation. We will present criteria for selecting part candidates for additive manufacturing. Some part characteristics are 'tailor made' for this process. Selecting the right parts for the process is the first step to maximizing productivity gains. We will also present specific lessons we learned about feature geometry that can and cannot be produced using additive manufacturing machines. Most liquid engine components were made using a two-step process. The base part was made using additive manufacturing and then traditional machining processes were used to produce the final part. The presentation will describe design accommodations needed to make the base part and lessons we learned about which features could be built directly and which require the final machine process. Tolerance capabilities, surface finish, and material thickness allowances will also be covered. Additive Manufacturing can produce internal passages that cannot be made using traditional approaches. It can also eliminate a significant amount of manpower by reducing part count and leveraging model-based design and analysis techniques. Information will be shared about performance enhancements and design efficiencies we experienced for certain categories of engine parts.

Design and Measurement of Metallic Post-Wall Waveguide Components

NARCIS (Netherlands)

Coenen, T.J.; Bekers, D.J.; Tauritz, J.L.; Vliet, F.E. van

2009-01-01

Abstract—In this paper we discuss the design and measurement of a set of metallic post-wall waveguide components for antenna feed structures. The components are manufactured on a single layer printed circuit board and excited by a grounded coplanar waveguide. For a straight transmission line, a 90°

High Temperature Transparent Furnace Development

Science.gov (United States)

Bates, Stephen C.

1997-01-01

This report describes the use of novel techniques for heat containment that could be used to build a high temperature transparent furnace. The primary objective of the work was to experimentally demonstrate transparent furnace operation at 1200 C. Secondary objectives were to understand furnace operation and furnace component specification to enable the design and construction of a low power prototype furnace for delivery to NASA in a follow-up project. The basic approach of the research was to couple high temperature component design with simple concept demonstration experiments that modify a commercially available transparent furnace rated at lower temperature. A detailed energy balance of the operating transparent furnace was performed, calculating heat losses through the furnace components as a result of conduction, radiation, and convection. The transparent furnace shells and furnace components were redesigned to permit furnace operation at at least 1200 C. Techniques were developed that are expected to lead to significantly improved heat containment compared with current transparent furnaces. The design of a thermal profile in a multizone high temperature transparent furnace design was also addressed. Experiments were performed to verify the energy balance analysis, to demonstrate some of the major furnace improvement techniques developed, and to demonstrate the overall feasibility of a high temperature transparent furnace. The important objective of the research was achieved: to demonstrate the feasibility of operating a transparent furnace at 1200 C.

Neurocognitive and somatic components of temperature increases during g-tummo meditation: legend and reality.

Directory of Open Access Journals (Sweden)

Maria Kozhevnikov

Full Text Available Stories of g-tummo meditators mysteriously able to dry wet sheets wrapped around their naked bodies during a frigid Himalayan ceremony have intrigued scholars and laypersons alike for a century. Study 1 was conducted in remote monasteries of eastern Tibet with expert meditators performing g-tummo practices while their axillary temperature and electroencephalographic (EEG activity were measured. Study 2 was conducted with Western participants (a non-meditator control group instructed to use the somatic component of the g-tummo practice (vase breathing without utilization of meditative visualization. Reliable increases in axillary temperature from normal to slight or moderate fever zone (up to 38.3°C were observed among meditators only during the Forceful Breath type of g-tummo meditation accompanied by increases in alpha, beta, and gamma power. The magnitude of the temperature increases significantly correlated with the increases in alpha power during Forceful Breath meditation. The findings indicate that there are two factors affecting temperature increase. The first is the somatic component which causes thermogenesis, while the second is the neurocognitive component (meditative visualization that aids in sustaining temperature increases for longer periods. Without meditative visualization, both meditators and non-meditators were capable of using the Forceful Breath vase breathing only for a limited time, resulting in limited temperature increases in the range of normal body temperature. Overall, the results suggest that specific aspects of the g-tummo technique might help non-meditators learn how to regulate their body temperature, which has implications for improving health and regulating cognitive performance.

Development of design Criteria for ITER In-vessel Components

International Nuclear Information System (INIS)

Sannazzaro, G.; Barabash, V.; Kang, S.C.; Fernandez, E.; Kalinin, G.; Obushev, A.; Martínez, V.J.; Vázquez, I.; Fernández, F.; Guirao, J.

2013-01-01

Absrtract: The components located inside the ITER vacuum chamber (in-vessel components – IC), due to their specific nature and the environments they are exposed to (neutron radiation, high heat fluxes, electromagnetic forces, etc.), have specific design criteria which are, in this paper, referred as Structural Design Criteria for In-vessel Components (SDC-IC). The development of these criteria started in the very early phase of the ITER design and followed closely the criteria of the RCC-MR code. Specific rules to include the effect of neutron irradiation were implemented. In 2008 the need of an update of the SDC-IC was identified to add missing specifications, to implement improvements, to modernise rules including recent evolutions in international codes and regulations (i.e. PED). Collaboration was set up between ITER Organization (IO), European (EUDA) and Russian Federation (RFDA) Domestic Agencies to generate a new version of SDC-IC. A Peer Review Group (PRG) composed by members of the ITER Organization and all ITER Domestic Agencies and code experts was set-up to review the proposed modifications, to provide comments, contributions and recommendations

Lower-Temperature Invert Design For Diffusion Barrier

International Nuclear Information System (INIS)

Bruce Stanley

2001-01-01

The objective of this analysis is to advance the state of the subsurface facilities design to primarily support the ''Yucca Mountain Science and Engineering Report'' (DOE 2001) and to also support the preparation and revision of System Description Document's Section 2 system descriptions (CRWMS M and O 2001, pp. 9 and 11). The results may also eventually support the License Application (CRWMS M and O 2001, p. 3). The Performance Assessment Department will be the primary user of the information generated and will be used in abstraction modeling for the lower-temperature scenario (CRWMS M and O 200 1, p. 27). This analysis will evaluate the invert relative to the lower- and higher-temperature conditions in accordance with the primary tasks below. Invert design is a major factor in allowing water entering the drift to pass freely and enter the drift floor without surface ponding and in limiting diffusive transport into the host rock. Specific cost effective designs will be conceptualized under the new lower-temperature conditions in this analysis. Interfacing activities and all aspects of Integrated Safety Management and Nuclear Culture principles are included in this work scope by adhering to the respective principles during this design activity and by incorporating safety into the design analysis (CRWMS M and O 2001, p. 8). Primary tasks of this analysis include identifying available design information from existing sources on the invert as a diffusive barrier, developing concepts that reduce the amount steel, and developing other design features that accommodate both lower- and higher-temperature operating modes (CRWMS M and O 2001, p.16)

Optimization of Cycle and Expander Design of an Organic Rankine Cycle Unit using Multi-Component Working Fluids

DEFF Research Database (Denmark)

Meroni, Andrea; Andreasen, Jesper Graa; Pierobon, Leonardo

2016-01-01

Organic Rankine cycle (ORC) power systems represent at-tractive solutions for power conversion from low temperatureheat sources, and the use of these power systems is gaining increasing attention in the marine industry. This paper proposesthe combined optimal design of cycle and expander...... for an organic Rankine cycle unit utilizing waste heat from low temperature heat sources. The study addresses a case where the minimum temperature of the heat source is constrained and a case where no constraint is imposed. The former case is the wasteheat recovery from jacket cooling water of a marine diesel...... engine onboard a large ship, and the latter is representative of a low-temperature geothermal, solar or waste heat recovery application. Multi-component working fluids are investigated, as they allow improving the match between the temperature pro-files in the heat exchangers and, consequently, reducing...

Design of High Temperature Reactor Vessel Using ANSYS Software

International Nuclear Information System (INIS)

Bandriyana; Kasmudin

2003-01-01

Design calculation and evaluation of material strength for high temperature reactor vessel based on the design of HTR-10 high temperature reactor vessel were carried out by using the ANSYS 5.4 software. ANSYS software was applied to calculate the combined load from thermal and pressure load. Evaluation of material strength was performed by calculate and determine the distribution of temperature, stress and strain in the thickness direction of vessel, and compared with its material strength for designed. The calculation was based on the inner wall temperature of vessel of 600 o C and the outer temperature of 500 and 600 o C. Result of calculation gave the maximum stress for outer temperature of 600 o C was 288 N/ mm 2 and strain of 0.000187. For outer temperature of 500 o C the maximum stress was 576 N/ mm 2 and strain of 0.003. Based on the analysis result, the material of steel SA 516-70 with limited stress for design of 308 N/ mm 2 can be used for vessel material with outer wall temperature of 600 o C

Structural materials for ITER in-vessel component design

Energy Technology Data Exchange (ETDEWEB)

Kalinin, G. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Gauster, W. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Matera, R. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Tavassoli, A.-A.F. [CEA Centre d`Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France); Rowcliffe, A. [Oak Ridge National Lab., TN (United States); Fabritsiev, S. [Research Inst. of Electrophysical Apparatus, St. Petersburg (Russian Federation); Kawamura, H. [JAERI, IMTR Project, Ibaraki (Japan). Blanket Irradiation Lab.

1996-10-01

The materials proposed for ITER in-vessel components have to exhibit adequate performance for the operating lifetime of the reactor or for specified replacement intervals. Estimates show that maximum irradiation dose to be up to 5-7 dpa (for 1 MWa/m{sup 2} in the basic performance phase (BPP)) within a temperature range from 20 to 300 C. Austenitic SS 316LN-ITER Grade was defined as a reference option for the vacuum vessel, blanket, primary wall, pipe lines and divertor body. Conventional technologies and mill products are proposed for blanket, back plate and manifold manufacturing. HIPing is proposed as a reference manufacturing method for the primary wall and blanket and as an option for the divertor body. The existing data show that mechanical properties of HIPed SS are no worse than those of forged 316LN SS. Irradiation will result in property changes. Minimum ductility has been observed after irradiation in an approximate temperature range between 250 and 350 C, for doses of 5-10 dpa. In spite of radiation-induced changes in tensile deformation behavior, the fracture remains ductile. Irradiation assisted corrosion cracking is a concern for high doses of irradiation and at high temperatures. Re-welding is one of the critical issues because of the need to replace failed components. It is also being considered for the replacement of shielding blanket modules by breeding modules after the BPP. (orig.).

Design prospect of remountable high-temperature superconducting magnet

Energy Technology Data Exchange (ETDEWEB)

Hashizume, Hidetoshi, E-mail: hidetoshi.hashizume@qse.tohoku.ac.jp; Ito, Satoshi

2014-10-15

The remountable (mountable and demountable repeatedly) high-temperature superconducting (HTS) magnet has been proposed for huge and complex superconducting magnets in future fusion reactors to fabricate and repair easily the magnet and access inner structural components. This paper summarizes progress in R and D activities of mechanical joints of HTS conductors in terms of the electrical resistance and heat transfer performance at the joint region. The latest experimental results show the low joint resistance, 4 nΩ under 70 kA current condition using REBCO HTS conductor with mechanical lap joint system, and for the cooling system the maximum heat flux of 0.4 MW/m{sup 2} is removed by using bronze sintered porous media with sub-cooled liquid nitrogen. These values indicate that there is large possibility to design the remountable HTS magnet for fusion reactors.

Designing an accurate system for temperature measurements

Directory of Open Access Journals (Sweden)

Kochan Orest

2017-01-01

Full Text Available The method of compensation of changes in temperature field along the legs of inhomogeneous thermocouple, which measures a temperature of an object, is considered in this paper. This compensation is achieved by stabilization of the temperature field along the thermocouple. Such stabilization does not allow the error due to acquired thermoelectric inhomogeneity to manifest itself. There is also proposed the design of the furnace to stabilize temperature field along the legs of the thermocouple which measures the temperature of an object. This furnace is not integrated with the thermocouple mentioned above, therefore it is possible to replace this thermocouple with a new one when it get its legs considerably inhomogeneous.. There is designed the two loop measuring system with the ability of error correction which can use simultaneously a usual thermocouple as well as a thermocouple with controlled profile of temperature field. The latter can be used as a reference sensor for the former.

Thick-Film and LTCC Passive Components for High-Temperature Electronics

Directory of Open Access Journals (Sweden)

A. Dziedzic

2013-04-01

Full Text Available At this very moment an increasing interest in the field of high-temperature electronics is observed. This is a result of development in the area of wide-band semiconductors’ engineering but this also generates needs for passives with appropriate characteristics. This paper presents fabrication as well as electrical and stability properties of passive components (resistors, capacitors, inductors made in thick-film or Low-Temperature Co-fired Ceramics (LTCC technologies fulfilling demands of high-temperature electronics. Passives with standard dimensions usually are prepared by screen-printing whereas combination of standard screen-printing with photolithography or laser shaping are recommenced for fabrication of micropassives. Attainment of proper characteristics versus temperature as well as satisfactory long-term high-temperature stability of micropassives is more difficult than for structures with typical dimensions for thick-film and LTCC technologies because of increase of interfacial processes’ importance. However it is shown that proper selection of thick-film inks together with proper deposition method permit to prepare thick-film micropassives (microresistors, air-cored microinductors and interdigital microcapacitors suitable for the temperature range between 150°C and 400°C.

Methods for designing building envelope components prepared for repair and maintenance

DEFF Research Database (Denmark)

Rudbeck, Claus Christian

2000-01-01

the deterministic and probabilistic approach. Based on an investigation of the data-requirement, user-friendliness and supposed accuracy (the accuracy of the different methods has not been evaluated due to the absence of field data) the method which combines the deterministic factor method with statistical...... to be prepared for repair and maintenance. Both of these components are insulation systems for flat roofs and low slope roofs; components where repair or replacement is very expensive if the roofing material fails in its function. The principle of both roofing insulation systems is that the insulation can...... of issues which are specified below:Further development of methods for designing building envelope components prepared for repair and maintenance, and ways of tracking and predicting performance through time once the components have been designed, implemented in a building design and built...

Time-domain ultra-wideband radar, sensor and components theory, analysis and design

CERN Document Server

Nguyen, Cam

2014-01-01

This book presents the theory, analysis, and design of ultra-wideband (UWB) radar and sensor systems (in short, UWB systems) and their components. UWB systems find numerous applications in the military, security, civilian, commercial and medicine fields. This book addresses five main topics of UWB systems: System Analysis, Transmitter Design, Receiver Design, Antenna Design and System Integration and Test. The developments of a practical UWB system and its components using microwave integrated circuits, as well as various measurements, are included in detail to demonstrate the theory, analysis and design technique. Essentially, this book will enable the reader to design their own UWB systems and components. In the System Analysis chapter, the UWB principle of operation as well as the power budget analysis and range resolution analysis are presented. In the UWB Transmitter Design chapter, the design, fabrication and measurement of impulse and monocycle pulse generators are covered. The UWB Receiver Design cha...

Method of forming components for a high-temperature secondary electrochemical cell

Science.gov (United States)

Mrazek, Franklin C.; Battles, James E.

1983-01-01

A method of forming a component for a high-temperature secondary electrochemical cell having a positive electrode including a sulfide selected from the group consisting of iron sulfides, nickel sulfides, copper sulfides and cobalt sulfides, a negative electrode including an alloy of aluminum and an electrically insulating porous separator between said electrodes. The improvement comprises forming a slurry of solid particles dispersed in a liquid electrolyte such as the lithium chloride-potassium chloride eutetic, casting the slurry into a form having the shape of one of the components and smoothing the exposed surface of the slurry, cooling the cast slurry to form the solid component, and removing same. Electrodes and separators can be thus formed.

ITER plasma facing components, design and development

International Nuclear Information System (INIS)

Vieider, G.; Cardella, A.; Akiba, M.; Matera, R.; Watson, R.

1991-01-01

The paper summarizes the collaborative effort of the ITER Conceptual Design Activity (CDA) on Plasma Facing Components (PFC) which focused on the following main tasks: (a) The definition of basic design concepts for the First Wall (FW) and Divertor Plates (DP), (b) the analysis of the performance and likely lifetime of these PFC designs including the identification of major critical issues, (c) the start of R and D work giving already first results, and the definition of the required further R and D program to support the contemplated ITER Engineering Design Activity (EDA). From the ITER CDA effort on PFC it is mainly concluded that: (a) The expected PFC operating conditions lead to design solutions at the limit of present technology in particular for the divertor, which may constrain the overall machine performance, (b) the development of convincing PFC designs requires an intensified R and D effort both on PFC technology and plasma physics. (orig.)

Assessment of New Components to be integrated in the LHC Room Temperature Vacuum System

CERN Document Server

Bregliozzi, G; Chiggiato, P

2014-01-01

Integration of new equipment in the long straight sections (LSS) of the LHC must be compatible with the TiZrV non-evaporable getter thin film that coats most of the 6-km-long room-temperature beam pipes. This paper focus on two innovative accelerator devices to be installed in the LSS during the long shutdown 1 (LS1): the beam gas vertex (BGV) and a beam bending experiment using a crystal collimator (LUA9). The BGV necessitates a dedicated pressure bump, generated by local gas injection, in order to create the required rate of inelastic beam-gas interactions. The LAU9 experiments aims at improving beam cleaning efficiency with the use of a crystal collimator. New materials like fibre optics, piezoelectric components, and glues are proposed in the original design of the two devices. The integration feasibility of these set-ups in the LSS is presented. In particular outgassing tests of special components, X-rays photoelectron spectroscopy analysis of NEG coating behaviour in presence of glues during bake-out, a...

Enactment of KEPIC MNH Based on 2007 ASME BPVC Section III Division 1, Subsection NH: Class 1 Components in Elevated Temperature Service

International Nuclear Information System (INIS)

Koo, Gyeong Hoi; Kim, J. B.; Lee, H. Y.; Park, C. G.

2008-11-01

This report is a draft of an enactment of KEPIC MNH based on 2007 ASME Boiler and Pressure Vessel Code, Section III, Division 1 Subsection NH for Class 1 Components in Elevated Temperature Service and contains of ASME Article NH-3000 design, the mandatory appendix I-14, and non-mandatory appendices T and X

Enactment of KEPIC MNH Based on 2007 ASME BPVC Section III Division 1, Subsection NH: Class 1 Components in Elevated Temperature Service

Energy Technology Data Exchange (ETDEWEB)

Koo, Gyeong Hoi; Kim, J. B.; Lee, H. Y.; Park, C. G

2008-11-15

This report is a draft of an enactment of KEPIC MNH based on 2007 ASME Boiler and Pressure Vessel Code, Section III, Division 1 Subsection NH for Class 1 Components in Elevated Temperature Service and contains of ASME Article NH-3000 design, the mandatory appendix I-14, and non-mandatory appendices T and X.

Design and development of R.F. LINAC accelerator components

International Nuclear Information System (INIS)

Abhay Kumar; Guha, S.; Balasubramaniam, R.; Jawale, S.B.

2003-01-01

Full text: Radio frequency linear accelerator, a high power electron LINAC technology, is being developed at BARC. These accelerators are considered to be the most compact and effective for a given power capacity. Important application areas of this LINAC include medical sterilization, food preservation, pollution control, semiconductor industries, radiation therapy and material science. Center for Design and Manufacture (CDM), BARC has been entrusted with the design, development and manufacturing of various mechanical components of the accelerator. Most critical and precision components out of them are Diagnostic chamber, Faraday cup, Drift tube and R.F. cavities. This paper deals with the design aspects in respect of Ultra high vacuum compatibility and the mechanism of operation. Also this paper discusses the state-of-art technology for machining of intricate contour using specially designed poly crystalline diamond tool and the inspection methodology developed to minimize the measurement errors on the machined contour. Silver brazing technique employed to join the LINAC cavities is also described in detail

Mechanical design of core components for a high performance light water reactor with a three pass core

International Nuclear Information System (INIS)

Fischer, Kai; Schneider, Tobias; Redon, Thomas; Schulenberg, Thomas; Starflinger, Joerg

2007-01-01

Nuclear reactors using supercritical water as coolant can achieve more than 500 deg. C core outlet temperature, if the coolant is heated up in three steps with intermediate mixing to avoid hot streaks. This method reduces the peak cladding temperatures significantly compared with a single heat up. The paper presents an innovative mechanical design which has been developed recently for such a High Performance Light Water Reactor. The core is built with square assemblies of 40 fuel pins each, using wire wraps as grid spacers. Nine of these assemblies are combined to a cluster having a common head piece and a common foot piece. A downward flow of additional moderator water, separated from the coolant, is provided in gaps between the assemblies and in a water box inside each assembly. The cluster head and foot pieces and mixing chambers, which are key components for this design, are explained in detail. (authors)

Combination scattering of dissociating gas applied to measurements of temperature and concentration of components

International Nuclear Information System (INIS)

Pashkov, V.A.; Kurganova, F.I.; Grishchuk, M.Kh.

1987-01-01

The method to calculate the combination scattering power of the components of the dissociating N 2 O 4 ↔ 2NO 2 → 2NO+O 2 gas subjected to the laser radiation effect is given. The combination scattering power has been calculated for temperatures 400-600 K, pressures 1-3 MPa, with the neodymium laser (λ=1.06 μm) as a source and the possibility of measuring the local temperatures and concentration of the given gas components with the help of the combination scattering has been analysed. It follows from the calculated data that combination scattering power of N 2 O 4 ↔ 2NO 2 ↔ 2NO+O 2 gas in excitation with the neodymium laser as a source is sufficient for detection. Gas temperature is likely to be measured with the minimum error relative to stokes and anti-stokes bands of the combination scattering, produced by nitrogen tetroxide. From calculated data it also follows that measurement of NO 2 concentration in the range 400-600 K is possible. At the same time combination scattering power, produced by NO and O 2 components is sufficient for measurement merely with the concentration of the components of the order of 10 18 molecules/cm 3 guaranteed in static conditions only at N 2 O 4 ↔ 2NO 2 ↔ 2NO+O 2 gas temperature 500 K and higher

NSSS Component Control System Design of Integral Reactor

International Nuclear Information System (INIS)

Lee, Joon Koo; Kwon, Ho Je; Jeong, Kwong Il; Park, Heui Youn; Koo, In Soo

2005-01-01

MMIS(Man Machine Interface System) of an integral reactor is composed of a Control Room, Plant Protection System, Control System and Monitoring System which are related with the overall plant operation. MMIS is being developed with a new design concept and digital technology to reduce the Human Factor Error and improve the systems' safety, reliability and availability. And CCS(component control system) is also being developed with a new design concept and digital hardware technology A fully digitalized system and design concept are introduced in the NSSS CCS

Development of flaw assesment methodology for elevated temperature components of FBR plants

International Nuclear Information System (INIS)

Shimakawa, Takashi; Takahashi, Yukio; Miura, Naoki; Nakayama, Yasunari; Sawai, Tatsuaki; Tooya, Yuuji

1999-01-01

Fracture mechanics is applicable for the safety assessment of FBR component if a crack is assumed to exist. Inelastic response should be taken into account due to high temperature operation of FBR components. However, methodology for the application of inelastic fracture mechanics has not been established sufficiently. CRIEPI has been conducted research projects to develop a flaw assessment guideline for FBR components. This guideline consists of evaluation methods for creep-fatigue crack propagation, ductile fracture and sodium leak rate. The summary of evaluation methods on creep-fatigue crack and ductile fracture is presented in this paper. (author)

Hot laboratory design on the basis of standardized components

International Nuclear Information System (INIS)

Cadrot, J.

1976-01-01

The paper describes the principal effects on hot laboratory design brought about over the last 15 years by the use of standardized components developed jointly with the CEA and the industrial associates of AFINE. After a rapid survey of the various advantages of standardization, the author turns to the specific case of a laboratory producing mixed plutonium and uranium oxide fuels, giving a brief description of the glove-boxes and ancillary equipment. He then deals with the design of an isotope production laboratory. The basic component is the DR 200 standard cell, which permits the civil engineering work to be effected on modular principles. Use of a safety-flow pressure regulating valve makes possible pneumatic automation of the production-cell internals. A substantial gain in output is the result. In the next section the paper refers to a pilot facility for irradiated fuel studies, and describes the components used, which require taking into account the high activities and intense radiations encountered in studies of this type. The author then demonstrates the flexibility with which standardized components can be adapted to different uses, thus solving many distinct problems, an example of which is represented by a semi-hot box for handling up to 100g of americium-241. Finally, the paper offers a rapid summary of the effects of standardization at the various stages concerned, from initial design to the commissioning of a hot laboratory. (author)

Application of Combined Sustained and Cyclic Loading Test Results to Alloy 617 Elevated Temperature Design Criteria

Energy Technology Data Exchange (ETDEWEB)

Wang, Yanli [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jetter, Robert I [Global Egineering and Technology, LLC, Coral Gables, FL (United States); Sham, Sam [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2014-08-25

Alloy 617 is a reference structural material for very high temperature components of advanced-gas cooled reactors with outlet temperatures in the range of 900-950°C . In order for designers to be able to use Alloy 617 for these high temperature components, Alloy 617 has to be approved for use in Section III (the nuclear section) of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. A plan has been developed to submit a draft code for Alloy 617 to ASME Section III by 2015. However, the current rules in Subsection NH for the evaluation of strain limits and creep-fatigue damage using simplified methods based on elastic analysis have been deemed inappropriate for Alloy 617 at temperatures above 1200°F (650°C). The rationale for this exclusion is that at higher temperatures it is not feasible to decouple plasticity and creep deformation, which is the basis for the current simplified rules. This temperature, 1200 °F, is well below the temperature range of interest for this material in High Temperature Gas Cooled Reactor (HTGR) applications. The only current alternative is, thus, a full inelastic analysis which requires sophisticated material models which have been formulated but not yet verified. To address this issue, proposed code rules have been developed which are based on the use of elastic-perfectly plastic (EPP) analysis methods and which are expected to be applicable to very high temperatures.

Flow Components in a NaK Test Loop Designed to Simulate Conditions in a Nuclear Surface Power Reactor

Science.gov (United States)

Polzin, Kurt A.; Godfroy, Thomas J.

2008-01-01

A test loop using NaK as the working fluid is presently in use to study material compatibility effects on various components that comprise a possible nuclear reactor design for use on the lunar surface. A DC electromagnetic (EM) pump has been designed and implemented as a means of actively controlling the NaK flow rate through the system and an EM flow sensor is employed to monitor the developed flow rate. These components allow for the matching of the flow rate conditions in test loops with those that would be found in a full-scale surface-power reactor. The design and operating characteristics of the EM pump and flow sensor are presented. In the EM pump, current is applied to a set of electrodes to produce a Lorentz body force in the fluid. A measurement of the induced voltage (back-EMF) in the flow sensor provides the means of monitoring flow rate. Both components are compact, employing high magnetic field strength neodymium magnets thermally coupled to a water-cooled housing. A vacuum gap limits the heat transferred from the high temperature NaK tube to the magnets and a magnetically-permeable material completes the magnetic circuit. The pump is designed to produce a pressure rise of 5 psi, and the flow sensor's predicted output is roughly 20 mV at the loop's nominal flow rate of 0.5 GPM.

Design and application for a high-temperature nuclear heat source

International Nuclear Information System (INIS)

Quade, R.N.

1980-01-01

Recent actions by OPEC have sharply increased interest in the United States in synfuels, with coal being the logical choice for the carbon source. Two coal liquefaction processes, direct and indirect, have been examined. Each can produce about 50% more output when coupled to an HTGR for process heat. The nuclear reactor designed for process heat has a power output of 842MW(t), a core outlet temperature of 950 0 C (1742 0 F), and an intermediate helium loop to separate the heat source from the process heat exchangers. Steam-methane reforming is the reference process. As part of the development of a nuclear process heat system, a computer code, Process Heat Reactor Evaluation and Design, is being developed. This code models both the reactor plant and a steam reforming plant. When complete, the program will have the capability to calculate an overall mass and heat balance, size the plant components, and estimate the plant cost for a wide variety of independent variables. (author)

Development of expert system for structural design of FBR components

International Nuclear Information System (INIS)

Ueda, Hiroyoshi; Uno, Masayoshi; Ogawa, Hiroshi; Shimakawa, Takashi; Yoshimura, Shinobu; Yagawa, Genki.

1995-01-01

The characteristics of structural design processes for nuclear components can be summarized as follows : (1) Many engineers belonging to different fields are working in parallel, exchanging a huge amount of data and information. (2) A final solution is determined after a number of iterative design processes. (3) Solutions have to be examined many times based on sophisticated design codes. (4) Sophisticated calculation methods such as the finite element method are frequently utilized, and experts' knowledge on such analyses plays important roles in the design process. Taking these issues into consideration, a new expert system for structural design is developed in the present study. Here, the object-oriented data flow mechanism and the blackboard model are utilized to systematize structural design processes in a computer. An automated finite element calculation module is implemented, and experts' knowledge is stored in knowledge base. In addition, a new algorithm is employed to automatically draw the design window, which is defined as an area of permissible solutions in a design parameter space. The developed system is successfully applied to obtain the design windows of four components selected from the demonstration FBR structures. (author)

Note: A new design for a low-temperature high-intensity helium beam source

Science.gov (United States)

Lechner, B. A. J.; Hedgeland, H.; Allison, W.; Ellis, J.; Jardine, A. P.

2013-02-01

A high-intensity supersonic beam source is a key component of any atom scattering instrument, affecting the sensitivity and energy resolution of the experiment. We present a new design for a source which can operate at temperatures as low as 11.8 K, corresponding to a beam energy of 2.5 meV. The new source improves the resolution of the Cambridge helium spin-echo spectrometer by a factor of 5.5, thus extending the accessible timescales into the nanosecond range. We describe the design of the new source and discuss experiments characterizing its performance. Spin-echo measurements of benzene/Cu(100) illustrate its merit in the study of a typical slow-moving molecular adsorbate species.

Flaw assessment procedure for high temperature reactor components

International Nuclear Information System (INIS)

Ainsworth, R.A.; Takahashi, Y.

1990-01-01

An interim high-temperature flaw assessment procedure is described. This is a result of a collaborative effort between Electric Power Research Institute in the USA, Central Research Institute of Electric Power Industry in Japan, and Nuclear Electric plc in the UK. The procedure addresses preexisting defects subject to creep-fatigue loading conditions. Laws employed to calculate the crack growth per cycle are defined in terms of fracture mechanics parameters and constants related to the component material. The crack growth laws may be integrated to calculate the remaining life of a component or to predict the amount of crack extension in a given period. Fatigue and creep crack growth per cycle are calculated separately, and the total crack extension is taken as the simple sum of the two contributions. An interaction between the two propagation modes is accounted for in the material properties in the separate calculations. In producing the procedure, limitations of the approach have been identified. Some of these limitations are to be addressed in an extension of the current collaborative program. 20 refs

Peculiar features of low temperature deformation and strengthening of Cu-Nb bimetal components

International Nuclear Information System (INIS)

Lototskaya, V.A.; Il'ichev, V.Ya.

1988-01-01

The Cu-Nb bimetal treated in different ways is studied under conditions of uniaxial tension within the temperature range of 4.2...20 K. Stresses of the components being in the bimetal itself are estimated from the load jumps of the strain curve caused by the niobium layer failure. Stresses of components in the bimetal and in its separated layers are found to be different. Variation in the stressed state of a colddrawn and annealed bimetals is, in this case, a factor which determines the stress difference. This variation is accounted for by different structural states of the copper layer under low-temperature localization of the plastic niobium deformation the plastic niobium deformation

Photovoltaic optimizer boost converters: Temperature influence and electro-thermal design

International Nuclear Information System (INIS)

Graditi, G.; Adinolfi, G.; Tina, G.M.

2014-01-01

Highlights: • The influence of temperature on DC–DC converter devices properties is considered. • An electro-thermal design method for PV power optimizer converters is proposed. • The electro-thermal design method proposed is applied to DR boost and SR boost. • Efficiency results of the designed SR converter and DR converters are presented. - Abstract: Objective: Photovoltaic (PV) systems can operate in presence of not uniform working conditions caused by continuously changing temperature and irradiance values and mismatching and shadowing phenomena. The more the PV system works in these conditions, the more its energy performances are negatively affected. Distributed Maximum Power Point Tracking (DMPPT) converters are now increasingly used to overcome this problem and to improve PV applications efficiency. A DMPPT system consists in a DC–DC converters equipped with a suitable controller dedicated to the Maximum Power Point Tracking (MPPT) of a single PV module. It is arranged either inside the junction-box or in a separate box close to the PV generator. Many power optimizers are now commercially available. In spite of different adopted DC–DC converter topologies, the shared interests of DMPPT systems designers are the high efficiency and reliability values. It is worth noting that to obtain so high performances converters, electronic components have to be carefully selected between the whole commercial availability and appropriately matched together. In this scenario, an electro-thermal design methodology is proposed and a reliability study by means of the Military Handbook 217F is carried out. Method: The developed DMPPT converters design method is constituted by many steps. In fact, beginning from installation site, PV generators and load data, this process selects power optimizers commercially available devices and it verifies their electro-thermal behavior to the aim to identify a set of suitable components for DMPPT applications. Repeating this

Endogenous and exogenous components in the circadian variation of core body temperature in humans

NARCIS (Netherlands)

Hiddinga, AE; Beersma, DGM; VandenHoofdakker, RH

Core body temperature is predominantly modulated by endogenous and exogenous components. In the present study we tested whether these two components can be reliably assessed in a protocol which lasts for only 120 h. In this so-called forced desynchrony protocol, 12 healthy male subjects (age 23.7

The RCC-MR design code for LMFBR components. A useful basis for fusion reactor design tools development

International Nuclear Information System (INIS)

Acker, D.; Chevereau, G.

1986-01-01

LMFBR and fusion reactors exhibit common features with regard to structural materials, temperature service level, loading types. So, design and construction rules used in France for LMFBR, that is to say RCC-MR Code, can constitute a good basis for fusion reactors design. Some original aspects of RCC-MR design rules are described, relating to unsignificant creep, ratchetting effect, fatigue and creep damage limits, creep damage evaluation, fatigue damage evaluation, buckling. The main originality of RCC-MR consists to propose comprehensive simplified rules based on elastic calculations and extended from classical cold temperatures to the elevated temperature domain. (author)

Secure wireless embedded systems via component-based design

DEFF Research Database (Denmark)

Hjorth, T.; Torbensen, R.

2010-01-01

This paper introduces the method secure-by-design as a way of constructing wireless embedded systems using component-based modeling frameworks. This facilitates design of secure applications through verified, reusable software. Following this method we propose a security framework with a secure c......, with full support for confidentiality, authentication, and integrity using keypairs. The approach has been demonstrated in a multi-platform home automation prototype that can remotely unlock a door using a PDA over the Internet....

Sodium immersible high temperature microphone design description

International Nuclear Information System (INIS)

Gavin, A.P.; Anderson, T.T.; Janicek, J.J.

1975-02-01

Argonne National Laboratory has developed a rugged high-temperature (HT) microphone for use as a sodium-immersed acoustic monitor in Liquid Metal Fast Breeder Reactors (LMFBRs). Microphones of this design have been extensively tested in room temperature water, in air up to 1200 0 F, and in sodium up to 1200 0 F. They have been successfully installed and employed as acoustic monitors in several operating liquid metal systems. The design, construction sequence, calibration, and testing of these microphones are described. 6 references. (U.S.)

Design for ASIC reliability for low-temperature applications

Science.gov (United States)

Chen, Yuan; Mojaradi, Mohammad; Westergard, Lynett; Billman, Curtis; Cozy, Scott; Burke, Gary; Kolawa, Elizabeth

2005-01-01

In this paper, we present a methodology to design for reliability for low temperature applications without requiring process improvement. The developed hot carrier aging lifetime projection model takes into account both the transistor substrate current profile and temperature profile to determine the minimum transistor size needed in order to meet reliability requirements. The methodology is applicable for automotive, military, and space applications, where there can be varying temperature ranges. A case study utilizing this methodology is given to design for reliability into a custom application-specific integrated circuit (ASIC) for a Mars exploration mission.

Study on system layout and component design in the HTTR hydrogen production system. Contract research

Energy Technology Data Exchange (ETDEWEB)

Nishihara, Tetsuo; Shimizu, Akira [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tanihira, Masanori [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Uchida, Shoji [Advanced Reactor Technology Co., Ltd., Tokyo (Japan)

2003-01-01

The global warming becomes a significant issue in the world so that it needs to reduce the CO{sub 2} emission. It is expected that hydrogen is in place of the fossil fuels such as coal and oil, and plays the important role to resolve the global warming. There are several hydrogen making processes such as water electrolysis and steam reforming of hydrocarbon. Steam reforming of hydrocarbon is a major hydrogen making process because of economy in industry. It utilizes the fossil fuels as process heat for chemical reaction and results in a large CO{sub 2} emission. New steam reforming system without fossil fuel can contribute to resolve the global warming. High temperature gas-cooled reactor (HTGR) has a unique feature to be able to supply a hot helium gas whose temperature is approximately 950degC at the reactor outlet. This makes HTGR possible to utilize for not only power generation but also process heat utilization. JAERI constructed the high temperature engineering test reactor (HTTR) that is a sort of HTGR in Oarai establishment and starts operation. Nuclear heat utilization is one of the R and D items of the HTTR. The steam reforming system coupling to the HTTR for hydrogen production has been designed. This report represents the system layout and design specification of key components in HTTR steam reforming system. (author)

Study on system layout and component design in the HTTR hydrogen production system. Contract research

International Nuclear Information System (INIS)

Nishihara, Tetsuo; Shimizu, Akira; Uchida, Shoji

2003-01-01

The global warming becomes a significant issue in the world so that it needs to reduce the CO 2 emission. It is expected that hydrogen is in place of the fossil fuels such as coal and oil, and plays the important role to resolve the global warming. There are several hydrogen making processes such as water electrolysis and steam reforming of hydrocarbon. Steam reforming of hydrocarbon is a major hydrogen making process because of economy in industry. It utilizes the fossil fuels as process heat for chemical reaction and results in a large CO 2 emission. New steam reforming system without fossil fuel can contribute to resolve the global warming. High temperature gas-cooled reactor (HTGR) has a unique feature to be able to supply a hot helium gas whose temperature is approximately 950degC at the reactor outlet. This makes HTGR possible to utilize for not only power generation but also process heat utilization. JAERI constructed the high temperature engineering test reactor (HTTR) that is a sort of HTGR in Oarai establishment and starts operation. Nuclear heat utilization is one of the R and D items of the HTTR. The steam reforming system coupling to the HTTR for hydrogen production has been designed. This report represents the system layout and design specification of key components in HTTR steam reforming system. (author)

C-Based Design Methodology and Topological Change for an Indian Agricultural Tractor Component

Science.gov (United States)

Matta, Anil Kumar; Raju, D. Ranga; Suman, K. N. S.; Kranthi, A. S.

2018-06-01

The failure of tractor components and their replacement has now become very common in India because of re-cycling, re-sale, and duplication. To over come the problem of failure we propose a design methodology for topological change co-simulating with software's. In the proposed Design methodology, the designer checks Paxial, Pcr, Pfailue, τ by hand calculations, from which refined topological changes of R.S.Arm are formed. We explained several techniques employed in the component for reduction, removal of rib material to change center of gravity and centroid point by using system C for mixed level simulation and faster topological changes. The design process in system C can be compiled and executed with software, TURBO C7. The modified component is developed in proE and analyzed in ANSYS. The topologically changed component with slot 120 × 4.75 × 32.5 mm at the center showed greater effectiveness than the original component.

C-Based Design Methodology and Topological Change for an Indian Agricultural Tractor Component

Science.gov (United States)

Matta, Anil Kumar; Raju, D. Ranga; Suman, K. N. S.; Kranthi, A. S.

2018-02-01

The failure of tractor components and their replacement has now become very common in India because of re-cycling, re-sale, and duplication. To over come the problem of failure we propose a design methodology for topological change co-simulating with software's. In the proposed Design methodology, the designer checks Paxial, Pcr, Pfailue, τ by hand calculations, from which refined topological changes of R.S.Arm are formed. We explained several techniques employed in the component for reduction, removal of rib material to change center of gravity and centroid point by using system C for mixed level simulation and faster topological changes. The design process in system C can be compiled and executed with software, TURBO C7. The modified component is developed in proE and analyzed in ANSYS. The topologically changed component with slot 120 × 4.75 × 32.5 mm at the center showed greater effectiveness than the original component.

Recent advances in design procedures for high temperature plant

International Nuclear Information System (INIS)

1988-01-01

Thirteen papers cover several aspects of design for high temperature plant. These include design codes, computerized structural analysis and mechanical properties of materials at high temperatures. Seven papers are relevant for fast reactors and these are indexed separately. These cover shakedown design, design codes for thin shells subjected to cyclic thermal loading, the inelastic behaviour of stainless steels and creep and crack propagation in reactor structures under stresses caused by thermal cycling loading. (author)

Development of computational methods of design by analysis for pressure vessel components

International Nuclear Information System (INIS)

Bao Shiyi; Zhou Yu; He Shuyan; Wu Honglin

2005-01-01

Stress classification is not only one of key steps when pressure vessel component is designed by analysis, but also a difficulty which puzzles engineers and designers at all times. At present, for calculating and categorizing the stress field of pressure vessel components, there are several computation methods of design by analysis such as Stress Equivalent Linearization, Two-Step Approach, Primary Structure method, Elastic Compensation method, GLOSS R-Node method and so on, that are developed and applied. Moreover, ASME code also gives an inelastic method of design by analysis for limiting gross plastic deformation only. When pressure vessel components design by analysis, sometimes there are huge differences between the calculating results for using different calculating and analysis methods mentioned above. As consequence, this is the main reason that affects wide application of design by analysis approach. Recently, a new approach, presented in the new proposal of a European Standard, CEN's unfired pressure vessel standard EN 13445-3, tries to avoid problems of stress classification by analyzing pressure vessel structure's various failure mechanisms directly based on elastic-plastic theory. In this paper, some stress classification methods mentioned above, are described briefly. And the computational methods cited in the European pressure vessel standard, such as Deviatoric Map, and nonlinear analysis methods (plastic analysis and limit analysis), are depicted compendiously. Furthermore, the characteristics of computational methods of design by analysis are summarized for selecting the proper computational method when design pressure vessel component by analysis. (authors)

Design and operation results of nitrogen gas baking system for KSTAR plasma facing components

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang-Tae [National Fusion Research Institute, 113 Gwahang-ro, Yuseong-gu, Daejeon 305-806 (Korea, Republic of); Kim, Young-Jin, E-mail: k43689@nfri.re.kr [National Fusion Research Institute, 113 Gwahang-ro, Yuseong-gu, Daejeon 305-806 (Korea, Republic of); Joung, Nam-Yong; Im, Dong-Seok; Kim, Kang-Pyo; Kim, Kyung-Min; Bang, Eun-Nam; Kim, Yaung-Soo [National Fusion Research Institute, 113 Gwahang-ro, Yuseong-gu, Daejeon 305-806 (Korea, Republic of); Yoo, Seong-Yeon [Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon 305-764 (Korea, Republic of)

2013-11-15

Highlights: • Vacuum pressure in a vacuum vessel arrived at 7.24 × 10{sup −8} mbar. • PFC temperature was reached maximum 250 °C by gas temperature at 300 °C. • PFC inlet gas temperature was changed 5 °C per hour during rising and falling. • PFC gas balancing was made temperature difference among them below 8.3 °C. • System has a pre-cooler and a three-way valve to save operation energy. -- Abstract: A baking system for the Korea Superconducting Tokamak Advanced Research (KSTAR) plasma facing components (PFCs) is designed and operated to achieve vacuum pressure below 5 × 10{sup −7} mbar in vacuum vessel with removing impurities. The purpose of this research is to prevent the fracture of PFC because of thermal stress during baking the PFC, and to accomplish stable operation of the baking system with the minimum life cycle cost. The uniformity of PFC temperature in each sector was investigated, when the supply gas temperature was varied by 5 °C per hour using a heater and the three-way valve at the outlet of a compressor. The alternative of the pipe expansion owing to hot gas and the cage configuration of the three-way valve were also studied. During the fourth campaign of the KSTAR in 2011, nitrogen gas temperature rose up to 300 °C, PFC temperature reached at 250 °C, the temperature difference among PFCs was maintained at below 8.3 °C, and vacuum pressure of up to 7.24 × 10{sup −8} mbar was achieved inside the vacuum vessel.

Investigation of effective factors of transient thermal stress of the MONJU-System components

Energy Technology Data Exchange (ETDEWEB)

Inoue, Masaaki; Hirayama, Hiroshi; Kimura, Kimitaka; Jinbo, M. [Toshiba Corp., Kawasaki, Kanagawa (Japan)

1999-03-01

Transient thermal stress of each system Component in the fast breeder reactor is an uncertain factor on it's structural design. The temperature distribution in a system component changes over a wide range in time and in space. An unified evaluation technique of thermal, hydraulic, and structural analysis, in which includes thermal striping, temperature stratification, transient thermal stress and the integrity of the system components, is required for the optimum design of tho fast reactor plant. Thermal boundary conditions should be set up by both the transient thermal stress analysis and the structural integrity evaluation of each system component. The reasonable thermal boundary conditions for the design of the MONJU and a demonstration fast reactor, are investigated. The temperature distribution analysis models and the thermal boundary conditions on the Y-piece structural parts of each system component, such as reactor vessel, intermediate heat exchanger, primary main circulation pump, steam generator, superheater and upper structure of reactor core, are illustrated in the report. (M. Suetake)

Proof of fatigue strength of ferritic and austenitic nuclear components

Energy Technology Data Exchange (ETDEWEB)

Roos, E.; Herter, K.H.; Schuler, X.; Weissenberg, T. [Materialpruefungsanstalt, Univ. Stuttgart (Germany)

2009-07-01

For the construction, design and operation of nuclear components and systems the appropriate technical codes and standards provide material data, detailed stress analysis procedures and a design philosophy which guarantees a reliable behaviour of the structural components throughout the specified lifetime. Especially for cyclic stress evaluation the different codes and standards provide different fatigue analyses procedures to be performed considering the various mechanical and thermal loading histories and geometric complexities of the components. For the fatigue design curves used as limiting criteria the influence of different factors like e.g., environment, surface finish and temperature must be taken into consideration in an appropriate way. Fatigue tests were performed with low alloy steels as well as with Nb- and Ti-stabilized German austenitic stainless steels in air and simulated high temperature boiling water reactor environment. The experimental results are compared and valuated with the mean data curves in air as well as with mean data curves under high temperature water environment published in the international literature. (orig.)

Components of the primary circuit of LWRs. Design, construction and calculation. Draft. Komponenten des Primaerkreises von Leichtwasserreaktoren. Auslegung, Konstruktion und Berechnung. Entwurf

Energy Technology Data Exchange (ETDEWEB)

1979-10-01

This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673/sup 0/K (400/sup 0/C). The primary circuit as the pressure continement of the reactor coolant comprises: Reactor pressure vessel (without internals), steam generator (primary loop), pressurizer, reactor coolant pump housing, interconnecting pipings between the components mentioned above and appropriate various valve and instrument casings, pipings branding off from the above components and interconnecting pipings, including the appropriate instrument casings, up to and including the first isolating valve, pressure shielding of control rod drives.

The mechanical behavior and reliability prediction of the HTR graphite component at various temperature and neutron dose ranges

International Nuclear Information System (INIS)

Fang, Xiang; Yu, Suyuan; Wang, Haitao; Li, Chenfeng

2014-01-01

Highlights: • The mechanical behavior of graphite component in HTRs under high temperature and neutron irradiation conditions is simulated. • The computational process of mechanical analysis is introduced. • Deformation, stresses and failure probability of the graphite component are obtained and discussed. • Various temperature and neutron dose ranges are selected in order to investigate the effect of in-core conditions on the results. - Abstract: In a pebble-bed high temperature gas-cooled reactor (HTR), nuclear graphite serves as the main structural material of the side reflectors. The reactor core is made up of a large number of graphite bricks. In the normal operation case of the reactor, the maximum temperature of the helium coolant commonly reaches about 750 °C. After around 30 years’ full power operation, the peak value of in-core fast neutron cumulative dose reaches to 1 × 10 22 n cm −2 (EDN). Such high temperature and neutron irradiation strongly impact the behavior of graphite component, causing obvious deformation. The temperature and neutron dose are unevenly distributed inside a graphite brick, resulting in stress concentrations. The deformation and stress concentration can both greatly affect safety and reliability of the graphite component. In addition, most of the graphite properties (such as Young's modulus and coefficient of thermal expansion) change remarkably under high temperature and neutron irradiations. The irradiation-induced creep also plays a very important role during the whole process, and provides a significant impact on the stress accumulation. In order to simulate the behavior of graphite component under various in-core conditions, all of the above factors must be considered carefully. In this paper, the deformation, stress distribution and failure probability of a side graphite component are studied at various temperature points and neutron dose levels. 400 °C, 500 °C, 600 °C and 750 °C are selected as the

Nuclear heat source component design considerations for HTGR process heat reactor plant concept

International Nuclear Information System (INIS)

McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

1982-01-01

Using alternate energy sources abundant in the U.S.A. to help curb foreign oil imports is vitally important from both national security and economic standpoints. Perhaps the most forwardlooking opportunity to realize national energy goals involves the integrated use of two energy sources that have an established technology base in the U.S.A., namely nuclear energy and coal. The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc.) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported (via an intermediate heat exchanger (IHX)) to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

Structural analysis technology for high-temperature design

International Nuclear Information System (INIS)

Greenstreet, W.L.

1977-01-01

Results from an ongoing program devoted to the development of verified high-temperature structural design technology applicable to nuclear reactor systems are described. The major aspects addressed by the program are (1) deformation behavior; (2) failure associated with creep rupture, brittle fracture, fatigue, creep-fatigue interactions, and crack propagation; and (3) the establishment of appropriate design criteria. This paper discusses information developed in the deformation behavior category. The material considered is type 304 stainless steel, and the temperatures range to 1100 0 F (593 0 C). In essence, the paper considers the ingredients necessary for predicting relatively high-temperature inelastic deformation behavior of engineering structures under time-varying temperature and load conditions and gives some examples. These examples illustrate the utility and acceptability of the computational methods identified and developed for prediting essential features of complex inelastic behaviors. Conditions and responses that can be encountered under nuclear reactor service conditions and invoked in the examples. (Auth.)

The RCC-MR design code for LMFBR components. A useful basic for fusion reactor design tools development

International Nuclear Information System (INIS)

Acker, D.; Chevereau, G.

1985-11-01

LMFBR and fusion reactors exhibit common features with regard to structural materials (Stainless steels), temperature service level (550-600 0 C), loading types. So, design and construction rules used in France for LMFBR, that is to say RCC-MR Code, can constitute a good basis for fusion reactors design. Some original aspects of RCC-MR design rules are described, relating to unsignificant creep, ratchetting effect, fatigue and creep damage limits, creep damage evaluation, fatigue damage evaluation, buckling. The main originality of RCC-MR consists to propose comprehensive simplified rules based on elastic calculations and extended from classical cold temperatures to the elevated temperature domain

Material design data of 2.25Cr-1Mo steel and hastelloy-x for the experimental multi-purpose very-high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Kodaira, Tsuneo; Suzuki, Michiaki; Uga, Takeo

1975-08-01

The preliminary structural design guidelines for the experimental multi-purpose very-high temperature gas-cooled reactor have recently been prepared. The components of the primary system operating at temperatures of creep dominant range are grouped in those of pressure and temperature boundaries respectively. In the material selection, 2 1/4Cr-1Mo steel is chosen for the former and Hastelloy-X for the latter taking into account of material properties at operating temperature. Deriving from the literature in the field, material design data of the alloys are established in design forms such as Sy, So, Sm, St, 100% of minimum stress to rupture, design fatigue curves, isochronous stress-strain curves, creep-fatigue interaction damage factor and so on, which are defined in ASME Code Section III, Code Case 1592. (auth.)

Development of Very High Temperature Reactor Design Technology

International Nuclear Information System (INIS)

Lee, Won Jae; Noh, J. M.; Kim, K. S.

2007-05-01

To develop design technologies for the VHTR (Very High Temperature Gas-cooled Reactor) of 950 .deg. C outlet temperature for an efficient hydrogen production, key studies were performed, which include evaluation technology for the performance and safety of VHTR, and development of design and analysis codes. First, to evaluate the performance of VHTR, a series of analyses has been carried out for core characteristics at 950 .deg. C, cooled-vessel adopting internal flow path through the graphite structure, compact heat exchanger with periodic channel configuration, intermediate loop system, risk/performance-informed method, and high temperature structural integrity. Through the analyses of major accidents such as HPCC and LPCC, safety evaluation of both VHTR and RCCS has been also performed. In addition, prototype codes have been developed for a nuclear design, system loop design, system performance analysis, air-ingress accident analysis, fission product/tritium transport analysis, graphite structure seismic analysis and hydrogen explosion analysis, and they are being verified and validated through a lot of international collaborations

High temperature creep-fatigue design

International Nuclear Information System (INIS)

Tavassoli, A. A. F.; Fournier, B.; Sauzay, M.

2010-01-01

Generation IV fission and future fusion reactors envisage development of more efficient high temperature concepts where materials performances are key to their success. This paper examines different types of high temperature creep-fatigue interactions and their implications on design rules for the structural materials retained in both programmes. More precisely, the paper examines current status of design rules for the stainless steel type 316L(N), the conventional Modified 9Cr-1Mo martensitic steel and the low activation Eurofer steel. Results obtained from extensive high temperature creep, fatigue and creep-fatigue tests performed on these materials and their welded joints are presented. These include sequential creep-fatigue and relaxation creep-fatigue tests with hold times in tension, in compression or in both. Effects of larger plastic deformations on fatigue properties are studied through cyclic creep tests or fatigue tests with extended hold time in creep. In most cases, mechanical test results are accompanied with microstructural and fractographic observations. In the case of martensitic steels, the effect of oxidation is examined by performing creep-fatigue tests on identical specimens in vacuum. Results obtained are analyzed and their implications on design allowable and creep-fatigue interaction diagrams are presented. While reasonable confidence is found in predicting creep-fatigue damage through existing code procedures for austenitic stainless steels, effects of cyclic softening and coarsening of microstructure of martensitic steels throughout the fatigue life on materials properties need to be taken into account for more precise damage calculations. In the long-term, development of ferritic/martensitic steels with stable microstructure, such as ODS steels, is proposed. (authors)

High temperature creep-fatigue design

Energy Technology Data Exchange (ETDEWEB)

Tavassoli, A. A. F.; Fournier, B.; Sauzay, M. [CEA Saclay, DEN DMN, F-91191 Gif Sur Yvette (France)

2010-07-01

Generation IV fission and future fusion reactors envisage development of more efficient high temperature concepts where materials performances are key to their success. This paper examines different types of high temperature creep-fatigue interactions and their implications on design rules for the structural materials retained in both programmes. More precisely, the paper examines current status of design rules for the stainless steel type 316L(N), the conventional Modified 9Cr-1Mo martensitic steel and the low activation Eurofer steel. Results obtained from extensive high temperature creep, fatigue and creep-fatigue tests performed on these materials and their welded joints are presented. These include sequential creep-fatigue and relaxation creep-fatigue tests with hold times in tension, in compression or in both. Effects of larger plastic deformations on fatigue properties are studied through cyclic creep tests or fatigue tests with extended hold time in creep. In most cases, mechanical test results are accompanied with microstructural and fractographic observations. In the case of martensitic steels, the effect of oxidation is examined by performing creep-fatigue tests on identical specimens in vacuum. Results obtained are analyzed and their implications on design allowable and creep-fatigue interaction diagrams are presented. While reasonable confidence is found in predicting creep-fatigue damage through existing code procedures for austenitic stainless steels, effects of cyclic softening and coarsening of microstructure of martensitic steels throughout the fatigue life on materials properties need to be taken into account for more precise damage calculations. In the long-term, development of ferritic/martensitic steels with stable microstructure, such as ODS steels, is proposed. (authors)

HYFIRE II: fusion/high-temperature electrolysis conceptual-design study. Annual report

International Nuclear Information System (INIS)

Fillo, J.A.

1983-08-01

As in the previous HYFIRE design study, the current study focuses on coupling a Tokamak fusion reactor with a high-temperature blanket to a High-Temperature Electrolyzer (HTE) process to produce hydrogen and oxygen. Scaling of the STARFIRE reactor to allow a blanket power to 6000 MW(th) is also assumed. The primary difference between the two studies is the maximum inlet steam temperature to the electrolyzer. This temperature is decreased from approx. 1300 0 to approx. 1150 0 C, which is closer to the maximum projected temperature of the Westinghouse fuel cell design. The process flow conditions change but the basic design philosophy and approaches to process design remain the same as before. Westinghouse assisted in the study in the areas of systems design integration, plasma engineering, balance-of-plant design, and electrolyzer technology

Stresses evolution at high temperature (200°C on the interface of thin films in magnetic components

Directory of Open Access Journals (Sweden)

Doumit Nicole

2014-07-01

Full Text Available In the field of electronics, the increase of operating temperatures is a major industrial and scientific challenge because it allows reducing mass and volume of components especially in the aeronautic domain. So minimizing our components reduce masses and the use of cooling systems. For that, the behaviours and interface stresses of our components (in particular magnetic inductors and transformers that are constituted of one magnetic layer (YIG or an alumina substrate (Al2O3 representing the substrate and a thin copper film are studied at high temperature (200°C. COMSOL Multiphysics is used to simulate our work and to validate our measurements results. In this paper, we will present stresses results according to the geometrical copper parameters necessary for the component fabrication. Results show that stresses increase with temperature and copper’s thickness while remaining always lower than 200MPa which is the rupture stress value.

Study of heat fluxes on plasma facing components in a tokamak from measurements of temperature by infrared thermography

International Nuclear Information System (INIS)

Daviot, R.

2010-05-01

The goal of this thesis is the development of a method of computation of those heat loads from measurements of temperature by infrared thermography. The research was conducted on three issues arising in current tokamaks but also future ones like ITER: the measurement of temperature on reflecting walls, the determination of thermal properties for deposits observed on the surface of tokamak components and the development of a three-dimensional, non-linear computation of heat loads. A comparison of several means of pyrometry, monochromatic, bi-chromatic and photothermal, is performed on an experiment of temperature measurement. We show that this measurement is sensitive to temperature gradients on the observed area. Layers resulting from carbon deposition by the plasma on the surface of components are modeled through a field of equivalent thermal resistance, without thermal inertia. The field of this resistance is determined, for each measurement points, from a comparison of surface temperature from infrared thermographs with the result of a simulation, which is based on a mono-dimensional linear model of components. The spatial distribution of the deposit on the component surface is obtained. Finally, a three-dimensional and non-linear computation of fields of heat fluxes, based on a finite element method, is developed here. Exact geometries of the component are used. The sensitivity of the computed heat fluxes is discussed regarding the accuracy of the temperature measurements. This computation is applied to two-dimensional temperature measurements of the JET tokamak. Several components of this tokamak are modeled, such as tiles of the divertor, upper limiter and inner and outer poloidal limiters. The distribution of heat fluxes on the surface of these components is computed and studied along the two main tokamak directions, poloidal and toroidal. Toroidal symmetry of the heat loads from one tile to another is shown. The influence of measurements spatial resolution

A Fixed Point VHDL Component Library for a High Efficiency Reconfigurable Radio Design Methodology

Science.gov (United States)

Hoy, Scott D.; Figueiredo, Marco A.

2006-01-01

Advances in Field Programmable Gate Array (FPGA) technologies enable the implementation of reconfigurable radio systems for both ground and space applications. The development of such systems challenges the current design paradigms and requires more robust design techniques to meet the increased system complexity. Among these techniques is the development of component libraries to reduce design cycle time and to improve design verification, consequently increasing the overall efficiency of the project development process while increasing design success rates and reducing engineering costs. This paper describes the reconfigurable radio component library developed at the Software Defined Radio Applications Research Center (SARC) at Goddard Space Flight Center (GSFC) Microwave and Communications Branch (Code 567). The library is a set of fixed-point VHDL components that link the Digital Signal Processing (DSP) simulation environment with the FPGA design tools. This provides a direct synthesis path based on the latest developments of the VHDL tools as proposed by the BEE VBDL 2004 which allows for the simulation and synthesis of fixed-point math operations while maintaining bit and cycle accuracy. The VHDL Fixed Point Reconfigurable Radio Component library does not require the use of the FPGA vendor specific automatic component generators and provide a generic path from high level DSP simulations implemented in Mathworks Simulink to any FPGA device. The access to the component synthesizable, source code provides full design verification capability:

Design Optimization Method for Composite Components Based on Moment Reliability-Sensitivity Criteria

Science.gov (United States)

Sun, Zhigang; Wang, Changxi; Niu, Xuming; Song, Yingdong

2017-08-01

In this paper, a Reliability-Sensitivity Based Design Optimization (RSBDO) methodology for the design of the ceramic matrix composites (CMCs) components has been proposed. A practical and efficient method for reliability analysis and sensitivity analysis of complex components with arbitrary distribution parameters are investigated by using the perturbation method, the respond surface method, the Edgeworth series and the sensitivity analysis approach. The RSBDO methodology is then established by incorporating sensitivity calculation model into RBDO methodology. Finally, the proposed RSBDO methodology is applied to the design of the CMCs components. By comparing with Monte Carlo simulation, the numerical results demonstrate that the proposed methodology provides an accurate, convergent and computationally efficient method for reliability-analysis based finite element modeling engineering practice.

Mechanical design assessments of structural components and auxiliaries of the Joint European Torus

International Nuclear Information System (INIS)

Sonnerup, L.

1986-01-01

The general design of the Joint European Torus (JET) is briefly described. The loads on its major structural components, at normal operation, and in cases of plasma instability and/or disruption, are discussed. The way these components have been assessed and optimised in relation to their loads is presented. A short account of mechanical design problems of auxiliary equipment is given. Finally, the state of operation of JET and its implications for the mechanical design is summarized. The mechanically most important components of the JET device are the support structure of the toroidal magnet, the vacuum vessel, the coils of the magnets and the pedestals supporting the weight of the torus. These components all participate in resisting and transmitting the primary forces during operation. (orig.)

High temperature structure design for FBRs and analysis technology

International Nuclear Information System (INIS)

Iwata, Koji

1986-01-01

In the case of FBRs, the operation temperature exceeds 500 deg C, therefore, the design taking the inelastic characteristics of structural materials, such as plasticity and creep, into account is required, and the high grade and detailed evaluation of design is demanded. This new high temperature structure design technology has been advanced in respective countries taking up experimental, prototype and demonstration reactors as the targets. The development of FBRs in Japan was begun with the experimental reactor 'Joyo' which has been operated since 1977, and now, the prototype FBR 'Monju' of 280 MWe is under construction, which is expected to attain the criticality in 1992. In order to realize FBRs which can compete with LWRs through the construction of a demonstration FBR, the construction of large scale plants and the heightening of the economy and reliability are necessary. The features and the role of FBR structural design, the method of high temperature structure design and the trend of its standardization, the trend of the structural analysis technology for FBRs such as inelastic analysis, buckling analysis and fluid and structure coupled vibration analysis, the present status of structural analysis programs, and the subjects for the future of high temperature structure design are explained. (Kako, I.)

Design and component specifications for high average power laser optical systems

Energy Technology Data Exchange (ETDEWEB)

O' Neil, R.W.; Sawicki, R.H.; Johnson, S.A.; Sweatt, W.C.

1987-01-01

Laser imaging and transport systems are considered in the regime where laser-induced damage and/or thermal distortion have significant design implications. System design and component specifications are discussed and quantified in terms of the net system transport efficiency and phase budget. Optical substrate materials, figure, surface roughness, coatings, and sizing are considered in the context of visible and near-ir optical systems that have been developed at Lawrence Livermore National Laboratory for laser isotope separation applications. In specific examples of general applicability, details of the bulk and/or surface absorption, peak and/or average power damage threshold, coating characteristics and function, substrate properties, or environmental factors will be shown to drive the component size, placement, and shape in high-power systems. To avoid overstressing commercial fabrication capabilities or component design specifications, procedures will be discussed for compensating for aberration buildup, using a few carefully placed adjustable mirrors. By coupling an aggressive measurements program on substrates and coatings to the design effort, an effective technique has been established to project high-power system performance realistically and, in the process, drive technology developments to improve performance or lower cost in large-scale laser optical systems. 13 refs.

Design and component specifications for high average power laser optical systems

International Nuclear Information System (INIS)

O'Neil, R.W.; Sawicki, R.H.; Johnson, S.A.; Sweatt, W.C.

1987-01-01

Laser imaging and transport systems are considered in the regime where laser-induced damage and/or thermal distortion have significant design implications. System design and component specifications are discussed and quantified in terms of the net system transport efficiency and phase budget. Optical substrate materials, figure, surface roughness, coatings, and sizing are considered in the context of visible and near-ir optical systems that have been developed at Lawrence Livermore National Laboratory for laser isotope separation applications. In specific examples of general applicability, details of the bulk and/or surface absorption, peak and/or average power damage threshold, coating characteristics and function, substrate properties, or environmental factors will be shown to drive the component size, placement, and shape in high-power systems. To avoid overstressing commercial fabrication capabilities or component design specifications, procedures will be discussed for compensating for aberration buildup, using a few carefully placed adjustable mirrors. By coupling an aggressive measurements program on substrates and coatings to the design effort, an effective technique has been established to project high-power system performance realistically and, in the process, drive technology developments to improve performance or lower cost in large-scale laser optical systems. 13 refs

Standardization Efforts for Mechanical Testing and Design of Advanced Ceramic Materials and Components

Science.gov (United States)

Salem, Jonathan A.; Jenkins, Michael G.

2003-01-01

Advanced aerospace systems occasionally require the use of very brittle materials such as sapphire and ultra-high temperature ceramics. Although great progress has been made in the development of methods and standards for machining, testing and design of component from these materials, additional development and dissemination of standard practices is needed. ASTM Committee C28 on Advanced Ceramics and ISO TC 206 have taken a lead role in the standardization of testing for ceramics, and recent efforts and needs in standards development by Committee C28 on Advanced Ceramics will be summarized. In some cases, the engineers, etc. involved are unaware of the latest developments, and traditional approaches applicable to other material systems are applied. Two examples of flight hardware failures that might have been prevented via education and standardization will be presented.

Mechanical design assessments of structural components and auxiliaries of the Joint European Torus

International Nuclear Information System (INIS)

Sonnerup, L.

1985-01-01

The general design of the Joint European Torus (JET) is briefly described. The loads on its major structural components, at normal operation, and in cases of plasma instability and/or disruption, are discussed. The way these components have been assessed and optimised in relation to their loads is presented. A short account of mechanical design problems of auxiliary equipment is given. Finally, the state of operation of JET and its implications for the mechanical design at the time of the conference will be summarized. The mechanically most important components of the JET device are the support structure of the toroidal magnet, th vacuum vessel, the coils of the magnets and the pedestals supporting the weight of the torus. These components all participate in resisting and transmitting the primary forces during operation. (orig.)

Finite element based design optimization of WENDELSTEIN 7-X divertor components under high heat flux loading

International Nuclear Information System (INIS)

Plankensteiner, A.; Leuprecht, A.; Schedler, B.; Scheiber, K.-H.; Greuner, H.

2007-01-01

In the divertor of the nuclear fusion experiment WENDELSTEIN 7-X (W7-X) plasma facing high heat flux target elements have to withstand severe loading conditions. The thermally induced mechanical stressing turns out to be most critical with respect to lifetime predictions of the target elements. Therefore, different design variants of those CFC flat tile armoured high heat flux components have been analysed via the finite element package ABAQUS aiming at derivation of an optimized component design under high heat flux conditions. The investigated design variants comprise also promising alterations in the cooling channel design and castellation of the CFC flat tiles which, however, from a system integration and manufacturing standpoint of view, respectively, are evaluated to be critical. Therefore, the numerical study as presented here mainly comprises a reference variant that is comparatively studied with a variant incorporating a bi-layer-type AMC-Cu/OF-Cu interlayer at the CFC/Cu-interface. The thermo-mechanical material characteristics are accounted for in the finite element models with elastic-plastic properties being assigned to the metallic sections CuCrZr, AMC-Cu and OF-Cu, respectively, and orthotropic nonlinear-elastic properties being used for the CFC sections. The calculated temporal and spatial evolution of temperatures, stresses, and strains for the individual design variants are evaluated with special attention being paid to stress measures, plastic strains, and damage parameters indicating the risk of failure of CFC and the CFC/Cu-interface, respectively. This way the finite element analysis allows to numerically derive an optimized design variant within the framework of expected operating conditions in W7-X

Fracture toughness evaluation of elastic-plastic J-integral for high temperature components of gas turbine in power plants

International Nuclear Information System (INIS)

Chung, Nam Yong; Kim, Moon Young; Kim, Jong Woo

1999-01-01

In the study, the analysis of elastic-plastic J-integral was performed in high temperature components for gas turbine based on elastic-plastic fracture mechanics. It had been operated on the range of about 700 deg C and degraded by high temperature. It was tested for material properties of used component because of material properties changing at high temperature condition. The elastic-plastic fracture mechanics parameter, J is obtained with finite element method. A method is suggested which determines J Ic applying analysis of elastic-plastic finite element method and results of experimental load-displacements with CT specimen. It is also investigated that J-integral is applied for the elastic-plastic analysis in high temperature components. The elastic-plastic fracture toughness. J Ic determined by finite element was obtained with high accuracy using the experimental method.=20

The PLC-based Industrial Temperature Control System: Design and Implementation

Directory of Open Access Journals (Sweden)

Wei Fanjie

2017-01-01

Full Text Available Targeting at the problem of slow response and low accuracy of the automatic temperature control system for material processing and boiler heating, a new design method is proposed to work with the PLC-based temperature control system, where the box temperature control may be achieved through the fan and the heating plate. The hardware design and software design of the system are analyzed in detail. In this paper, a combination of the traditional PID control and the more popular fuzzy control is taken as the control program to achieve the overall design of the control algorithm. Followed by the simulation in the MATLAB software, the designed system is highlighted by its the characteristics of impressive stability, precision and robustness.

Advanced targeted monitoring of high temperature components in power plants

Energy Technology Data Exchange (ETDEWEB)

Roos, E; Maile, K; Jovanovic, A [MPA Stuttgart (Germany)

1999-12-31

The article presents the idea of targeted monitoring of high-temperature pressurized components in fossil-fueled power plants, implemented within a modular software system and using, in addition to pressure and temperature data, also displacement and strain measurement data. The concept has been implemented as a part of a more complex company-oriented Internet/Intranet system of MPA Stuttgart (ALIAS). ALIAS enables to combine smoothly the monitoring results with those of the off-line analysis, e. g. sensitivity analyses, comparison with preceding experience (case studies), literature search, search in material databases -(experimental and standard data), nonlinear FE-analysis, etc. The concept and the system have been implemented in real plant conditions several power plants in Germany and Europe: one of these applications and its results are described more in detail in the presentation. (orig.) 9 refs.

Advanced targeted monitoring of high temperature components in power plants

Energy Technology Data Exchange (ETDEWEB)

Roos, E.; Maile, K.; Jovanovic, A. [MPA Stuttgart (Germany)

1998-12-31

The article presents the idea of targeted monitoring of high-temperature pressurized components in fossil-fueled power plants, implemented within a modular software system and using, in addition to pressure and temperature data, also displacement and strain measurement data. The concept has been implemented as a part of a more complex company-oriented Internet/Intranet system of MPA Stuttgart (ALIAS). ALIAS enables to combine smoothly the monitoring results with those of the off-line analysis, e. g. sensitivity analyses, comparison with preceding experience (case studies), literature search, search in material databases -(experimental and standard data), nonlinear FE-analysis, etc. The concept and the system have been implemented in real plant conditions several power plants in Germany and Europe: one of these applications and its results are described more in detail in the presentation. (orig.) 9 refs.

Design of stirling engine operating at low temperature difference

Directory of Open Access Journals (Sweden)

Sedlák Josef

2018-01-01

Full Text Available There are many sources of free energy available in the form of heat that is often simply wasted. The aim of this paper is to design and build a low temperature differential Stirling engine that would be powered exclusively from heat sources such as waste hot water or focused solar rays. A prototype is limited to a low temperature differential modification because of a choice of ABSplus plastic as a construction material for its key parts. The paper is divided into two parts. The first part covers a brief history of Stirling engine and its applications nowadays. Moreover, it describes basic principles of its operation that are supplemented by thermodynamic relations. Furthermore, an analysis of applied Fused Deposition Modelling has been done since the parts with more complex geometry had been manufactured using this additive technology. The second (experimental part covers 4 essential steps of a rapid prototyping method - Computer Aided Design of the 3D model of Stirling engine using parametric modeller Autodesk Inventor, production of its components using 3D printer uPrint, assembly and final testing. Special attention was devoted to last two steps of the process since the surfaces of the printed parts were sandpapered and sprayed. Parts, where an ABS plus plastic would have impeded the correct function, had been manufactured from aluminium and brass by cutting operations. Remaining parts had been bought in a hardware store as it would be uneconomical and unreasonable to manufacture them. Last two chapters of the paper describe final testing, mention the problems that appeared during its production and propose new approaches that could be used in the future to improve the project.

DEVELOPMENT OF METHODOLOGY FOR DESIGNING TESTABLE COMPONENT STRUCTURE OF DISCIPLINARY COMPETENCE

Directory of Open Access Journals (Sweden)

Vladimir I. Freyman

2014-01-01

Full Text Available The aim of the study is to present new methods of quality results assessment of the education corresponding to requirements of Federal State Educational Standards (FSES of the Third Generation developed for the higher school. The urgency of search of adequate tools for quality competency measurement and its elements formed in the course of experts’ preparation are specified. Methods. It is necessary to consider interference of competency components such as knowledge, abilities, possession in order to make procedures of assessment of students’ achievements within the limits of separate discipline or curriculum section more convenient, effective and exact. While modeling of component structure of the disciplinary competence the testable design of components is used; the approach borrowed from technical diagnostics. Results. The research outcomes include the definition and analysis of general iterative methodology for testable designing component structure of the disciplinary competence. Application of the proposed methodology is illustrated as the example of an abstract academic discipline with specified data and index of labour requirement. Methodology restrictions are noted; practical recommendations are given. Scientific novelty. Basic data and a detailed step-by-step implementation phase of the proposed common iterative approach to the development of disciplinary competence testable component structure are considered. Tests and diagnostic tables for different options of designing are proposed. Practical significance. The research findings can help promoting learning efficiency increase, a choice of adequate control devices, accuracy of assessment, and also efficient use of personnel, temporal and material resources of higher education institutions. Proposed algorithms, methods and approaches to procedure of control results organization and realization of developed competences and its components can be used as methodical base while

The impact of component performance on the overall cycle performance of small-scale low temperature organic Rankine cycles

Science.gov (United States)

White, M.; Sayma, A. I.

2015-08-01

Low temperature organic Rankine cycles offer a promising technology for the generation of power from low temperature heat sources. Small-scale systems (∼10kW) are of significant interest, however there is a current lack of commercially viable expanders. For a potential expander to be economically viable for small-scale applications it is reasonable to assume that the same expander must have the ability to be implemented within a number of different ORC applications. It is therefore important to design and optimise the cycle considering the component performance, most notably the expander, both at different thermodynamic conditions, and using alternative organic fluids. This paper demonstrates a novel modelling methodology that combines a previously generated turbine performance map with cycle analysis to establish at what heat source conditions optimal system performance can be achieved using an existing turbine design. The results obtained show that the same turbine can be effectively utilised within a number of different ORC applications by changing the working fluid. By selecting suitable working fluids, this turbine can be used to convert pressurised hot water at temperatures between 360K and 400K, and mass flow rates between 0.45kg/s and 2.7kg/s, into useful power with outputs between 1.5kW and 27kW. This is a significant result since it allows the same turbine to be implemented into a variety of applications, improving the economy of scale. This work has also confirmed the suitability of the candidate turbine for a range of low temperature ORC applications.

Development of very high temperature reactor design technology

International Nuclear Information System (INIS)

Lee, Won Jae; Noh, Jan Man

2012-04-01

or an efficient production of nuclear hydrogen, the VHTR (Very High Temperature Gas-cooled Reactor) of 950 .deg. C outlet temperature and the interfacing system for the hydrogen production are required. We have developed various evaluation technologies for the performance and safety of VHTR through the accomplishment of this project. First, to evaluate the performance of VHTR, a series of analyses has been performed such as core characteristics at 950 .deg. C, applicability of cooled-vessel, intermediate loop system and high temperature structural integrity. Through the analyses of major accidents such as HPCC and LPCC and the analysis of the risk/performance-informed method, VHTR safety evaluation has been also performed. In addition, various design analysis codes have been developed for a nuclear design, system loop design, system performance analysis, fission product/tritium transport analysis, core thermo-fluid analysis, system layout analysis, graphite structure seismic analysis and hydrogen exposion analysis, and they are being verified and validated through a lot of international collaborations

Design and assembly of a torsion pendulum for the measurement of internal friction at low temperatures

International Nuclear Information System (INIS)

San Juan, J. M.; Gallego, I.; No, M. L.

2001-01-01

In this work we describe the assembly, operation and specifications of an inverted torsion pendulum designed to measure internal friction at low temperatures (from 4.2 K to 500 K). The high precision mechanics allow us to obtain internal friction spectra with low levels of noise from amplitudes as small as 2x10''7. The inertia components of the pendulum have been built with specific materials, so that the resonance frequency of the pendulum can be changed within two orders of magnitude (0.1-10Hz). In addition, the sample can be in situ deformed at any temperature and can be inserted into the pendulum at liquid nitrogen temperature. The operation of the pendulum, all the control p recesses and data acquisition are completely automated. (Author) 4 refs

A conceptual design of main components sizing for UMT PHEV powertrain

Science.gov (United States)

Haezah, M. N.; Norbakyah, J. S.; Atiq, W. H.; Salisa, A. R.

2015-12-01

This paper presents a conceptual design of main components sizing for Universiti Malaysia Terengganu plug-in hybrid electric vehicle (UMT PHEV) powertrain. In the design of hybrid vehicles, it is important to identify a proper component sizes. Component sizing significantly affects vehicle performance, fuel economy and emissions. The proposed UMT PHEV has only one electric machine (EM) which functions as either a motor or generator at a time and using batteries and ultracapacitors as an energy storage system (ESS). In this work, firstly, energy and power requirements based on parameters, specifications and performance requirements of vehicle are calculated. Then, the parameters for internal combustion engine, EM and ESS are selected based on the developed Kuala Terengganu drive cycle. The results obtained from this analysis are within reasonable range and satisfactory.

RCC-M - Design and Conception Rules for Mechanical Components of PWR Nuclear Islands

International Nuclear Information System (INIS)

2007-01-01

The design and construction rules applicable to mechanical components of PWR Nuclear Islands (RCC-M) are a part of the collection of design and construction rules for nuclear power plants. It covers the rules applicable to the design and manufacture of pressure boundaries of mechanical equipment of pressurized water reactors (PWR). The pressure components subject to the RCC-M are specified in A 4000. They include the reactor fluid systems (primary, secondary and auxiliary systems) and other components which are not subject to pressure: vessel internals, supports for pressure components subject to the RCC-M, nuclear island storage tanks. When a pressure equipment is subject to the RCC-M, all its elements subject to pressure are also, in accordance with the provisions of A 4000, and these elements are the same class as the component. In this case all the provisions of the RCC-M are applicable: design, procurement, manufacture, inspection and pressure testing. Elements which are not subject to pressure and which are subject to the RCC-M may be covered within the Code by limited specific provisions (procurement of materials for example). The other rules applicable to this equipment must be in contractual form. The assemblies comprising pressure equipment assembled by a manufacturer to constitute an integrated and functional whole, shall be subject to the rules indicated in this Code. Main objectives of Code Requirements are to ensure the integrity and mechanical stability over the equipment design life. Function ability and operability of equipment are not directly addressed in the Code. The RCC-M contributes to ensuring compliance with regulatory requirements. These requirements depend on the applicable regulatory context. The RCC-M is representative of the state of the art as concerns the design and manufacture of PWR components, ensuring an overall safety level tested through experience. The RCC-M consists of five sections, which provide rules for the design and

RCC-M: Design and construction rules for mechanical components of PWR nuclear islands

International Nuclear Information System (INIS)

2017-01-01

AFCEN's RCC-M code concerns the mechanical components designed and manufactured for pressurized water reactors (PWR). It applies to pressure equipment in nuclear islands in safety classes 1, 2 and 3, and certain non-pressure components, such as vessel internals, supporting structures for safety class components, storage tanks and containment penetrations. RCC-M covers the following technical subjects: sizing and design, choice of materials and procurement. Fabrication and control, including: associated qualification requirements (procedures, welders and operators, etc.), control methods to be implemented, acceptance criteria for detected defects, documentation associated with the different activities covered, and quality assurance. The design, manufacture and inspection rules defined in RCC-M leverage the results of the research and development work pioneered in France, Europe and worldwide, and which have been successfully used by industry to design and build PWR nuclear islands. AFCEN's rules incorporate the resulting feedback. Use: France's last 16 nuclear units (P'4 and N4); 4 CP1 reactors in South Africa (2) and Korea (2); 44 M310 (4), CPR-1000 (28), CPR-600 (6), HPR-1000 (4) and EPR (2) reactors in service or undergoing construction in China; 4 EPR reactors in Europe: Finland (1), France (1) and UK (2). Content: Section I - nuclear island components, subsection 'A': general rules, subsection 'B': class 1 components, subsection 'C': class 2 components, subsection 'D': class 3 components, subsection 'E': small components, subsection 'G': core support structures, subsection 'H': supports, subsection 'J': low pressure or atmospheric storage tanks, subsection 'P': containment penetration, subsection 'Q': qualification of active mechanical components, subsection 'Z': technical appendices; section II - materials; section III - examination

Evaluating comfort with varying temperatures: a graphic design tool

Energy Technology Data Exchange (ETDEWEB)

Evans, J.M. [Research Centre Habitat and Energy, Faculty of Architecture, Design and Urbanism, University of Buenos Aires, Ciudad Universitaria (Argentina)

2002-07-01

This paper considers the need to define comfort of indoor and outdoor spaces in relation to the daily variations of temperature. A graphical tool is presented, which indicates the daily swings of temperature, shown as a single point on a graph representing the average temperature and the maximum temperature swing. This point can be compared with the comfort zones for different activity levels, such as sedentary activity, sleeping, indoor and outdoor circulation according to the design proposals for different spaces. The graph allows the representation of climatic variables, the definition of comfort zones, the selection of bio climatic design resources and the evaluation of indoor temperatures, measured in actual buildings or obtained from computer simulations. The development of the graph is explained and examples given with special emphasis on the use of thermal mass. (author)

Method and alloys for fabricating wrought components for high-temperature gas-cooled reactors

International Nuclear Information System (INIS)

Thompson, L.D.; Johnson, W.R.

1983-01-01

Wrought, nickel-based alloys, suitable for components of a high-temperature gas-cooled reactor exhibit strength and excellent resistance to carburization at elevated temperatures and include aluminum and titanium in amounts and ratios to promote the growth of carburization resistant films while preserving the wrought character of the alloys. These alloys also include substantial amounts of molybdenum and/or tungsten as solid-solution strengtheners. Chromium may be included in concentrations less than 10% to assist in fabrication. Minor amounts of carbon and one or more carbide-forming metals also contribute to high-temperature strength. The range of compositions of these alloys is given. (author)

Design, construction, qualification and reliability of main components, from the safety aspect

International Nuclear Information System (INIS)

Crette, J.P.

1982-01-01

In FRANCE, the design and construction of reliable components, which condition the safe operation and availability of breeder plants, is based on the experience acquired during the operation of RAPSODIE, PHENIX and the various test facilities. The technical progress achieved on all main components is illustrated by examples taken from the CREYS-MALVILLE plant. In parallel with the development of these components, an extensive program covering research, development and the definition of design, construction and inspection rules, together with scheduling and quality assurance methods, prepares the industrialization of this reactor system, in compliance with the rules and recommendations issued by the pertinent safety authorities

Computers as components principles of embedded computing system design

CERN Document Server

Wolf, Marilyn

2012-01-01

Computers as Components: Principles of Embedded Computing System Design, 3e, presents essential knowledge on embedded systems technology and techniques. Updated for today's embedded systems design methods, this edition features new examples including digital signal processing, multimedia, and cyber-physical systems. Author Marilyn Wolf covers the latest processors from Texas Instruments, ARM, and Microchip Technology plus software, operating systems, networks, consumer devices, and more. Like the previous editions, this textbook: Uses real processors to demonstrate both technology and tec

Survey on Cooled-Vessel Designs in High Temperature Gas-Cooled Reactors

International Nuclear Information System (INIS)

Kim, Min-Hwan; Lee, Won-Jae

2006-01-01

The core outlet temperature of the coolant in the high temperature gas-cooled reactors (HTGR) has been increased to improve the overall efficiency of their electricity generation by using the Brayton cycle or their nuclear hydrogen production by using thermo-chemical processes. The increase of the outlet temperature accompanies an increase of the coolant inlet temperature. A high coolant inlet temperature results in an increase of the reactor pressure vessel (RPV) operation temperature. The conventional steels, proven vessel material in light water reactors, cannot be used as materials for the RPV in the elevated temperatures which necessitate its design to account for the creep effects. Some ferritic or martensitic steels like 2 1/4Cr-1Mo and 9Cr-1Mo-V are very well established creep resistant materials for a temperature range of 400 to 550 C. Although these materials have been used in a chemical plant, there is limited experience with using these materials in nuclear reactors. Even though the 2 1/4Cr-1Mo steel was used to manufacture the RPV for HTR-10 of Japan Atomic Energy Agency(JAEA), a large RPV has not been manufactured by using this material or 9Cr-1Mo-V steel. Due to not only its difficulties in manufacturing but also its high cost, the JAEA determined that they would exclude these materials from the GTHTR design. For the above reasons, KAERI has been considering a cooled-vessel design as an option for the RPV design of a NHDD plant (Nuclear Hydrogen Development and Demonstration). In this study, we surveyed several HTGRs, which adopt the cooled-vessel concept for their RPV design, and discussed their design characteristics. The survey results in design considerations for the NHDD cooled-vessel design

Design and development of gas turbine high temperature reactor 300

International Nuclear Information System (INIS)

Kunitomi, Kazuhiko; Katanishi, Shoji; Takada, Shoji; Yan, Xing; Takizuka, Takakazu

2003-01-01

JAERI (Japan Atomic Energy Research Institute) has been designing a Japan's original gas turbine high temperature reactor, GTHTR300 (Gas Turbine High Temperature Reactor 300). The greatly simplified design based on salient features of the HTGR (High Temperature Gas-cooled reactor) with a closed helium gas turbine enables the GTHTR300 a high efficient and economically competitive reactor to be deployed in early 2010s. Also, the GTHTR300 fully taking advantage of various experiences accumulated in design, construction and operation of the HTTR (High Temperature Engineering Test Reactor) and fossil gas turbine systems reduces technological development concerning a reactor system and electric generation system. Original features of this system are core design with two-year refueling interval, conventional steel material usage for a reactor pressure vessel, innovative plant flow scheme and horizontally installed gas turbine unit. Due to these salient features, the capital cost of the GTHTR300 is less than a target cost of 200 thousands Yen/kWe, and the electric generation cost is close to a target cost of 4 Yen/kWh. This paper describes the original design features focusing on reactor core design, fuel design, in-core structure design and reactor pressure vessel design except PCU design. Also, R and D for developing the power conversion unit is briefly described. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

Software Engineering Environment for Component-based Design of Embedded Software

DEFF Research Database (Denmark)

Guo, Yu

2010-01-01

as well as application models in a computer-aided software engineering environment. Furthermore, component models have been realized following carefully developed design patterns, which provide for an efficient and reusable implementation. The components have been ultimately implemented as prefabricated...... executable objects that can be linked together into an executable application. The development of embedded software using the COMDES framework is supported by the associated integrated engineering environment consisting of a number of tools, which support basic functionalities, such as system modelling......, validation, and executable code generation for specific hardware platforms. Developing such an environment and the associated tools is a highly complex engineering task. Therefore, this thesis has investigated key design issues and analysed existing platforms supporting model-driven software development...

Design of reactor components (non replaceable) of 500 MWe PHWR for enhanced life

International Nuclear Information System (INIS)

Dwivedi, K.P.; Seth, V.K.

1994-01-01

A nuclear power station is characterised by large initial cost and low operating cost. So a plant which is capable of operating for a longer period of time will be economically more attractive. In the past approach had been to design a nuclear power plant for 30 to 40 years of life time. However, with the improvement in technology and incorporation of redundant and diverse safety features it is now possible to design a nuclear power plant for longer life. Now internationally it is being realised that without sacrificing safety features, plant life should be extended till the cost of maintenance or refurbishment is larger than the cost of the replacement capacity. In order to meet the objective of long life, for the components which cannot be easily replaced the life time of about 100 years is being considered as the design objective. For other items replacement, layout space, shielding, access route and lifting capacity and component design are receiving additional emphasis so as to provide a long total station life time. With the above background, design improvements to enhance the life of reactor components for 500 MWe PHWR namely calandria, end shields and calandria vault liners which cannot be replaced and on which any repair is extremely difficult, have been made. This paper deals with design life of these components and the modifications incorporated in the design. (author). 3 refs., 2 tabs., 3 figs

Material and component progress within ARCHER for advanced high temperature reactor

International Nuclear Information System (INIS)

Buckthorpe, D.E.; Davies, M.; Pra, F.; Bonnamy, P.; Fokkens, J.; Heijna, M.; Bout, N. de; Vreeling, A.; Bourlier, F.; Lhachemi, D.; Woayehune, A.; Dubiez-le-Goff, S.; Hahner, P.; Futterer, M.; Berka, J.; Kalivodora, J.; Pouchon, M.A.; Schmitt, R.; Homerin, P.; Marsden, B.; Mummery, P.; Mutch, G.; Ponca, D.; Buhl, P.; Hoffmann, M.; Rondet, F.; Pecherty, A.; Baurand, F.; Alenda, F.; Esch, M.; Kohlz, N.; Reed, J.; Fachinger, J.; Klower, Dr.

2014-01-01

The ARCHER (Advanced High-Temperature Reactors for Cogeneration of Heat and Electricity R and D) integrated project started in 2011 as part of the European Commission 7. Framework Programme (FP7) for a period of four years to perform High Temperature Reactor technology R and D in support of reactor demonstration. The project consortium encompasses conventional and Nuclear Industry, Utilities, Technical Support Organizations, Research and Development Organizations and Academia. The activities involved contribute to the Generation IV (GIF) International Forum and collaborate with related projects in the US, China, Japan, and the Republic of Korea in cooperation with IAEA and ISTC. This paper addresses the progress of the work on materials and component technologies within ARCHER over the first two years of the project. (authors)

Advanced BWR core component designs and the implications for SFD analysis

International Nuclear Information System (INIS)

Ott, L.J.

1997-01-01

Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories in 1986, no experimental data base existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper will present the lessons learned from the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core models in the current generation of SFD codes. The DF-4 and CORA BWR test assemblies were modeled on the core component designs circa 1985; that is, the 8 x 8 fuel assembly with two water rods and a cruciform control blade constructed of B 4 C-filled tubelets. Within the past ten years, the state-of-the-art with respect to BWR core component development has out-distanced the current SFD experimental data base and SFD code capabilities. For example, modern BWR control blade design includes hafnium at the tips and top of each control blade wing for longer blade operating lifetimes; also water rods have been replaced by larger water channels for better neutronics economy; and fuel assemblies now contain partial-length fuel rods, again for better neutronics economy. This paper will also discuss the implications of these advanced fuel assembly and core component designs on severe accident progression and on the current SFD code capabilities

Utilization of Minor Actinides as a Fuel Component for Ultra-Long Life VHTR Configurations: Designs, Advantages and Limitations

International Nuclear Information System (INIS)

Tsvetkov, Pavel V.

2009-01-01

This project assessed the advantages and limitations of using minor actinides as a fuel component to achieve ultra-long life Very High Temperature Reactor (VHTR) configurations. Researchers considered and compared the capabilities of pebble-bed and prismatic core designs with advanced actinide fuels to achieve ultra-long operation without refueling. Since both core designs permit flexibility in component configuration, fuel utilization, and fuel management, it is possible to improve fissile properties of minor actinides by neutron spectrum shifting through configuration adjustments. The project studied advanced actinide fuels, which could reduce the long-term radio-toxicity and heat load of high-level waste sent to a geologic repository and enable recovery of the energy contained in spent fuel. The ultra-long core life autonomous approach may reduce the technical need for additional repositories and is capable to improve marketability of the Generation IV VHTR by allowing worldwide deployment, including remote regions and regions with limited industrial resources. Utilization of minor actinides in nuclear reactors facilitates developments of new fuel cycles towards sustainable nuclear energy scenarios.

Utilization of Minor Actinides as a Fuel Component for Ultra-Long Life Bhr Configurations: Designs, Advantages and Limitations

Energy Technology Data Exchange (ETDEWEB)

Dr. Pavel V. Tsvetkov

2009-05-20

This project assessed the advantages and limitations of using minor actinides as a fuel component to achieve ultra-long life Very High Temperature Reactor (VHTR) configurations. Researchers considered and compared the capabilities of pebble-bed and prismatic core designs with advanced actinide fuels to achieve ultra-long operation without refueling. Since both core designs permit flexibility in component configuration, fuel utilization, and fuel management, it is possible to improve fissile properties of minor actinides by neutron spectrum shifting through configuration adjustments. The project studied advanced actinide fuels, which could reduce the long-term radio-toxicity and heat load of high-level waste sent to a geologic repository and enable recovery of the energy contained in spent fuel. The ultra-long core life autonomous approach may reduce the technical need for additional repositories and is capable to improve marketability of the Generation IV VHTR by allowing worldwide deployment, including remote regions and regions with limited industrial resources. Utilization of minor actinides in nuclear reactors facilitates developments of new fuel cycles towards sustainable nuclear energy scenarios.

Design considerations for CRBRP heat transport system piping operating at elevated temperatures

International Nuclear Information System (INIS)

Pollono, L.P.; Mello, R.M.

1979-01-01

The heat transport system sodium piping for the Clinch River Breeder Reactor Plant (CRBRP) within the reactor containment building must withstand high temperatures for long periods of time. Each phase of the mechanical design process of the piping system is influenced by elevated temperature considerations which include material thermal creep effects, ratchetting caused by rapid temperature transients and stress relaxation, and material degradation effects. The structural design philosophy taken to design the CRBRP piping operating in a high temperature environment is described. The resulting design of the heat transport system piping is presented along with a discussion of special features that resulted from the elevated temperature considerations

Friction and wear studies of nuclear power plant components in pressurized high temperature water environments

International Nuclear Information System (INIS)

Ko, P.L.; Zbinden, M.; Taponat, M.C.; Robertson, M.F.

1997-01-01

The present paper is part of a series of papers aiming to present the friction and wear results of a collaborative study on nuclear power plant components tested in pressurized high temperature water. The high temperature test facilities and the methodology in presenting the kinetics and wear results are described in detail. The results of the same material combinations obtained from two very different high temperature test facilities (NRCC and EDF) are presented and discussed. (K.A.)

Creep fatigue assessment for EUROFER components

Energy Technology Data Exchange (ETDEWEB)

Özkan, Furkan, E-mail: oezkan.furkan@partner.kit.edu; Aktaa, Jarir

2015-11-15

Highlights: • Design rules for creep fatigue assessment are developed to EUROFER components. • Creep fatigue assessment tool is developed in FORTRAN code with coupling MAPDL. • Durability of the HCPB-TBM design is discussed under typical fusion reactor loads. - Abstract: Creep-fatigue of test blanket module (TBM) components built from EUROFER is evaluated based on the elastic analysis approach in ASME Boiler Pressure Vessel Code (BPVC). The required allowable number of cycles design fatigue curve and stress-to-rupture curve to estimate the creep-fatigue damage are used from the literature. Local stress, strain and temperature inputs for the analysis of creep-fatigue damage are delivered by the finite element code ANSYS utilizing the Mechanical ANSYS Parametric Design Language (MAPDL). A developed external FORTRAN code used as a post processor is coupled with MAPDL. Influences of different pulse durations (hold-times) and irradiation on creep-fatigue damage for the preliminary design of the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) are discussed for the First Wall component of the TBM box.

ATR-IR study of skin components: Lipids, proteins and water. Part I: Temperature effect

Science.gov (United States)

Olsztyńska-Janus, S.; Pietruszka, A.; Kiełbowicz, Z.; Czarnecki, M. A.

2018-01-01

In this work we report the studies of the effect of temperature on skin components, such as lipids, proteins and water. Modifications of lipids structure induced by increasing temperature (from 20 to 90 °C) have been studied using ATR-IR (Attenuated Total Reflectance Infrared) spectroscopy, which is a powerful tool for characterization of the molecular structure and properties of tissues, such as skin. Due to the small depth of penetration (0.6-5.6 μm), ATR-IR spectroscopy probes only the outermost layer of the skin, i.e. the stratum corneum (SC). The assignment of main spectral features of skin components allows for the determination of phase transitions from the temperature dependencies of band intensities [e.g. νas(CH2) and νs(CH2)]. The phase transitions were determined by using two methods: the first one was based on the first derivative of the Boltzmann function and the second one employed tangent lines of sigmoidal, aforementioned dependencies. The phase transitions in lipids were correlated with modifications of the structure of water and proteins.

A CANDU designed for more tolerance to failures in large components

International Nuclear Information System (INIS)

Spinks, N.J.; Barclay, F.W.; Allen, P.J.; Yee, F.

1988-06-01

Current designs of CANDU reactors have several groups of fuel channels each served by an upstream coolant supply-train consisting of an outlet header, a steam generator, one or more pumps in parallel and an inlet header. Postulated failures in these large components put the heaviest demands on the safety systems. For example, the rupture of a header sets the requirements for the speed of shutdown and for the speed and capacity of emergency coolant injection, and it has a large impact on containment design. A CANDU design is being investigated to reduce the impact of failures in large components. Each group of fuel channels is supplied by more than one train so that if one train fails the rest continue to work. Reverse flow limiters reduce the loss-of-coolant from the unbroken trains to a broken supply train. The paper describes several design options for making the piping connections from multi supply-trains to fuel channels. It discusses progress in design and testing of flow limiters. A preliminary analysis is given of affected accidents

Reliability analysis of component-level redundant topologies for solid-state fault current limiter

Science.gov (United States)

Farhadi, Masoud; Abapour, Mehdi; Mohammadi-Ivatloo, Behnam

2018-04-01

Experience shows that semiconductor switches in power electronics systems are the most vulnerable components. One of the most common ways to solve this reliability challenge is component-level redundant design. There are four possible configurations for the redundant design in component level. This article presents a comparative reliability analysis between different component-level redundant designs for solid-state fault current limiter. The aim of the proposed analysis is to determine the more reliable component-level redundant configuration. The mean time to failure (MTTF) is used as the reliability parameter. Considering both fault types (open circuit and short circuit), the MTTFs of different configurations are calculated. It is demonstrated that more reliable configuration depends on the junction temperature of the semiconductor switches in the steady state. That junction temperature is a function of (i) ambient temperature, (ii) power loss of the semiconductor switch and (iii) thermal resistance of heat sink. Also, results' sensitivity to each parameter is investigated. The results show that in different conditions, various configurations have higher reliability. The experimental results are presented to clarify the theory and feasibility of the proposed approaches. At last, levelised costs of different configurations are analysed for a fair comparison.

Design, Qualification and Integration Testing of the High-Temperature Resistance Temperature Device for Stirling Power System

Science.gov (United States)

Chan, Jack; Hill, Dennis H.; Elisii, Remo; White, Jonathan R.; Lewandowski, Edward J.; Oriti, Salvatore M.

2015-01-01

The Advanced Stirling Radioisotope Generator (ASRG), developed from 2006 to 2013 under the joint sponsorship of the United States Department of Energy (DOE) and National Aeronautics and Space Administration (NASA) to provide a high-efficiency power system for future deep space missions, employed Sunpower Incorporated's Advanced Stirling Convertors (ASCs) with operating temperature up to 840 C. High-temperature operation was made possible by advanced heater head materials developed to increase reliability and thermal-to-mechanical conversion efficiency. During a mission, it is desirable to monitor the Stirling hot-end temperature as a measure of convertor health status and assist in making appropriate operating parameter adjustments to maintain the desired hot-end temperature as the radioisotope fuel decays. To facilitate these operations, a Resistance Temperature Device (RTD) that is capable of high-temperature, continuous long-life service was designed, developed and qualified for use in the ASRG. A thermal bridge was also implemented to reduce the RTD temperature exposure while still allowing an accurate projection of the ASC hot-end temperature. NASA integrated two flight-design RTDs on the ASCs and assembled into the high-fidelity Engineering Unit, the ASRG EU2, at Glenn Research Center (GRC) for extended operation and system characterization. This paper presents the design implementation and qualification of the RTD, and its performance characteristics and calibration in the ASRG EU2 testing.

Design of the ITER Plasma-Facing Components

Energy Technology Data Exchange (ETDEWEB)

Merola, M.

2009-07-01

The ITER plasma-facing components cover an area of about 850 m{sup 2} and consist of the Divertor, the Blanket and the Test Blanket Modules (TBMs) with their corresponding frames. The Divertor is located at the bottom of the plasma chamber and is aimed at exhausting the major part of the plasma thermal power (including alpha power) and at minimizing the helium and impurity content in the plasma. It consists of 54 cassette assemblies. Each assembly has 3 plasma-facing components (PFCs), namely the inner and outer target and the dome, which are mounted onto a steel support structure, the cassette body. The targets directly intercept the magnetic field lines and are designed to withstand heat fluxes as high as 20 MW/m{sup 2}. CFC is the reference design solution for the armour of the lower part of the targets. However, the resultant high erosion rate could potentially limit machine operation in the DT phase (due to co-deposition with T). Therefore, prior to the DT phase, the divertor PFCs will be replaced with a new set entirely covered with W armour. The Divertor is a RH Class 1 component, which is planned to be replaced 3 times during the 20 years of the ITER operation. The construction phase of the ITER Divertor is being launched. The Blanket covers the largest fraction of the plasma-facing surface. Each of the 440 Blanket modules consists of a first wall (FW) panel, which is mechanically attached onto a Shield Module (SM). The design heat flux is set up to 1 or 5 MW/m{sup 2}. The FW panels are covered by Be tiles, which are joined onto a copper alloy (CuCrZr) heat sink, which is in turn intimately joined onto a 316L(N) stainless steel part. The SM is a block of 316L(N)-IG steel, where an array of cooling channels are obtained by machining and welding. The TBMs are mock-ups of DEMO breeding blankets. There are three ITER equatorial ports devoted to TBM testing, each of them allocating two TBMs, inserted in a thick steel frame. The frame is a water-cooled 316L

2D surface temperature measurement of plasma facing components with modulated active pyrometry

International Nuclear Information System (INIS)

Amiel, S.; Loarer, T.; Pocheau, C.; Roche, H.; Gauthier, E.; Aumeunier, M.-H.; Courtois, X.; Jouve, M.; Balorin, C.; Moncada, V.; Le Niliot, C.; Rigollet, F.

2014-01-01

In nuclear fusion devices, such as Tore Supra, the plasma facing components (PFC) are in carbon. Such components are exposed to very high heat flux and the surface temperature measurement is mandatory for the safety of the device and also for efficient plasma scenario development. Besides this measurement is essential to evaluate these heat fluxes for a better knowledge of the physics of plasma-wall interaction, it is also required to monitor the fatigue of PFCs. Infrared system (IR) is used to manage to measure surface temperature in real time. For carbon PFCs, the emissivity is high and known (ε ∼ 0.8), therefore the contribution of the reflected flux from environment and collected by the IR cameras can be neglected. However, the future tokamaks such as WEST and ITER will be equipped with PFCs in metal (W and Be/W, respectively) with low and variable emissivities (ε ∼ 0.1–0.4). Consequently, the reflected flux will contribute significantly in the collected flux by IR camera. The modulated active pyrometry, using a bicolor camera, proposed in this paper allows a 2D surface temperature measurement independently of the reflected fluxes and the emissivity. Experimental results with Tungsten sample are reported and compared with simultaneous measurement performed with classical pyrometry (monochromatic and bichromatic) with and without reflective flux demonstrating the efficiency of this method for surface temperature measurement independently of the reflected flux and the emissivity

Design and structural calculation of nuclear power plant mechanical components

International Nuclear Information System (INIS)

Amaral, J.A.R. do

1986-01-01

The mechanical components of a nuclear power plant must show high quality and safety due to the presence of radioactivity. Besides the perfect functioning during the rigid operating conditions, some postulated loadings are foreseen, like earthquake and loss of coolant accidents, which must be also considered in the design. In this paper, it is intended to describe the design and structural calculations concept and development, the interactions with the piping and civil designs, as well as their influences in the licensing process with the authorities. (Author) [pt

Replaceable LMFBR core components

International Nuclear Information System (INIS)

Evans, E.A.; Cunningham, G.W.

1976-01-01

Much progress has been made in understanding material and component performance in the high temperature, fast neutron environment of the LMFBR. Current data have provided strong assurance that the initial core component lifetime objectives of FFTF and CRBR can be met. At the same time, this knowledge translates directly into the need for improved core designs that utilize improved materials and advanced fuels required to meet objectives of low doubling times and extended core component lifetimes. An industrial base for the manufacture of quality core components has been developed in the US, and all procurements for the first two core equivalents for FFTF will be completed this year. However, the problem of fabricating recycled plutonium while dramatically reducing fabrication costs, minimizing personnel exposure, and protecting public health and safety must be addressed

Organizational Design Analysis of Fleet Readiness Center Southwest Components Department

National Research Council Canada - National Science Library

Montes, Jose F

2007-01-01

.... The purpose of this MBA Project is to analyze the proposed organizational design elements of the FRCSW Components Department that resulted from the integration of the Naval Aviation Depot at North Island (NADEP N.I...

Design principles and overall aspects to proof the integrity of pressurized components

International Nuclear Information System (INIS)

Roos, E.; Herter, K.-H.; Schuler, X.

2005-01-01

Technical codes and standards used for the construction, design and operation of nuclear components and systems provides the material data required, detailed stress analysis procedures and a design philosophy which guarantees a reliable behaviour of the systems, structures and components (SSC) throughout the specified life time. It is important that the design concept accounts for most possible damage mechanisms and failure modes and provides rational margins of safety against each type of damage mechanism and failure mode. The design criteria according to codes and standards are the basic rules upon which the mechanical behaviour of the SSC is based. For cyclic stress evaluation the different codes and standards provides fatigue analyses to be performed considering the various loading histories (mechanical and thermal loads) and geometric complexities of the SSC. Essentially the philosophy for the mechanical design in all of the codes and standards broadly encompasses the two approaches of Design-by-Rule (DBR) and Design-by-Analysis (DBA). Design-by-Experiment (DBE) and Design-by-Fracture Mechanics (DBFA) are in special cases additional possibilities for the design as well as for the proof of integrity of SSC. Based on the German Basis Safety Concept a general concept to ensure the integrity of pressurised components is developed. As a premise for a systematically approach it is indispensable to show that the as-built status of quality (actual material characteristics, actual as-built configurations, design, actual loading) is according to the requirements given in the guidelines and standards, to show that sufficient knowledge of possible failure mechanism (e.g. no inadmissible dynamic loading, no corrosion) is available and to show that the as-built status of quality can be guaranteed for the succeeding operation. The calculation methods and fracture mechanics approaches are verified by numerous experimental data. (authors)

Evaluation of Embedded System Component Utilized in Delivery Integrated Design Project Course

Science.gov (United States)

Junid, Syed Abdul Mutalib Al; Hussaini, Yusnira; Nazmie Osman, Fairul; Razak, Abdul Hadi Abdul; Idros, Mohd Faizul Md; Karimi Halim, Abdul

2018-03-01

This paper reports the evaluation of the embedded system component utilized in delivering the integrated electronic engineering design project course. The evaluation is conducted based on the report project submitted as to fulfil the assessment criteria for the integrated electronic engineering design project course named; engineering system design. Six projects were assessed in this evaluation. The evaluation covers the type of controller, programming language and the number of embedded component utilization as well. From the evaluation, the C-programming based language is the best solution preferred by the students which provide them flexibility in the programming. Moreover, the Analog to Digital converter is intensively used in the projects which include sensors in their proposed design. As a conclusion, in delivering the integrated design project course, the knowledge over the embedded system solution is very important since the high density of the knowledge acquired in accomplishing the project assigned.

Design evolution and integration of the ITER in-vessel components

International Nuclear Information System (INIS)

Martin, A.; Calcagno, B.; Chappuis, Ph.; Daly, E.; Dellopoulos, G.; Furmanek, A.; Gicquel, S.; Heitzenroeder, P.; Jiming, Chen; Kalish, M.; Kim, D.-H.; Khomiakov, S.; Labusov, A.; Loarte, A.; Loughlin, M.; Merola, M.; Mitteau, R.; Polunovski, E.; Raffray, R.; Sadakov, S.

2013-01-01

Highlights: ► The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. ► A set of in-vessel vertical stabilization (VS) coils and a set of in-vessel Edge Localized Mode (ELM) control coils have been implemented. ► The blanket system has been redesigned to include first wall (FW) shaping, to upgrade the FW heat removal capability and to allow for an “in situ” replacement. ► The blanket manifold system has been redesigned to improve leak detection and localisation. ► The introduction of a new set of in-vessel coils and the design evolution of the blanket system while the ITER project was entering the procurement phase have proven to be a major engineering challenge. -- Abstract: The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. A set of in-vessel vertical stabilization (VS) coils and a set of in-vessel Edge Localized Mode (ELM) control coils have been implemented. The blanket system has been redesigned to include first wall (FW) shaping, to upgrade the FW heat removal capability and to allow for an “in situ” replacement. The blanket manifold system has been redesigned to improve leak detection and localisation. The introduction of a new set of in-vessel coils and the design evolution of the blanket system while the ITER project was entering the procurement phase have proven to be a major engineering challenge. This paper describes the status of the redesign of the in-vessel components and the associated integration issues

Recommendations for fatigue design of welded joints and components

CERN Document Server

Hobbacher, A F

2016-01-01

This book provides a basis for the design and analysis of welded components that are subjected to fluctuating forces, to avoid failure by fatigue. It is also a valuable resource for those on boards or commissions who are establishing fatigue design codes. For maximum benefit, readers should already have a working knowledge of the basics of fatigue and fracture mechanics. The purpose of designing a structure taking into consideration the limit state for fatigue damage is to ensure that the performance is satisfactory during the design life and that the survival probability is acceptable. The latter is achieved by the use of appropriate partial safety factors. This document has been prepared as the result of an initiative by Commissions XIII and XV of the International Institute of Welding (IIW).

Elevated service water temperature systems analysis for a nuclear power plant

International Nuclear Information System (INIS)

Lewis, T.; Hurt, W.

1992-01-01

This paper describes analyses performed to support the evaluation of the effects of elevated Service Water (SW) temperatures on the operation of a Pressurized Water Reactor. The purpose of the analyses is to provide justification of continued plant operation with SW temperatures up to 5 degrees F (3 degrees C) above the original temperature design limit. The study involved evaluation of the following major components or plant transients: Containment Design Basis Accident (DBA), Emergency Diesel Generator (EDG), Plant Cooldown, Engineered Safety Feature (ESF) Room Coolers, Engineered Safety Feature Pumps, and Assessment for Impact on Normal Operation. The principal objective was related to raising the design maximum temperature of the SW system from 95 degrees F (35 degrees C) to 100 degrees F (38 degrees C). since the Service Water system is safety related, an serves a plant during both normal and design basis conditions, a wide variety of components must be analyzed under various operating modes. The evaluation of systems and components affected by elevated SW temperature is presented, along with conclusions

Design Basis of Core Components and their Realization in the frame of the EPR'sTM Core Component Development

International Nuclear Information System (INIS)

Schebitz, Florian; Mekmouche, Abdelhalim

2008-01-01

Rod Cluster Control Assemblies (RCCAs), Thimble Plug Assemblies (TPAs), Primary Neutron Sources (PNS) and Secondary Neutron Sources (SNS) are essential for the operation of a Nuclear Power Plant. Different functional requirements ask for different components and geometries. Therefore three different core components are used within the primary circuit: - The RCCA, which contains the absorber materials, is used to regulate and shut down the nuclear chain reaction. Under these demanding conditions different effects are determining the lifetime of the RCCA and in particular of the control rods. Several improvements like ion-nitriding of the cladding, lengthening of the bottom end plug, helium backfilling and reduction of the absorber diameter in the bottom part, which have already been introduced with the HARMONI TM RCCA, show a real improvement in terms of lifetime. - The TPAs are used at positions without RCCAs and neutron sources to limit the by-pass flow-rate in the fuel assembly guide tubes. The advanced TPA design results from a perfect combination of French and German design experience feedback. Benefits like homogenized hydraulic flow and improved manageability in terms of handling tools show the joined experience. - The neutron sources are used to enhance the flux level when the core is sub-critical so as to facilitate the core start-up control by the neutron flux detectors. Primary and secondary neutron sources are designed in a common way with reviewed and improved methodology. As there are different ways and conditions to operate core components, several designs are available. For the EPR TM , the best methods and products have been chosen. All chosen components contribute to an optimized and safe operation of the EPR TM . (authors)

Thermocouple design for measuring temperatures of small insects

Science.gov (United States)

A.A. Hanson; R.C. Venette

2013-01-01

Contact thermocouples often are used to measure surface body temperature changes of insects during cold exposure. However, small temperature changes of minute insects can be difficult to detect, particularly during the measurement of supercooling points. We developed two thermocouple designs, which use 0.51 mm diameter or 0.127 mm diameter copper-constantan wires, to...

Analysis of optimal design of low temperature economizer

Science.gov (United States)

Song, J. H.; Wang, S.

2017-11-01

This paper has studied the Off-design characteristic of low temperature economizer system based on thermodynamics analysis. Based on the data from one 1000 MW coal-fired unit, two modes of operation are contrasted and analyzed. One is to fix exhaust gas temperature and the other one is to take into account both of the average temperature difference and the exhaust gas temperature. Meanwhile, the cause of energy saving effect change is explored. Result shows that: in mode 1, the amount of decrease in coal consumption reduces from 1.11 g/kWh (under full load) to 0.54 g/kWh (under half load), and in mode 2, when the load decreases from 90% to 50%, the decrease in coal consumption reduces from 1.29 g/kWh to 0.84 g/kWh. From the result, under high load, the energy saving effect is superior, and under lower work load, energy saving effect declines rapidly when load is reduced. When load changes, the temperature difference of heat transfer, gas flow, the flue gas heat rejection and the waste heat recovery change. The energy saving effect corresponding changes result in that the energy saving effect under high load is superior and more stable. However, rational adjustment to the temperature of outlet gas can alleviate the decline of the energy saving effect under low load. The result provides theoretical analysis data for the optimal design and operation of low temperature economizer system of power plant.

Heat experiment design to estimate temperature dependent thermal properties

International Nuclear Information System (INIS)

Romanovski, M

2008-01-01

Experimental conditions are studied to optimize transient experiments for estimating temperature dependent thermal conductivity and volumetric heat capacity. A mathematical model of a specimen is the one-dimensional heat equation with boundary conditions of the second kind. Thermal properties are assumed to vary nonlinearly with temperature. Experimental conditions refer to the thermal loading scheme, sampling times and sensor location. A numerical model of experimental configurations is studied to elicit the optimal conditions. The numerical solution of the design problem is formulated on a regularization scheme with a stabilizer minimization without a regularization parameter. An explicit design criterion is used to reveal the optimal sensor location, heating duration and flux magnitude. Results obtained indicate that even the strongly nonlinear experimental design problem admits the aggregation of its solution and has a strictly defined optimal measurement scheme. Additional region of temperature measurements with allowable identification error is revealed.

Design, maintenance and lifetime of nuclear components

International Nuclear Information System (INIS)

Noel, R.L.; Eisenhut, D.G.; Carey, J.J.; Reynes, L.J.

1989-01-01

Division D of SMiRT deals with experience feedback relating to the in-service behavior of nuclear components, the design and construction of this equipment, its maintenance and the evaluation and management of its lifetime. The nuclear industry now having reached maturity, with more than 300 units in service worldwide, these problems are now of predominant importance to the activity of the industry and in its development programs. This applies particularly to the problems relating to the lifetime of nuclear plants, problems which are rightly of such concern both to the utilities, in view of the enormous investments involved, and also to the safety authorities. These contributions have been reviewed for the purpose of analyzing the essential points. This analysis highlights the considerable advances achieved during the recent decades in design and maintenance methods and practices. It also identifies the areas in which progress still remains to be made

Designing components using smartMOVE electroactive polymer technology

Science.gov (United States)

Rosenthal, Marcus; Weaber, Chris; Polyakov, Ilya; Zarrabi, Al; Gise, Peter

2008-03-01

Designing components using SmartMOVE TM electroactive polymer technology requires an understanding of the basic operation principles and the necessary design tools for integration into actuator, sensor and energy generation applications. Artificial Muscle, Inc. is collaborating with OEMs to develop customized solutions for their applications using smartMOVE. SmartMOVE is an advanced and elegant way to obtain almost any kind of movement using dielectric elastomer electroactive polymers. Integration of this technology offers the unique capability to create highly precise and customized motion for devices and systems that require actuation. Applications of SmartMOVE include linear actuators for medical, consumer and industrial applications, such as pumps, valves, optical or haptic devices. This paper will present design guidelines for selecting a smartMOVE actuator design to match the stroke, force, power, size, speed, environmental and reliability requirements for a range of applications. Power supply and controller design and selection will also be introduced. An overview of some of the most versatile configuration options will be presented with performance comparisons. A case example will include the selection, optimization, and performance overview of a smartMOVE actuator for the cell phone camera auto-focus and proportional valve applications.

Mathematical Model-Based Temperature Preparation of Liquid-Propellant Components Cooled by Liquid Nitrogen in the Heat Exchanger with a Coolant

Directory of Open Access Journals (Sweden)

S. K. Pavlov

2014-01-01

Full Text Available Before fuelling the tanks of missiles, boosters, and spacecraft with liquid-propellant components (LPC their temperature preparation is needed. The missile-system ground equipment performs this operation during prelaunch processing of space-purpose missiles (SPM. Usually, the fuel cooling is necessary to increase its density and provide heat compensation during prelaunch operation of SPM. The fuel temperature control systems (FTCS using different principles of operation and types of coolants are applied for fuel cooling.To determine parameters of LPC cooling process through the fuel heat exchange in the heat exchanger with coolant, which is cooled by liquid nitrogen upon contact heat exchange in the coolant reservoir, a mathematical model of this process and a design technique are necessary. Both allow us to determine design parameters of the cooling system and the required liquid nitrogen reserve to cool LPC to the appropriate temperature.The article presents an overview of foreign and domestic publications on cooling processes research and implementation using cryogenic products such as liquid nitrogen. The article draws a conclusion that it is necessary to determine the parameters of LPC cooling process through the fuel heat exchange in the heat exchanger with coolant, which is liquid nitrogen-cooled upon contact heat exchange in the coolant reservoir allowing to define rational propellant cooling conditions to the specified temperature.The mathematical model describes the set task on the assumption that a heat exchange between the LPC and the coolant in the heat exchanger and with the environment through the walls of tanks and pipelines of circulation loops is quasi-stationary.The obtained curves allow us to calculate temperature changes of LPC and coolant, cooling time and liquid nitrogen consumption, depending on the process parameters such as a flow rate of liquid nitrogen, initial coolant temperature, pump characteristics, thermal

Applying dynamic mold temperature control to cosmetic package design

Directory of Open Access Journals (Sweden)

Hsiao Shih-Wen

2017-01-01

Full Text Available Owing to the fashion trend and the market needs, this study developed the eco-cushion compact. Through the product design and the advanced process technology, many issues have improved, for instance, the inconvenience of transportation, the lack of multiuse capability, the increase of costs, and the low yield rate. The eco-cushion compact developed in this study was high quality, low cost, and meets the requirements of the eco market. The study aimed at developing a reusable container. Dynamic mold temperature control was introduced in the injection modeling process. The innovation in the product was its multi-functional formula invention, eco-product design, one-piece powder case design, and multifunctional design in the big powder case, mold flow and development of dynamic mold temperature control. Finally, through 3D drawing and modeling, and computer assistance for mold flow and verification to develop and produce models. During the manufacturing process, in order to solve the problems of tightness and warping, development and manufacture of dynamic mold temperature control were introduced. This decreased the injection cycle and residual stress, and deformation of the products has reduced to less than 0.2 mm, and the air tightness increased. In addition, air leakage was less than 2% and the injection cycle decreased to at least 10%. The results of the study can be extended and applied on the future design on cosmetic package and an alternative can be proposed to solve the problems of air tightness and warping. In this study, dynamic mold temperature control is considered as a design with high price-performance ratio, which can be adopted on industrial application for practical benefit and improvement.

High-Temperature Gas-Cooled Test Reactor Point Design

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

2016-04-01

A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

Temperature Programmed Desorption of Quench-condensed Krypton and Acetone in Air; Selective Concentration of Ultra-trace Gas Components.

Science.gov (United States)

Suzuki, Taku T; Sakaguchi, Isao

2016-01-01

Selective concentration of ultra-trace components in air-like gases has an important application in analyzing volatile organic compounds in the gas. In the present study, we examined quench-condensation of the sample gas on a ZnO substrate below 50 K followed by temperature programmed desorption (TPD) (low temperature TPD) as a selective gas concentration technique. We studied two specific gases in the normal air; krypton as an inert gas and acetone as a reactive gas. We evaluated the relationship between the operating condition of low temperature TPD and the lowest detection limit. In the case of krypton, we observed the selective concentration by exposing at 6 K followed by thermal desorption at about 60 K. On the other hand, no selectivity appeared for acetone although trace acetone was successfully concentrated. This is likely due to the solvent effect by a major component in the air, which is suggested to be water. We suggest that pre-condensation to remove the water component may improve the selectivity in the trace acetone analysis by low temperature TPD.

The Effects of Temperature and Oxidation on Deuterium Retention in Solid and Liquid Lithium Films on Molybdenum Plasma-Facing Components

Science.gov (United States)

Capece, Angela

2014-10-01

Liquid metal plasma-facing components (PFCs) enable in-situ renewal of the surface, thereby offering a solution to neutron damage, erosion, and thermal fatigue experienced by solid PFCs. Lithium in particular has a high chemical affinity for hydrogen, which has resulted in reduced recycling and enhanced plasma performance on many fusion devices including TFTR, T11-M, FTU, CDX-U, LTX, TJ-II, and NSTX. A key component to the improvement in plasma performance is deuterium retention in Li; however, this process is not well understood in the complex tokamak environment. Recent surface science experiments conducted at the Princeton Plasma Physics Laboratory have used electron spectroscopy and temperature programmed desorption to understand the mechanisms for D retention in Li coatings on Mo substrates. The experiments were designed to give monolayer-control of Li films and were conducted in ultrahigh vacuum under controlled environments. An electron cyclotron resonance plasma source was used to deliver a beam of deuterium ions to the surface over a range of ion energies. Our work shows that D is retained as LiD in metallic Li films. However, when oxygen is present in the film, either by diffusion from the subsurface at high temperature or as a contaminant during the deposition process, Li oxides are formed that retain D as LiOD. Experiments indicate that LiD is more thermally stable than LiOD, which decomposes to liberate D2 gas and D2O at temperatures 100 K lower than the LiD decomposition temperature. Other experiments show how D retention varies with substrate temperature to provide insight into the differences between solid and liquid lithium films. This work was supported by DOE Contract No. DE AC02-09CH11466.

Sizes of secondary plant components for modularized IRIS balance of plant design

International Nuclear Information System (INIS)

Williamson, Martin; Townsend, Lawrence

2003-01-01

Herein we report on a conceptual design for a balance of plant (BOP) layout to coordinate with IRIS-like plants. The report consists of results of calculations that sizes of various BOP components. These calculations include the thermodynamic analyses and general sizing of the components in order to determine plant capability and plant layout for studies on modularity and transportability. Mathematical modeling of the BOP system involves a modified ORCENT2 code as well as standard heat transfer methods. Using typical values for PWR type plants, a general BOP design, and IRIS steam generator values, an ORCENT2 heat balance is carried out for the secondary side of the plant. Using the ORCENT2 output, standard heat transfer methods are then used to calculate system performance and component sizes. (author)

Magnetic superelevation design of Halbach permanent magnet guideway for high-temperature superconducting maglev

Science.gov (United States)

Lei, Wuyang; Qian, Nan; Zheng, Jun; Huang, Huan; Zhang, Ya; Deng, Zigang

2017-07-01

To improve the curve negotiating ability of high-temperature superconducting (HTS) maglev system, a special structure of magnetic superelevation for double-pole Halbach permanent magnet guideway (PMG) was designed. The most significant feature of this design is the asymmetrical PMG that forms a slanting magnetic field without affecting the smoothness of the PMG surface. When HTS maglev vehicle runs through curves with magnetic superelevation, the vehicle will slant due to asymmetry in magnetic field and the flux-pinning effect of onboard HTS bulks. At the same time, one component of the levitation force provides a part of the centripetal force that reduces lateral acceleration of the vehicle and thus enhances its curve negotiating ability. Furthermore, the slant angle of magnetic superelevation can be adjusted by changing the materials and the thickness of the added permanent magnets. This magnetic superelevation method, together with orographic uplift, can be applied to different requirements of PMG designs. Besides, the applicability of this method would benefit future development of high-speed HTS maglev system.

Bulk ultrasonic NDE of metallic components at high temperature using magnetostrictive transducers

Science.gov (United States)

Ashish, Antony Jacob; Rajagopal, Prabhu; Balasubramaniam, Krishnan; Kumar, Anish; Rao, B. Purnachandra; Jayakumar, Tammana

2017-02-01

Online ultrasonic NDE at high-temperature is of much interest to the power, process and automotive industries in view of possible savings in downtime. This paper describes a novel approach to developing ultrasonic transducers capable of high-temperature in-situ operation using the principle of magnetostriction. Preliminary design from previous research by the authors [1] is extended for operation at 1 MHz, and at elevated temperatures by amorphous metallic strips as the magnetostrictive core. Ultrasonic signals in pulse-echo mode are experimentally obtained from the ultrasonic transducer thus developed, in a simulated high-temperature environment of 350 °C for 10 hours. Advantages and challenges for practical deployment of this approach are discussed.

Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

International Nuclear Information System (INIS)

Mohanty, Subhasish; Majumdar, Saurindranath; Srinivasan, Makuteswara

2013-01-01

Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite

Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

Energy Technology Data Exchange (ETDEWEB)

Mohanty, Subhasish, E-mail: smohanty@anl.gov [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Majumdar, Saurindranath [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Srinivasan, Makuteswara [U.S. Nuclear Regulatory Commission, Washington, DC 20555 (United States)

2013-07-15

Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite.

[Design of traditional Chinese medicines with antihypertensive components based on medicinal property combination modes].

Science.gov (United States)

Liao, Su-Fen; Yan, Su-Rong; Guo, Wei-Jia; Luo, Ji; Sun, Jing; Dong, Fang; Wang, Yun; Qiao, Yan-Jiang

2014-07-01

Multi-component traditional Chinese medicines are an innovative research mode for traditional Chinese medicines. Currently, there are many design methods for developing multi-component traditional Chinese medicines, but their common feature is the lack of effective connection of the traditional Chinese medicine theory. In this paper, the authors discussed the multi-component traditional Chinese medicine design methods based on medicinal property combination modes, provided the combination methods with the characteristics of traditional Chinese medicine for the prescription combinations, and proved its feasibly with hypertension cases.

Design of temperature monitoring system based on CAN bus

Science.gov (United States)

Zhang, Li

2017-10-01

The remote temperature monitoring system based on the Controller Area Network (CAN) bus is designed to collect the multi-node remote temperature. By using the STM32F103 as main controller and multiple DS18B20s as temperature sensors, the system achieves a master-slave node data acquisition and transmission based on the CAN bus protocol. And making use of the serial port communication technology to communicate with the host computer, the system achieves the function of remote temperature storage, historical data show and the temperature waveform display.

Group-wise ANOVA simultaneous component analysis for designed omics experiments

NARCIS (Netherlands)

Saccenti, Edoardo; Smilde, Age K.; Camacho, José

2018-01-01

Introduction: Modern omics experiments pertain not only to the measurement of many variables but also follow complex experimental designs where many factors are manipulated at the same time. This data can be conveniently analyzed using multivariate tools like ANOVA-simultaneous component analysis

Seismic Design of ITER Component Cooling Water System-1 Piping

Science.gov (United States)

Singh, Aditya P.; Jadhav, Mahesh; Sharma, Lalit K.; Gupta, Dinesh K.; Patel, Nirav; Ranjan, Rakesh; Gohil, Guman; Patel, Hiren; Dangi, Jinendra; Kumar, Mohit; Kumar, A. G. A.

2017-04-01

The successful performance of ITER machine very much depends upon the effective removal of heat from the in-vessel components and other auxiliary systems during Tokamak operation. This objective will be accomplished by the design of an effective Cooling Water System (CWS). The optimized piping layout design is an important element in CWS design and is one of the major design challenges owing to the factors of large thermal expansion and seismic accelerations; considering safety, accessibility and maintainability aspects. An important sub-system of ITER CWS, Component Cooling Water System-1 (CCWS-1) has very large diameter of pipes up to DN1600 with many intersections to fulfill the process flow requirements of clients for heat removal. Pipe intersection is the weakest link in the layout due to high stress intensification factor. CCWS-1 piping up to secondary confinement isolation valves as well as in-between these isolation valves need to survive a Seismic Level-2 (SL-2) earthquake during the Tokamak operation period to ensure structural stability of the system in the Safe Shutdown Earthquake (SSE) event. This paper presents the design, qualification and optimization of layout of ITER CCWS-1 loop to withstand SSE event combined with sustained and thermal loads as per the load combinations defined by ITER and allowable limits as per ASME B31.3, This paper also highlights the Modal and Response Spectrum Analyses done to find out the natural frequency and system behavior during the seismic event.

Off-design performance analysis of Kalina cycle for low temperature geothermal source

International Nuclear Information System (INIS)

Li, Hang; Hu, Dongshuai; Wang, Mingkun; Dai, Yiping

2016-01-01

Highlights: • The off-design performance analysis of Kalina cycle is conducted. • The off-design models are established. • The genetic algorithm is used in the design phase. • The sliding pressure control strategy is applied. - Abstract: Low temperature geothermal sources with brilliant prospects have attracted more and more people’s attention. Kalina cycle system using ammonia water as working fluid could exploit geothermal energy effectively. In this paper, the quantitative analysis of off-design performance of Kalina cycle for the low temperature geothermal source is conducted. The off-design models including turbine, pump and heat exchangers are established preliminarily. Genetic algorithm is used to maximize the net power output and determine the thermodynamic parameters in the design phase. The sliding pressure control strategy applied widely in existing Rankine cycle power plants is adopted to response to the variations of geothermal source mass flow rate ratio (70–120%), geothermal source temperature (116–128 °C) and heat sink temperature (0–35 °C). In the off-design research scopes, the guidance for pump rotational speed adjustment is listed to provide some reference for off-design operation of geothermal power plants. The required adjustment rate of pump rotational speed is more sensitive to per unit geothermal source temperature than per unit heat sink temperature. Influence of the heat sink variation is greater than that of the geothermal source variation on the ranges of net power output and thermal efficiency.

Design alternatives, components key to optimum flares

International Nuclear Information System (INIS)

Cunha-Leite, O.

1992-01-01

A properly designed flare works as an emissions control system with greater than 98% combustion efficiency. The appropriate use of steam, natural gas, and air-assisted flare tips can result in smokeless combustion. Ground flare, otherwise the elevated flare is commonly chosen because it handles larger flow releases more economically. Flaring has become more complicated than just lighting up waste gas. Companies are increasingly concerned about efficiency. In addition, U.S. Occupational Safety and Health Administration (OSHA) and U.S. Environmental Protection Agency (EPA) have become more active, resulting in tighter regulations on both safety and emissions control. These regulations have resulted in higher levels of concern and involvement in safety and emissions matters, not to mention smoke, noise, glare, and odor. This first to two articles on flare design and components looks at elevated flares, flare tips, incinerator-type flares, flare pilots, and gas seals. Part 2 will examine knockout drums, liquid-seal drums, ignition systems, ground flares, vapor recovery systems, and flare noise

Cooling concepts for HTS components

International Nuclear Information System (INIS)

Binneberg, A.; Buschmann, H.; Neubert, J.

1993-01-01

HTS components require that low-cost, reliable cooling systems be used. There are no general solutions to such systems. Any cooling concept has to be tailored to the specific requirements of a system. The following has to he taken into consideration when designing cooling concepts: - cooling temperature - constancy and controllability of the cooling temperature - cooling load and refrigerating capacity - continuous or discontinuous mode - degree of automation - full serviceability or availability before evacuation -malfunctions caused by microphonic, thermal or electromagnetic effects -stationary or mobile application - investment and operating costs (orig.)

STRUCTURAL DESIGN CRITERIA FOR TARGET/BLANKET SYSTEM COMPONENT MATERIALS FOR THE ACCELERATOR PRODUCTION OF TRITIUM PROJECT

Energy Technology Data Exchange (ETDEWEB)

W. JOHNSON; R. RYDER; P. RITTENHOUSE

2001-01-01

The design of target/blanket system components for the Accelerator Production of Tritium (APT) plant is dependent on the development of materials properties data specified by the designer. These data are needed to verify that component designs are adequate. The adequacy of the data will be related to safety, performance, and economic considerations, and to other requirements that may be deemed necessary by customers and regulatory bodies. The data required may already be in existence, as in the open technical literature, or may need to be generated, as is often the case for the design of new systems operating under relatively unique conditions. The designers' starting point for design data needs is generally some form of design criteria used in conjunction with a specified set of loading conditions and associated performance requirements. Most criteria are aimed at verifying the structural adequacy of the component, and often take the form of national or international standards such as the ASME Boiler and Pressure Vessel Code (ASME B and PV Code) or the French Nuclear Structural Requirements (RCC-MR). Whether or not there are specific design data needs associated with the use of these design criteria will largely depend on the uniqueness of the conditions of operation of the component. A component designed in accordance with the ASME B and PV Code, where no unusual environmental conditions exist, will utilize well-documented, statistically-evaluated developed in conjunction with the Code, and will not be likely to have any design data needs. On the other hand, a component to be designed to operate under unique APT conditions, is likely to have significant design data needs. Such a component is also likely to require special design criteria for verification of its structural adequacy, specifically accounting for changes in materials properties which may occur during exposure in the service environment. In such a situation it is common for the design criteria

Design of Water Temperature Control System Based on Single Chip Microcomputer

Science.gov (United States)

Tan, Hanhong; Yan, Qiyan

2017-12-01

In this paper, we mainly introduce a multi-function water temperature controller designed with 51 single-chip microcomputer. This controller has automatic and manual water, set the water temperature, real-time display of water and temperature and alarm function, and has a simple structure, high reliability, low cost. The current water temperature controller on the market basically use bimetal temperature control, temperature control accuracy is low, poor reliability, a single function. With the development of microelectronics technology, monolithic microprocessor function is increasing, the price is low, in all aspects of widely used. In the water temperature controller in the application of single-chip, with a simple design, high reliability, easy to expand the advantages of the function. Is based on the appeal background, so this paper focuses on the temperature controller in the intelligent control of the discussion.

Components of the primary circuit of LWRs

International Nuclear Information System (INIS)

1980-01-01

This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673 K (400 C). The primary circuit as the pressure containment of the reactor coolant comprises: Reactor pressure vessel (without internals), steam generator (primary loop), pressurizer, reactor coolant pump housing, interconnecting pipings between the components mentioned above and appropriate various valve and instrument casings, pipings branding from the above components and interconnecting pipings, including the appropriate instrument casings, up to and including the first isolating valve, pressure shielding of control rod drives. (orig.) [de

Structural mechanics of nuclear plant components

International Nuclear Information System (INIS)

Roche, R.

1986-10-01

Sound structural analysis are needed for designing safe and reliable components, hence his play is very important in nuclear industry. This report is a provisional writing on the good practice in structural mechanics. Emphasis is put on non elastic analysis, damage appraisal, fatigue, fracture mechanics and also on elevated temperature behaviour [fr

Thermal Analysis of Iodine Satellite (iSAT) from Preliminary Design Review (PDR) to Critical Design Review (CDR)

Science.gov (United States)

Mauro, Stephanie

2016-01-01

The Iodine Satellite (iSAT) is a 12U cubesat with a primary mission to demonstrate the iodine fueled Hall Effect Thruster (HET) propulsion system. The spacecraft (SC) will operate throughout a one year mission in an effort to mature the propulsion system for use in future applications. The benefit of the HET is that it uses a propellant, iodine, which is easy to store and provides a high thrust-to-mass ratio. This paper will describe the thermal analysis and design of the SC between Preliminary Design Review (PDR) and Critical Design Review (CDR). The design of the satellite has undergone many changes due to a variety of challenges, both before PDR and during the time period discussed in this paper. Thermal challenges associated with the system include a high power density, small amounts of available radiative surface area, localized temperature requirements of the propulsion components, and unknown orbital parameters. The thermal control system is implemented to maintain component temperatures within their respective operational limits throughout the mission, while also maintaining propulsion components at the high temperatures needed to allow gaseous iodine propellant to flow. The design includes heaters, insulation, radiators, coatings, and thermal straps. Currently, the maximum temperatures for several components are near to their maximum operation limit, and the battery is close to its minimum operation limit. Mitigation strategies and planned work to solve these challenges will be discussed.

Using containment analysis to improve component cooling water heat exchanger limits

International Nuclear Information System (INIS)

Da Silva, H.C.; Tajbakhsh, A.

1995-01-01

The Comanche Peak Steam Electric Station design requires that exit temperatures from the Component Cooling Water Heat Exchanger remain below 330.37 K during the Emergency Core Cooling System recirculation stage, following a hypothetical Loss of Coolant Accident (LOCA). Due to measurements indicating a higher than expected combination of: (a) high fouling factor in the Component Cooling Water Heat Exchanger with (b) high ultimate heat sink temperatures, that might lead to temperatures in excess of the 330.37 K limit, if a LOCA were to occur, TUElectric adjusted key flow rates in the Component Cooling Water network. This solution could only be implemented with improvements to the containment analysis methodology of record. The new method builds upon the CONTEMPT-LT/028 code by: (a) coupling the long term post-LOCA thermohydraulics with a more detailed analytical model for the complex Component Cooling Water Heat Exchanger network and (b) changing the way mass and energy releases are calculated after core reflood and steam generator energy is dumped to the containment. In addition, a simple code to calculate normal cooldowns was developed to confirm RHR design bases were met with the improved limits

Design and Manufacturing of Young 3 and 4 NSSS Components

International Nuclear Information System (INIS)

Chung, Chungwoon

1989-01-01

Korea nuclear unit 11 and 12 (Young 3 and 4) project, which is the 6th nuclear construction project in Korea, has been implemented since 1987. The project is scheduled to commence commercial operation by March 1995 and March 1996, respectively. The project is executed in such a manner that local firms play the leading role. In parallel with the project, nationwide technical self-reliance program for nuclear power plant construction is activated. Accordingly, the clear-cut division and achievement of responsibilities assigned to local firms will determine the success of this project and future nuclear projects. The local manufacturer takes responsibility for on-time delivery of safety-assured and reliable equipment and also for achieving technical self-reliance in component design and manufacturing. This paper describes the objectives to be achieved by the local manufacturer in the execution of design and manufacturing of NSSS components for the project and action plans taken and/or to be taken to achieve those objectives

Principles of designing cyber-physical system of producing mechanical assembly components at Industry 4.0 enterprise

Science.gov (United States)

Gurjanov, A. V.; Zakoldaev, D. A.; Shukalov, A. V.; Zharinov, I. O.

2018-03-01

The task of developing principles of cyber-physical system constitution at the Industry 4.0 company of the item designing components of mechanical assembly production is being studied. The task has been solved by analyzing the components and technologies, which have some practical application in the digital production organization. The list of components has been defined and the authors proposed the scheme of the components and technologies interconnection in the Industry 4.0 of mechanical assembly production to make an uninterrupted manufacturing route of the item designing components with application of some cyber-physical systems.

Design/licensing of on-site package for core component

International Nuclear Information System (INIS)

Ogasawara, K.; Chohzuka, T.; Shimura, T.; Kikuchi, T.; Fujiwara, R.; Karigome, S.; Takani, M.

1993-01-01

For storage of used core components which are produced from reactors, Tohoku EPCO decided to construct a site bunker at Onagawa site. It was also decided to develop and fabricate one packaging to transport core components from the reactor buildings to the site bunker. The packaging will be used within the power station; therefore, it shall comply with 'The Law for the Business of Electric Power' and relevant Notification. The main requirements of the packaging are as follows: 1) The number of contents, such as channel boxes and control rods, shall be as large as possible. 2) The weight and the outer dimensions of the packaging shall be within the limitation of the reactor building and the site bunker. 3) Materials shall be selected from those which have been already applied for existing packagings and utilized without any problems. 4) It shall be considered during design of trunnions that handling equipment, such as lifting beam, can be used for not only this packaging but also for existing spent fuel packagings. The design of the packaging is completed and has been licensed. The packaging is scheduled to be utilized from November, 1993. (J.P.N.)

Design and development of gas turbine high temperature reactor 300 (GTHTR300)

International Nuclear Information System (INIS)

Kunitomi, Kazuhiko; Katanishi, Shoji; Takada, Shoji; Takizuka, Takakazu; Yan, Xing; Kosugiyama, Shinichi

2003-01-01

JAERI (Japan Atomic Energy Research Institute) started design and development of the high temperature gas cooled reactor with a gas turbine electric generation system, GTHTR300, in April 2001. Design originalities of the GTHTR300 are a horizontally mounted highly efficient gas turbine system and an ultimately simplified safety system such as no containment building and no active emergency core cooling. These design originalities are proposed based on design and operational experiences in conventional gas turbine systems and Japan's first high temperature gas cooled reactor (HTTR: High Temperature Engineering Test Reactor) so that many R and Ds are not required for the development. Except these original design features, devised core design, fuel design and plant design are adopted to meet design requirements and attain a target cost. This paper describes the unique design features focusing on the safety design, reactor core design and gas turbine system design together with a preliminary result of the safety evaluation carried out for a typical severe event. This study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

A new electrodynamic balance (EDB) design for low-temperature studies: application to immersion freezing of pollen extract bioaerosols

Science.gov (United States)

Tong, H.-J.; Ouyang, B.; Nikolovski, N.; Lienhard, D. M.; Pope, F. D.; Kalberer, M.

2015-03-01

In this paper we describe a newly designed cold electrodynamic balance(CEDB) system, built to study the evaporation kinetics and freezing properties of supercooled water droplets. The temperature of the CEDB chamber at the location of the levitated water droplet can be controlled in the range -40 to +40 °C, which is achieved using a combination of liquid nitrogen cooling and heating by positive temperature coefficient heaters. The measurement of liquid droplet radius is obtained by analysing the Mie elastic light scattering from a 532 nm laser. The Mie scattering signal was also used to characterise and distinguish droplet freezing events; liquid droplets produce a regular fringe pattern, whilst the pattern from frozen particles is irregular. The evaporation rate of singly levitated water droplets was calculated from time-resolved measurements of the radii of evaporating droplets and a clear trend of the evaporation rate on temperature was measured. The statistical freezing probabilities of aqueous pollen extracts (pollen washing water) are obtained in the temperature range -4.5 to -40 °C. It was found that that pollen washing water from water birch (Betula fontinalis occidentalis) pollen can act as ice nuclei in the immersion freezing mode at temperatures as warm as -22.45 (±0.65) °C. Furthermore it was found that the protein-rich component of the washing water was significantly more ice-active than the non-proteinaceous component.

Mechanical and materials engineering of modern structure and component design

CERN Document Server

Altenbach, Holm

2015-01-01

This book presents the latest findings on mechanical and materials engineering as applied to the design of modern engineering materials and components. The contributions cover the classical fields of mechanical, civil and materials engineering, as well as bioengineering and advanced materials processing and optimization. The materials and structures discussed can be categorized into modern steels, aluminium and titanium alloys, polymers/composite materials, biological and natural materials, material hybrids and modern nano-based materials. Analytical modelling, numerical simulation, state-of-the-art design tools and advanced experimental techniques are applied to characterize the materials’ performance and to design and optimize structures in different fields of engineering applications.

The MainSTREAM Component Platform: A Holistic Approach to Microfluidic System Design

DEFF Research Database (Denmark)

Sabourin, David; Skafte-Pedersen, Peder; Søe, Martin Jensen

2013-01-01

A microfluidic component library for building systems driving parallel or serial microfluidic-based assays is presented. The components are a miniaturized eight-channel peristaltic pump, an eight-channel valve, sample-to-waste liquid management, and interconnections. The library of components...... of reaction chips; (2) highly parallel pumping and routing/valving capability; (3) methods to interface pumps and chip-to-liquid management systems; (4) means to construct a portable system; (5) reconfigurability/flexibility in system design; (6) means to interface to microscopes; and (7) compatibility...

Metrology for WEST components design and integration optimization

International Nuclear Information System (INIS)

Brun, C.; Archambeau, G.; Blanc, L.; Bucalossi, J.; Chantant, M.; Gargiulo, L.; Hermenier, A.; Le, R.; Pilia, A.

2015-01-01

Highlights: • Metrology methods. • Interests of metrology campaign to optimize margins by reducing uncertainties. • Assembly problems are solved and validated on a numerical mock up. • Post treatment of full 3DScan of the vacuum vessel. - Abstract: On WEST new components will be implemented in an existing environment, emphasis has to be put on the metrology to optimize the design and the assembly. Hence, at a particular stage of the project, several components have to coexist in the limited vessel. Therefore, all the difficulty consists in validating the mechanical interfaces between existing components and new one; minimize the risk of the assembling and to maximize the plasma volume. The CEA/IRFM takes the opportunity of the ambitious project to sign a partnership with an industrial specialized in multipurpose metrology domains. To optimize the assembly procedure, the IRFM Assembly group works in strong collaboration with its industrial, to define and plan the campaigns of metrology. The paper will illustrate the organization, methods and results of the dedicated metrology campaigns have been defined and carried out in the WEST dis/assembly phase. To conclude, the future needs of metrology at CEA/IRFM will be exposed to define the next steps.

Design and Application of an Ontology for Component-Based Modeling of Water Systems

Science.gov (United States)

Elag, M.; Goodall, J. L.

2012-12-01

Many Earth system modeling frameworks have adopted an approach of componentizing models so that a large model can be assembled by linking a set of smaller model components. These model components can then be more easily reused, extended, and maintained by a large group of model developers and end users. While there has been a notable increase in component-based model frameworks in the Earth sciences in recent years, there has been less work on creating framework-agnostic metadata and ontologies for model components. Well defined model component metadata is needed, however, to facilitate sharing, reuse, and interoperability both within and across Earth system modeling frameworks. To address this need, we have designed an ontology for the water resources community named the Water Resources Component (WRC) ontology in order to advance the application of component-based modeling frameworks across water related disciplines. Here we present the design of the WRC ontology and demonstrate its application for integration of model components used in watershed management. First we show how the watershed modeling system Soil and Water Assessment Tool (SWAT) can be decomposed into a set of hydrological and ecological components that adopt the Open Modeling Interface (OpenMI) standard. Then we show how the components can be used to estimate nitrogen losses from land to surface water for the Baltimore Ecosystem study area. Results of this work are (i) a demonstration of how the WRC ontology advances the conceptual integration between components of water related disciplines by handling the semantic and syntactic heterogeneity present when describing components from different disciplines and (ii) an investigation of a methodology by which large models can be decomposed into a set of model components that can be well described by populating metadata according to the WRC ontology.

Operating temperatures for an LMFBR

International Nuclear Information System (INIS)

Bhoje, S.B.; Chellapandi, P.

1993-01-01

The scope of the present paper is limited to structural mechanics aspects that are associated with this technology. However, for the purpose of comprehensive presentation, all the other related issues are also highlighted. For this study, a Prototype Fast Breeder Reactor (PFBR) with 500 MWe capacity is taken as the reference design. Accordingly, some critical high temperature components of PFBR are analysed in- detail for elastic, inelastic and viscoplastic behaviour towards life prediction as per the requirement of design codes (RCC-MR 87) which form basis for justifying the possibility of higher operating temperatures for LMFBRs. Since operation with higher primary sodium outlet temperature in association with higher ΔT across the core is one of the efficient techniques towards making LMFBRs cost effective, operating Temperature limits are determined for a typical pool type FBR of 500 MWe capacity. Analysis indicates that control plug in the hot pool is the most critical component which limits the operating temperature to 820 K with a ΔT across the core of 160 K. By improving the thermal hydraulic design in conjunction with the structural design optimisation at the plate-shell junctions of control plug, possibility exists to go up to 840-850 K for primary outlet sodium with a T of 160 K across the core. This will result in producing steam of about 790-800 K (520 deg. C). Apart from improving the thermal hydraulic design to mitigate the transient thermal stresses, following are also needed to demonstrate higher safety margins in the design. Reduction of thermal transients, for an example, the temperature drop in the primary sodium outlet can be reduced by decreasing the sodium flow rate to the core, during a reactor scram. Welds should be avoided at the plate-shell junctions of control plug. A complete ring with necessary fillet radius may be forged as a single piece. In case of reactor vessel, a pullout option is better for redan-stand pipe junction

Synthesis of results obtained on sodium components and technology through the Generation IV International Forum SFR Component Design and Balance-of-Plant Project

International Nuclear Information System (INIS)

Sienicki, J.J.; Rodriguez, G.; Kisohara, N.; Kim, J. B.; Gerber, A.; Ashurko, Y.; Toyama, S.

2013-01-01

Status: The viability of designing SFR components and BOP has been demonstrated with design, construction and operation of previous sodium-cooled reactors. The main objective of this R&D project is related to system performance, or by development on the use of AECS in the BOP that could allow further cost improvements. Objective: To conduct collaborative research and development of components and BOP for the SFR System. The Project has to satisfy the GIF’s criteria of safety, economy, sustainability, proliferation resistance and physical protection. Activities within this Project are addressing experimental and analytical evaluation of advanced ISI&R, LBB assessment, development of AECS with Brayton cycles, advanced SG technologies. Project activities will be based in part on the extensive historical R&D experience with component design and balance of plant for sodium-cooled fast reactors

Engineering design and fabrication of X-Band components

CERN Document Server

Filippova, M; Solodko, A; Riddone, G; Syratchev, I

2011-01-01

The CLIC RF frequency has been changed in 2008 from the initial 30 GHz to the European X-band 11.994 GHz permitting beam independent power production using klystrons for the accelerating structure testing. X-band klystron test facilities at 11.424 GHz are operated at SLAC and at KEK [1], and they are used by the CLIC study in the framework of the X-band structure collaboration for testing accelerating structures scaled to that frequency [2]. CERN is currently building a klystron test-stand operating at 11.994 GHz. In addition X-FEL projects at PSI and Sincrotrone Trieste operate at 11.4 GHz. Therefore several RF components accommodating frequencies from 11.424 to 11.994 GHz are required. The engineering design of these RF components (high power and compact loads, bi-directional couplers, X-band splitters, hybrids, phase shifters, variable power attenuators) and the main fabrication processes are presented here.

Selective extraction of intracellular components from the microalga Chlorella vulgaris by combined pulsed electric field-temperature treatment

NARCIS (Netherlands)

Postma, P.R.; Pataro, G.; Capitoli, M.; Barbosa, M.J.; Wijffels, R.H.; Eppink, M.H.M.; Olivieri, G.; Ferrari, G.

2016-01-01

The synergistic effect of temperature (25-65°C) and total specific energy input (0.55-1.11kWhkg_DW ^-1) by pulsed electric field (PEF) on the release of intracellular components from the microalgae Chlorella vulgaris was studied. The combination of PEF with temperatures from

Considerations on the design of a helium circulator for a high temperature modular reactor system

International Nuclear Information System (INIS)

Dumm, K.; Donaldson, J.

1988-01-01

A modular helium cooled, high temperature reactor system with a thermal output of 200 MW per reactor has been developed by the KWU group for cogeneration of electricity and process steam. The flow of the reactor coolant - Helium at 60 bars and 250/700 deg. C is maintained by one circulator per reactor. The circulator is driven by a variable speed Siemens asynchronous motor and is submerged in the helium primary system. For operational reasons high reliability and availability of the circulator is required. The operational requirements for the circulator design are presented in this paper. The actual design has been carried out in close cooperation with the designer and manufacturer of all submerged circulators operating in AGR plants in Great Britain, James Howden Co. Renfrew, Scotland. Design solutions received so far and mainly based on sufficiently proven components - such as oil bath lubricated bearing systems - will be described. Special attention will be paid on the necessary test work; especially for the prototype to confirm the lay out. (author). 9 figs

Principle of maximum entropy for reliability analysis in the design of machine components

Science.gov (United States)

Zhang, Yimin

2018-03-01

We studied the reliability of machine components with parameters that follow an arbitrary statistical distribution using the principle of maximum entropy (PME). We used PME to select the statistical distribution that best fits the available information. We also established a probability density function (PDF) and a failure probability model for the parameters of mechanical components using the concept of entropy and the PME. We obtained the first four moments of the state function for reliability analysis and design. Furthermore, we attained an estimate of the PDF with the fewest human bias factors using the PME. This function was used to calculate the reliability of the machine components, including a connecting rod, a vehicle half-shaft, a front axle, a rear axle housing, and a leaf spring, which have parameters that typically follow a non-normal distribution. Simulations were conducted for comparison. This study provides a design methodology for the reliability of mechanical components for practical engineering projects.

Fabrication and Characterizations of Materials and Components for Intermediate Temperature Fuel Cells and Water Electrolysers

DEFF Research Database (Denmark)

Jensen, Annemette Hindhede; Prag, Carsten Brorson; Li, Qingfeng

The worldwide development of fuel cells and electrolysers has so far almost exclusively addressed either the low temperature window (20-200 °C) or the high temperature window (600-1000 °C). This work concerns the development of key materials and components of a new generation of fuel cells...... and electrolysers for operation in the intermediate temperature range from 200 to 400 °C. The intermediate temperature interval is of importance for the use of renewable fuels. Furthermore electrode kinetics is significantly enhanced compared to when operating at low temperature. Thus non-noble metal catalysts...... might be used. One of the key materials in the fuel cell and electrolyser systems is the electrolyte. Proton conducting materials such as cesium hydrogen phosphates, zirconium hydrogen phosphates and tin pyrophosphates have been investigated by others and have shown interesting potential....

Design considerations: gas turbines for electric power generation

International Nuclear Information System (INIS)

Moon, D.M.

1979-01-01

The gas turbine represents one of the most sophisticated designs from the standpoint of time dependent deformation behavior. The large size of the equipment, which limits the amount of full scale testing, together with the demanding performance requirements and high level of reliability desired places a high degree of emphasis on the high temperature deformation design process. As an example of the various design considerations used in this equipment, a brief overview of the turbine will be given, highlighting the materials, stress, temperatures, and load history experienced by the major components. Particular attention will then be focused on the vane segment design considerations. This component is not only structurally complicated, but experiences steep temperature gradients imposed by internal cooling and large temperature transients during cyclic duty operation which have to be addressed in the design procedure. Based on this discussion the limitations of the current design procedures will be highlighted and the areas requiring additional research inputs will be discussed

Component Functional Allocations of the ESF Multi-loop Controller for the KNICS ESF-CCS Design

International Nuclear Information System (INIS)

Hur, Seop; Choi, Jong Kyun; Kim, Dong Hoon; Kim, Ho; Kim, Seong Tae

2006-01-01

The safety related components in nuclear power plants are traditionally controlled by single-loop controllers. Traditional single-loop controller systems utilize dedicated processors for each component but that components independence is compromised through a sharing of power supplies, auxiliary logic modules and auxiliary I/O cards. In the new design of the ESF-CCS, the multi-loop controllers with data networks are widely used. Since components are assigned to ESF-CCS functional groups in a manner consistent with their process relationship, the effects of the failures are predictable and manageable. Therefore, the key issues for the design of multi-loop controller is to allocate the components to the each multi-loop controller through plant and function analysis and grouping. This paper deals with an ESF component functional allocation which is performed through allocation criteria and a fault analysis

Design and evaluation of a pressure sensor for high temperature nuclear application

International Nuclear Information System (INIS)

Yancey, M.E.

1981-11-01

The goal of this technical development task was the development of a small eddy-current pressure sensor for use within a high temperature nuclear environment. The sensor is designed for use at pressures and temperatures of up to 17.23 MPa and 650 0 F. The design of the sensor incorporated features to minimize possible errors due to temperature transients present in nuclear applications. This report describes a prototype pressure sensor that was designed, the associated 100 kHz signal conditioning electronics, and the evaluation tests which were conducted

A Brief Description of High Temperature Solid Oxide Fuel Cell’s Operation, Materials, Design, Fabrication Technologies and Performance

Directory of Open Access Journals (Sweden)

Muneeb Irshad

2016-03-01

Full Text Available Today’s world needs highly efficient systems that can fulfill the growing demand for energy. One of the promising solutions is the fuel cell. Solid oxide fuel cell (SOFC is considered by many developed countries as an alternative solution of energy in near future. A lot of efforts have been made during last decade to make it commercial by reducing its cost and increasing its durability. Different materials, designs and fabrication technologies have been developed and tested to make it more cost effective and stable. This article is focused on the advancements made in the field of high temperature SOFC. High temperature SOFC does not need any precious catalyst for its operation, unlike in other types of fuel cell. Different conventional and innovative materials have been discussed along with properties and effects on the performance of SOFC’s components (electrolyte anode, cathode, interconnect and sealing materials. Advancements made in the field of cell and stack design are also explored along with hurdles coming in their fabrication and performance. This article also gives an overview of methods required for the fabrication of different components of SOFC. The flexibility of SOFC in terms fuel has also been discussed. Performance of the SOFC with varying combination of electrolyte, anode, cathode and fuel is also described in this article.

CRBR reactor structures design. BRC meeting presentation

International Nuclear Information System (INIS)

Pennell, W.E.

1975-01-01

Some of the more important developments in LMFBR structures design technology are described and the application of the technology to design of the CRBR reactor components is illustrated. The LMFBR is both a high-temperature and a high-ΔT machine. High-temperature operation (up to 1100 0 F) requires that the designer consider the effects of thermal creep as a deformation mechanism and stress rupture as a failure mode. The large ΔT across the core coupled with a low core thermal inertia and the high conductivity of the sodium coolant combine to produce severe temperature gradients during a reactor scram. Structures designed to operate in this environment must be both light and stiff to minimize transient thermal stresses and prevent unacceptable flow-induced vibrations. Thermal shields may be required to protect the load-bearing structure. At CRBR core-component goal fluence levels, the predicted magnitude of core-component dimensional changes due to irradiation swelling and creep is very large compared with the more familiar dimensional changes associated with thermal expansion and thermal creep. The design of the core components, and in particular the core restraint system, is dominated by the need to accommodate the effects of irradiation swelling, creep and du []tility loss considerations. (auth)

Alternative designs of high-temperature superconducting synchronous generators

OpenAIRE

Goddard, K. F.; Lukasik, B.; Sykulski, J. K.

2010-01-01

This paper discusses the different possible designs of both cored and coreless superconducting synchronous generators using high-temperature superconducting (HTS) tapes, with particular reference to demonstrators built at the University of Southampton using BiSCCO conductors. An overview of the electromagnetic, thermal, and mechanical issues is provided, the advantages and drawbacks of particular designs are highlighted, the need for compromises is explained, and practical solutions are offer...

Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR)

Energy Technology Data Exchange (ETDEWEB)

Ingersoll, D.T.

2004-07-29

A new reactor plant concept is presented that combines the benefits of ceramic-coated, high-temperature particle fuel with those of clean, high-temperature, low-pressure molten salt coolant. The Advanced High-Temperature Reactor (AHTR) concept is a collaboration of Oak Ridge National Laboratory, Sandia National Laboratories, and the University of California at Berkeley. The purpose of the concept is to provide an advanced design capable of satisfying the top-level functional requirements of the U.S. Department of Energy Next Generation Nuclear Plant (NGNP), while also providing a technology base that is sufficiently robust to allow future development paths to higher temperatures and larger outputs with highly competitive economics. This report summarizes the status of the AHTR preconceptual design. It captures the results from an intense effort over a period of 3 months to (1) screen and examine potential feasibility concerns with the concept; (2) refine the conceptual design of major systems; and (3) identify research, development, and technology requirements to fully mature the AHTR design. Several analyses were performed and are presented to quantify the AHTR performance expectations and to assist in the selection of several design parameters. The AHTR, like other NGNP reactor concepts, uses coated particle fuel in a graphite matrix. But unlike the other NGNP concepts, the AHTR uses molten salt rather than helium as the primary system coolant. The considerable previous experience with molten salts in nuclear environments is discussed, and the status of high-temperature materials is reviewed. The large thermal inertia of the system, the excellent heat transfer and fission product retention characteristics of molten salt, and the low-pressure operation of the primary system provide significant safety attributes for the AHTR. Compared with helium coolant, a molten salt cooled reactor will have significantly lower fuel temperatures (150-200-C lower) for the

Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR)

International Nuclear Information System (INIS)

Ingersoll, D.T.

2004-01-01

A new reactor plant concept is presented that combines the benefits of ceramic-coated, high-temperature particle fuel with those of clean, high-temperature, low-pressure molten salt coolant. The Advanced High-Temperature Reactor (AHTR) concept is a collaboration of Oak Ridge National Laboratory, Sandia National Laboratories, and the University of California at Berkeley. The purpose of the concept is to provide an advanced design capable of satisfying the top-level functional requirements of the U.S. Department of Energy Next Generation Nuclear Plant (NGNP), while also providing a technology base that is sufficiently robust to allow future development paths to higher temperatures and larger outputs with highly competitive economics. This report summarizes the status of the AHTR preconceptual design. It captures the results from an intense effort over a period of 3 months to (1) screen and examine potential feasibility concerns with the concept; (2) refine the conceptual design of major systems; and (3) identify research, development, and technology requirements to fully mature the AHTR design. Several analyses were performed and are presented to quantify the AHTR performance expectations and to assist in the selection of several design parameters. The AHTR, like other NGNP reactor concepts, uses coated particle fuel in a graphite matrix. But unlike the other NGNP concepts, the AHTR uses molten salt rather than helium as the primary system coolant. The considerable previous experience with molten salts in nuclear environments is discussed, and the status of high-temperature materials is reviewed. The large thermal inertia of the system, the excellent heat transfer and fission product retention characteristics of molten salt, and the low-pressure operation of the primary system provide significant safety attributes for the AHTR. Compared with helium coolant, a molten salt cooled reactor will have significantly lower fuel temperatures (150-200-C lower) for the

Design Basis of Core Components and their Realization in the frame of the EPR's{sup TM} Core Component Development

Energy Technology Data Exchange (ETDEWEB)

Schebitz, Florian [AREVA NP GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany); Mekmouche, Abdelhalim [AREVA NP SAS, 10 rue Juliette Recamier, 69456 Lyon Cedex 06 (France)

2008-07-01

Rod Cluster Control Assemblies (RCCAs), Thimble Plug Assemblies (TPAs), Primary Neutron Sources (PNS) and Secondary Neutron Sources (SNS) are essential for the operation of a Nuclear Power Plant. Different functional requirements ask for different components and geometries. Therefore three different core components are used within the primary circuit: - The RCCA, which contains the absorber materials, is used to regulate and shut down the nuclear chain reaction. Under these demanding conditions different effects are determining the lifetime of the RCCA and in particular of the control rods. Several improvements like ion-nitriding of the cladding, lengthening of the bottom end plug, helium backfilling and reduction of the absorber diameter in the bottom part, which have already been introduced with the HARMONI{sup TM} RCCA, show a real improvement in terms of lifetime. - The TPAs are used at positions without RCCAs and neutron sources to limit the by-pass flow-rate in the fuel assembly guide tubes. The advanced TPA design results from a perfect combination of French and German design experience feedback. Benefits like homogenized hydraulic flow and improved manageability in terms of handling tools show the joined experience. - The neutron sources are used to enhance the flux level when the core is sub-critical so as to facilitate the core start-up control by the neutron flux detectors. Primary and secondary neutron sources are designed in a common way with reviewed and improved methodology. As there are different ways and conditions to operate core components, several designs are available. For the EPR{sup TM}, the best methods and products have been chosen. All chosen components contribute to an optimized and safe operation of the EPR{sup TM}. (authors)

Deep Trek High Temperature Electronics Project

Energy Technology Data Exchange (ETDEWEB)

Bruce Ohme

2007-07-31

This report summarizes technical progress achieved during the cooperative research agreement between Honeywell and U.S. Department of Energy to develop high-temperature electronics. Objects of this development included Silicon-on-Insulator (SOI) wafer process development for high temperature, supporting design tools and libraries, and high temperature integrated circuit component development including FPGA, EEPROM, high-resolution A-to-D converter, and a precision amplifier.

Components of the LWR primary circuit. Pt. 2

International Nuclear Information System (INIS)

1984-01-01

This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673 K (400 0 C). The primary circuit as the pressure containment of the reactor coolant comprises: Reactor pressure vessel (without internals), steam generator (primary loop), pressurizer, reactor coolant pump housing, interconnecting pipings between the components mentioned above and appropriate various valve and instrument casings, pipings branding from the above components and interconnecting pipings, including the appropriate instrument casings, up to and including the first isolating valve, pressure shielding of control rod drives. (orig.) [de

Design of the Mechanical Components of a Dual Axis Solar Tracker

OpenAIRE

Romero Llanas, Amador

2013-01-01

This work is about the design of a solar tracker with the objective of following the sun throughout the day. In order to achieve that objective, the solar tracker has two degrees of freedom. The different mechanical components necessary to build the structure has been designed, calculated and verified. Apart from that, the whole structure has been drawn using the 3D mechanical CAD program SolidWorks. The plans have been drawn too.

Small punch creep test: A promising methodology for high temperature plant components life evaluation

Energy Technology Data Exchange (ETDEWEB)

Tettamanti, S [CISE SpA, Milan (Italy); Crudeli, R [ENEL SpA, Milan (Italy)

1999-12-31

CISE and ENEL are involved for years in a miniaturization creep methodology project to obtain similar non-destructive test with the same standard creep test reliability. The goal can be reached with `Small punch creep test` that collect all the requested characteristics; quasi nondestructive disk specimens extracted both on external or internal side of components, than accurately machined and tested on little and cheap apparatus. CISE has developed complete creep small punch procedure that involved peculiar test facility and correlation`s law comparable with the more diffused isostress methodology for residual life evaluation on ex-serviced high temperature plant components. The aim of this work is to obtain a simple and immediately applicable relationship useful for plant maintenance managing. More added work is need to validate the Small Punch methodology and for relationship calibration on most diffusion high temperature structural materials. First obtained results on a comparative work on ASTM A355 P12 ex-serviced pipe material are presented joint with a description of the Small Punch apparatus realized in CISE. (orig.) 6 refs.

Small punch creep test: A promising methodology for high temperature plant components life evaluation

Energy Technology Data Exchange (ETDEWEB)

Tettamanti, S. [CISE SpA, Milan (Italy); Crudeli, R. [ENEL SpA, Milan (Italy)

1998-12-31

CISE and ENEL are involved for years in a miniaturization creep methodology project to obtain similar non-destructive test with the same standard creep test reliability. The goal can be reached with `Small punch creep test` that collect all the requested characteristics; quasi nondestructive disk specimens extracted both on external or internal side of components, than accurately machined and tested on little and cheap apparatus. CISE has developed complete creep small punch procedure that involved peculiar test facility and correlation`s law comparable with the more diffused isostress methodology for residual life evaluation on ex-serviced high temperature plant components. The aim of this work is to obtain a simple and immediately applicable relationship useful for plant maintenance managing. More added work is need to validate the Small Punch methodology and for relationship calibration on most diffusion high temperature structural materials. First obtained results on a comparative work on ASTM A355 P12 ex-serviced pipe material are presented joint with a description of the Small Punch apparatus realized in CISE. (orig.) 6 refs.

Safety philosophy and design principles for systems and components of nuclear power plant: external event

International Nuclear Information System (INIS)

Lopes, J.P.G.

1986-01-01

In nuclear power plants, some systems and components are designed to withstand external impacts. Such systems and components are those which have to perform their functions even during and after the occurrences of an earthquake, for example, fulfilling the safety objectives and avoiding the release of radioactive material to the environment. The aim of this report is to introduce the safety philosophy and design principles for systems/components to perform their functions during and after the occurrence of an earthquake, as applied by NUCLEN for Angra 2 and 3. (Author) [pt

Numerical simulation of the laser welding process for the prediction of temperature distribution on welded aluminium aircraft components

Science.gov (United States)

Tsirkas, S. A.

2018-03-01

The present investigation is focused to the modelling of the temperature field in aluminium aircraft components welded by a CO2 laser. A three-dimensional finite element model has been developed to simulate the laser welding process and predict the temperature distribution in T-joint laser welded plates with fillet material. The simulation of the laser beam welding process was performed using a nonlinear heat transfer analysis, based on a keyhole formation model analysis. The model employs the technique of element ;birth and death; in order to simulate the weld fillet. Various phenomena associated with welding like temperature dependent material properties and heat losses through convection and radiation were accounted for in the model. The materials considered were 6056-T78 and 6013-T4 aluminium alloys, commonly used for aircraft components. The temperature distribution during laser welding process has been calculated numerically and validated by experimental measurements on different locations of the welded structure. The numerical results are in good agreement with the experimental measurements.

Elevated temperature design of KALIMER reactor internals accounting for creep and stress-rupture effects

International Nuclear Information System (INIS)

Koo, Gyeong Hoi; Yoo, Bong

2000-01-01

In most LMFBR (Liquid Metal Fast Breed Reactor) design, the operating temperature is very high and the time-dependent creep and stress-rupture effects become so important in reactor structural design. Therefore, unlike with conventional PWR, the normal operating conditions can be basically dominant design loading because the hold time at elevated temperature condition is so long and enough to result in severe total creep ratcheting strains during total service lifetime. In this paper, elevated temperature design of the conceptually designed baffle annulus regions of KALIMER (Korea Advanced Liquid Metal Reactor) reactor internal structures is carried out for normal operating conditions which have the operating temperature 530 deg. C and the total service lifetime of 30 years. For the elevated temperature design of reactor internal structures, the ASME Code Case N-201-4 is used. Using this code, the time-dependent stress limits, the accumulated total inelastic strain during service lifetime, and the creep-fatigue damages are evaluated with the calculation results by the elastic analysis under conservative assumptions. The application procedures of elevated temperature design of the reactor internal structures using ASME code case N-201-4 with the elastic analysis method are described step by step in detail. This paper will be useful guide for actual application of elevated temperature design of various reactor types accounting for creep and stress-rupture effects. (author)

Material characterisation and preliminary mechanical design for the HL-LHC shielded beam screens operating at cryogenic temperatures

CERN Document Server

Garion, C; Koettig, T; Machiocha, W; Morrone, M

2015-01-01

The High Luminosity LHC project (HL-LHC) aims at increasing the luminosity (rate of collisions) in the Large Hadron Collider (LHC) experiments by a factor of 10 beyond the original design value (from 300 to 3000 fb-1). It relies on new superconducting magnets, installed close to the interaction points, equipped with new beam screen. This component has to ensure the vacuum performance together with shielding the cold mass from physics debris and screening the cold bore cryogenic system from beam induced heating. The beam screen operates in the range 40-60 K whereas the magnet cold bore temperature is 1.9 K. A tungsten-based material is used to absorb the energy of particles. In this paper, measurements of the mechanical and physical properties of such tungsten material are shown at room and cryogenic temperature. In addition, the design and the thermal mechanical behaviour of the beam screen assembly are presented also. They include the heat transfer from the tungsten absorbers to the cooling pipes and the sup...

Design and analysis of automobile components using industrial procedures

Science.gov (United States)

Kedar, B.; Ashok, B.; Rastogi, Nisha; Shetty, Siddhanth

2017-11-01

Today’s automobiles depend upon mechanical systems that are crucial for aiding in the movement and safety features of the vehicle. Various safety systems such as Antilock Braking System (ABS) and passenger restraint systems have been developed to ensure that in the event of a collision be it head on or any other type, the safety of the passenger is ensured. On the other side, manufacturers also want their customers to have a good experience while driving and thus aim to improve the handling and the drivability of the vehicle. Electronics systems such as Cruise Control and active suspension systems are designed to ensure passenger comfort. Finally, to ensure optimum and safe driving the various components of a vehicle must be manufactured using the latest state of the art processes and must be tested and inspected with utmost care so that any defective component can be prevented from being sent out right at the beginning of the supply chain. Therefore, processes which can improve the lifetime of their respective components are in high demand and much research and development is done on these processes. With a solid base research conducted, these processes can be used in a much more versatile manner for different components, made up of different materials and under different input conditions. This will help increase the profitability of the process and also upgrade its value to the industry.

Internet MEMS design tools based on component technology

Science.gov (United States)

Brueck, Rainer; Schumer, Christian

1999-03-01

The micro electromechanical systems (MEMS) industry in Europe is characterized by small and medium sized enterprises specialized on products to solve problems in specific domains like medicine, automotive sensor technology, etc. In this field of business the technology driven design approach known from micro electronics is not appropriate. Instead each design problem aims at its own, specific technology to be used for the solution. The variety of technologies at hand, like Si-surface, Si-bulk, LIGA, laser, precision engineering requires a huge set of different design tools to be available. No single SME can afford to hold licenses for all these tools. This calls for a new and flexible way of designing, implementing and distributing design software. The Internet provides a flexible manner of offering software access along with methodologies of flexible licensing e.g. on a pay-per-use basis. New communication technologies like ADSL, TV cable of satellites as carriers promise to offer a bandwidth sufficient even for interactive tools with graphical interfaces in the near future. INTERLIDO is an experimental tool suite for process specification and layout verification for lithography based MEMS technologies to be accessed via the Internet. The first version provides a Java implementation even including a graphical editor for process specification. Currently, a new version is brought into operation that is based on JavaBeans component technology. JavaBeans offers the possibility to realize independent interactive design assistants, like a design rule checking assistants, a process consistency checking assistants, a technology definition assistants, a graphical editor assistants, etc. that may reside distributed over the Internet, communicating via Internet protocols. Each potential user thus is able to configure his own dedicated version of a design tool set dedicated to the requirements of the current problem to be solved.

Temperature field analysis of single layer TiO2 film components induced by long-pulse and short-pulse lasers

International Nuclear Information System (INIS)

Wang Bin; Zhang Hongchao; Qin Yuan; Wang Xi; Ni Xiaowu; Shen Zhonghua; Lu Jian

2011-01-01

To study the differences between the damaging of thin film components induced by long-pulse and short-pulse lasers, a model of single layer TiO 2 film components with platinum high-absorptance inclusions was established. The temperature rises of TiO 2 films with inclusions of different sizes and different depths induced by a 1 ms long-pulse and a 10 ns short-pulse lasers were analyzed based on temperature field theory. The results show that there is a radius range of inclusions that corresponds to high temperature rises. Short-pulse lasers are more sensitive to high-absorptance inclusions and long-pulse lasers are more easily damage the substrate. The first-damage decision method is drawn from calculations.

Embedded DAQ System Design for Temperature and Humidity Measurement

International Nuclear Information System (INIS)

Memon, T.R.

2013-01-01

In this work, we have proposed a cost effective DAQ (Data Acquisition) system design useful for local industries by using user friendly LABVIEW (Laboratory Virtual Instrumentation Electronic Workbench). The proposed system can measure and control different industrial parameters which can be presented in graphical icon format. The system design is proposed for 8-channels, whereas tested and recorded for two parameters i.e. temperature and RH (Relative Humidity). Both parameters are set as per upper and lower limits and controlled using relays. Embedded system is developed using standard microcontroller to acquire and process the analog data and plug-in for further processing using serial interface with PC using LABVIEW. The designed system is capable of monitoring and recording the corresponding linkage between temperature and humidity in industrial unit's and indicates the abnormalities within the process and control those abnormalities through relays. (author)

Some thoughts on the future of probabilistic structural design of nuclear components

International Nuclear Information System (INIS)

Stancampiano, P.A.

1978-01-01

This paper presents some views on the future role of probabilistic methods in the structural design of nuclear components. The existing deterministic design approach is discussed and compared to the probabilistic approach. Some of the objections to both deterministic and probabilistic design are listed. Extensive research and development activities are required to mature the probabilistic approach suficiently to make it cost-effective and competitive with current deterministic design practices. The required research activities deal with probabilistic methods development, more realistic casual failure mode models development, and statistical data models development. A quasi-probabilistic structural design approach is recommended which accounts for the random error in the design models. (Auth.)

Successful Bullying Prevention Programs: Influence of Research Design, Implementation Features, and Program Components

Directory of Open Access Journals (Sweden)

Bryanna Hahn Fox

2012-12-01

Full Text Available Bullying prevention programs have been shown to be generally effective in reducing bullying and victimization. However, the effects are relatively small in randomized experiments and greater in quasi-experimental and age-cohort designs. Programs that are more intensive and of longer duration (for both children and teachers are more effective, as are programs containing more components. Several program components are associated with large effect sizes, including parent training or meetings and teacher training. These results should inform the design and evaluation of anti-bullying programs in the future, and a system ofaccreditation of effective programs.

Multi parametric sensitivity study applied to temperature measurement of metallic plasma facing components in fusion devices

International Nuclear Information System (INIS)

Aumeunier, M-H.; Corre, Y.; Firdaouss, M.; Gauthier, E.; Loarer, T.; Travere, J-M.; Gardarein, J-L.; EFDA JET Contributor

2013-06-01

In nuclear fusion experiments, the protection system of the Plasma Facing Components (PFCs) is commonly ensured by infrared (IR) thermography. Nevertheless, the surface monitoring of new metallic plasma facing component, as in JET and ITER is being challenging. Indeed, the analysis of infrared signals is made more complicated in such a metallic environment since the signals will be perturbed by the reflected photons coming from high temperature regions. To address and anticipate this new measurement environment, predictive photonic models, based on Monte-Carlo ray tracing (SPEOS R CAA V5 Based), have been performed to assess the contribution of the reflective part in the total flux collected by the camera and the resulting temperature error. This paper deals with the effects of metals features, as the emissivity and reflectivity models, on the accuracy of the surface temperature estimation. The reliability of the features models is discussed by comparing the simulation with experimental data obtained with the wide angle IR thermography system of JET ITER like wall. The impact of the temperature distribution is studied by considering two different typical plasma scenarios, in limiter (ITER start-up scenario) and in X-point configurations (standard divertor scenario). The achievable measurement performances of IR system and risks analysis on its functionalities are discussed. (authors)

Design and construction of electronic components for a ''Novillo'' Tokamak

International Nuclear Information System (INIS)

Lopez C, R.

1986-07-01

The goal of this effort was to design, construct and make functional the electronic components for a ''Novillo'' Tokamak currently being experimentally investigated at the National Institute of Nuclear Research in Mexico. The problem was to develop programmable electronic switches capable of discharging high voltage kilowatt energies stored in capacitator banks onto the coils of the Tokamak. (author)

Selection of fiber-optical components for temperature measurement for satellite applications

Science.gov (United States)

Putzer, P.; Kuhenuri Chami, N.; Koch, A. W.; Hurni, A.; Roner, M.; Obermaier, J.; Lemke, N. M. K.

2017-11-01

investigate the radiation induced wavelength shift. The FBGs react on temperature and strain change, so a decoupling of both physical effects must be assured to allow a precise measurement over large temperature ranges and corresponding potential mechanical stress, passed from the structure to the sensor. This potential source of error is addressed with the design of a strain-decoupled temperature transducer to which the FBGs are glued. The design of the transducer and measurement results of a bending test are provided within this paper. An outlook of the usage of fiber-optical sensing in space applications will be given. One promising field of application are the so called photonically-wired spacecraft panels, where optical fibers with integrated FBGs are being integrated in panels for temperature measurements and high-speed data transfer at the same time.

Conceptual Design of Structural Components of a Dual Cooled Fuel

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyung-Kyu; Lee, Young-Ho; Lee, Kang-Hee; Kim, Jae-Yong; Yoon, Kyung-Ho

2008-01-15

A dual cooled fuel, featured by an internal as well as an external coolant flow passage of a fuel rod, was suggested to enable a large-scaled power-uprate of PWR plant and launched as one of the National Nuclear R and D Projects in 2007. It is necessary to make the dual cooled fuel be compatible with an OPR-1000 system to maximize the economy. Also, the structural components of the dual cooled fuel should be designed to realize their features. To this end, a conceptual design of a spacer grid, outer and center guide tubes, and top and bottom end pieces has been carried out in the project 'Development of Design Technology for Dual Cooled Fuel Structure'. For the spacer grids, it is suggested that springs and dimples are located at or near the cross points of the straps due to a considerably narrowed rod-to-rod gap. Candidate shapes of the grids were also developed and applied for domestic patents. For the outer and center guide tubes, a dual tube like a fuel rod was suggested to make the subchannel areas around the guide tubes be similar to those around the fuel rods of enlarged diameter. It was applied for the domestic patent as well. For the top and bottom end pieces, the shape and pattern have been changed from the conventional ones reflecting the fuel rods' changes. Technical issues and method of resolution for each components were listed up for a basic design works in the following years.

Materials and Components for Low Temperature Solid Oxide Fuel Cells – an Overview

Directory of Open Access Journals (Sweden)

D. Radhika

2013-06-01

Full Text Available This article summarizes the recent advancements made in the area of materials and components for low temperature solid oxide fuel cells (LT-SOFCs. LT-SOFC is a new trend in SOFCtechnology since high temperature SOFC puts very high demands on the materials and too expensive to match marketability. The current status of the electrolyte and electrode materials used in SOFCs, their specific features and the need for utilizing them for LT-SOFC are presented precisely in this review article. The section on electrolytes gives an overview of zirconia, lanthanum gallate and ceria based materials. Also, this review article explains the application of different anode, cathode and interconnect materials used for SOFC systems. SOFC can result in better performance with the application of liquid fuels such methanol and ethanol. As a whole, this review article discusses the novel materials suitable for operation of SOFC systems especially for low temperature operation.

Design of components of reinforced concrete stressed by seismic loads

International Nuclear Information System (INIS)

Sitka, R.

1980-01-01

The example of the type of frame investigated shows that the ductility of the system assumed for standard dimensioning of such a frame lies between two and four. According to the system and the loading different requirements may result for the cross-section, that will have to be observed in design. Derived from these requirements rules are given for the design of frames stiffening in horizontal direction that will guarantee a minimum level of ductility. These rules concern the design of joint and node regions, utilization of the compressive force of the concrete as well as guidance and graduation of the reinforcement according to stud and bolt. By means of some examples of damaged components the effects of violating these rules are made clear. (orig./DG) [de

Design and testing of high temperature micro-ORC test stand using Siloxane as working fluid

Science.gov (United States)

Turunen-Saaresti, Teemu; Uusitalo, Antti; Honkatukia, Juha

2017-03-01

Organic Rankine Cycle is a mature technology for many applications e.g. biomass power plants, waste heat recovery and geothermal power for larger power capacity. Recently more attention is paid on an ORC utilizing high temperature heat with relatively low power. One of the attractive applications of such ORCs would be utilization of waste heat of exhaust gas of combustion engines in stationary and mobile applications. In this paper, a design procedure of the ORC process is described and discussed. The analysis of the major components of the process, namely the evaporator, recuperator, and turbogenerator is done. Also preliminary experimental results of an ORC process utilizing high temperature exhaust gas heat and using siloxane MDM as a working fluid are presented and discussed. The turbine type utilized in the turbogenerator is a radial inflow turbine and the turbogenerator consists of the turbine, the electric motor and the feed pump. Based on the results, it was identified that the studied system is capable to generate electricity from the waste heat of exhaust gases and it is shown that high molecular weight and high critical temperature fluids as the working fluids can be utilized in high-temperature small-scale ORC applications. 5.1 kW of electric power was generated by the turbogenerator.

Design of a Novel Two-Component Hybrid Dermal Scaffold for the Treatment of Pressure Sores.

Science.gov (United States)

Sharma, Vaibhav; Kohli, Nupur; Moulding, Dale; Afolabi, Halimat; Hook, Lilian; Mason, Chris; García-Gareta, Elena

2017-11-01

The aim of this study is to design a novel two-component hybrid scaffold using the fibrin/alginate porous hydrogel Smart Matrix combined to a backing layer of plasma polymerized polydimethylsiloxane (Sil) membrane to make the fibrin-based dermal scaffold more robust for the treatment of the clinically challenging pressure sores. A design criteria are established, according to which the Sil membranes are punched to avoid collection of fluid underneath. Manual peel test shows that native silicone does not attach to the fibrin/alginate component while the plasma polymerized silicone membranes are firmly bound to fibrin/alginate. Structural characterization shows that the fibrin/alginate matrix is intact after the addition of the Sil membrane. By adding a Sil membrane to the original fibrin/alginate scaffold, the resulting two-component scaffolds have a significantly higher shear or storage modulus G'. In vitro cell studies show that dermal fibroblasts remain viable, proliferate, and infiltrate the two-component hybrid scaffolds during the culture period. These results show that the design of a novel two-component hybrid dermal scaffold is successful according to the proposed design criteria. To the best of the authors' knowledge, this is the first study that reports the combination of a fibrin-based scaffold with a plasma-polymerized silicone membrane. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Thermal loads and their effect on integrity of mechanical systems and components

International Nuclear Information System (INIS)

Koenig, G.; Schoeckle, F.

2010-01-01

The initial step to establish a required quality status of systems and components is performed during the state of design. Main goal of the design is to consider every possible damage mechanism of the future operation (by specification of loads, medium and environment and the selection of the materials). The knowledge during the state of design determines the reliability of the component. Regarding the thermal loads, especially, only global parameters are specified usually (transients of flow and temperature connected to specified operation). These global transients are analyzed according to the standards. In operation, the safety (integrity) resp. remaining life of a component is determined by the real operation history. As experience showed, failures, defects and not specified (new) loads were discovered during operation, e.g. stratification effects in feedwater pipes and in surge lines or thermal effects in the region of valves due to switching or internal leakage. Standard surveillance in operation is performed using plant transducers that can only monitor global loads. However, problems usually are of local nature. Thermal loads like - turbulent temperatures due to mixing of media with different temperatures - temperature differences across shells or in regions of nozzles/thermal sleeves - temperature differences in piping cross sections (local and global stratification effects) - temperature differences along sections of piping systems have to be monitored by use of local instrumentation. During analysis, both the local loads and construction details have to be considered, in detail, using appropriate calculation / analysis tools. The complexity of the loads requires a comprehensive procedure: - determine the types of loads resulting from measured temperature transients - perform sensitivity studies to identify the load type that results in relevant stresses - evaluate the stresses of the significant loads - assess these stresses according to component

Design and Development of a PC- Based temperature monitoring ...

African Journals Online (AJOL)

The design of the work involves a circuit that measures the surrounding temperature using appropriate sensors and the sensor output is then converted to digital signals after due processing and conditioning of the signals. There is also an interface circuit configured to make it compatible with the PC hardware. This design ...

Progress report on the varying temperature experiment

Energy Technology Data Exchange (ETDEWEB)

Qualls, A.L.; Hurst, M.T.; Raby, D.G. [Oak Ridge National Lab., TN (United States)] [and others

1997-08-01

A capsule has been designed that permits four specimen sets to be irradiated in an RB* location in the High Flux Isotope reactor (HFIR) with distinct temperature histories. During the reporting period critical component prototyping was completed. The results have lead to some design and operational changes from that previously reported. The primary design changes are (1) compression seals in the specimen holes of the beryllium holders, and (2) oxide-dispersion strengthened aluminum alloy (DISPAL) specimen sleeves in all holders. Details of the capsule design are presented in the previous issue of this publication. Four, axially displaced temperature zones are independently controlled. Holder temperatures are monitored by thermocouples and controlled by a combination of adjustable temperature control gas mixtures and auxiliary heaters. The high temperature holders are located in the center of the experimental region, which is centered on the reactor mid-plane, and the low temperature holders are located at the ends of the experimental region.

Progress report on the varying temperature experiment

International Nuclear Information System (INIS)

Qualls, A.L.; Hurst, M.T.; Raby, D.G.

1997-01-01

A capsule has been designed that permits four specimen sets to be irradiated in an RB* location in the High Flux Isotope reactor (HFIR) with distinct temperature histories. During the reporting period critical component prototyping was completed. The results have lead to some design and operational changes from that previously reported. The primary design changes are (1) compression seals in the specimen holes of the beryllium holders, and (2) oxide-dispersion strengthened aluminum alloy (DISPAL) specimen sleeves in all holders. Details of the capsule design are presented in the previous issue of this publication. Four, axially displaced temperature zones are independently controlled. Holder temperatures are monitored by thermocouples and controlled by a combination of adjustable temperature control gas mixtures and auxiliary heaters. The high temperature holders are located in the center of the experimental region, which is centered on the reactor mid-plane, and the low temperature holders are located at the ends of the experimental region

Reactor System Design

International Nuclear Information System (INIS)

Chi, S. K.; Kim, G. K.; Yeo, J. W.

2006-08-01

SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

Optimal design of multi-state weighted k-out-of-n systems based on component design

International Nuclear Information System (INIS)

Li Wei; Zuo, Ming J.

2008-01-01

This paper presents a study on design optimization of multi-state weighted k-out-of-n systems. The studied system reliability model is more general than the traditional k-out-of-n system model. The system and its components are capable of assuming a whole range of performance levels, varying from perfect functioning to complete failure. A utility value corresponding to each state is used to indicate the corresponding performance level. A widely studied reliability optimization problem is the 'component selection problem', which involves selection of components with known reliability and cost characteristics. Less adequately addressed has been the problem of determining system cost and utility based on the relationships between component reliability, cost and utility. This paper addresses this topic. All the optimization problems dealt with in this paper can be categorized as either minimizing the expected total system cost subject to system reliability requirements, or maximizing system reliability subject to total system cost limitation. The resulting optimization problems are too complicated to be solved by traditional optimization approaches; therefore, genetic algorithm (GA) is used to solve them. Our results show that GA is a powerful tool for solving these kinds of problems

Design and Testing of Improved Spacesuit Shielding Components

International Nuclear Information System (INIS)

Ware, J.; Ferl, J.; Wilson, J.W.; Clowdsley, M.S.; DeAngelis, G.; Tweed, J.; Zeitlin, C.J.

2002-01-01

In prior studies of the current Shuttle Spacesuit (SSA), where basic fabric lay-ups were tested for shielding capabilities, it was found that the fabric portions of the suit give far less protection than previously estimated due to porosity and non-uniformity of fabric and LCVG components. In addition, overall material transmission properties were less than optimum. A number of alternate approaches are being tested to provide more uniform coverage and to use more efficient materials. We will discuss in this paper, recent testing of new material lay-ups/configurations for possible use in future spacesuit designs

The scope of additive manufacturing in cryogenics, component design, and applications

NARCIS (Netherlands)

Stautner, W.; Vanapalli, S.; Weiss, K.-P.; Chen, R.; Amm, K.; Budesheim, E.; Ricci, J.

2017-01-01

Additive manufacturing techniques using composites or metals are rapidly gaining momentum in cryogenic applications. Small or large, complex structural components are now no longer limited to mere design studies but can now move into the production stream thanks to new machines on the market that

Effect of design factors on surface temperature and wear in disk brakes

Science.gov (United States)

Santini, J. J.; Kennedy, F. E.; Ling, F. F.

1976-01-01

The temperatures, friction, wear and contact conditions that occur in high energy disk brakes are studied. Surface and near surface temperatures were monitored at various locations in a caliper disk brake during drag type testing, with friction coefficient and wear rates also being determined. The recorded transient temperature distributions in the friction pads and infrared photographs of the rotor disk surface both showed that contact at the friction surface was not uniform, with contact areas constantly shifting due to nonuniform thermal expansion and wear. The effect of external cooling and of design modifications on friction, wear and temperatures was also investigated. It was found that significant decreases in surface temperature and in wear rate can be achieved without a reduction in friction either by slotting the contacting face of the brake pad or by modifying the design of the pad support to improve pad compliance. Both design changes result in more uniform contact conditions on the friction surface.

Basics of Low-temperature Refrigeration

CERN Document Server

Alekseev, A.

2014-07-17

This chapter gives an overview of the principles of low temperature refrigeration and the thermodynamics behind it. Basic cryogenic processes - Joule-Thomoson process, Brayton process as well as Claude process - are described and compared. A typical helium laboratory refrigerator based on Claude process is used as a typical example of a low-temperature refrigeration system. A description of the hardware components for helium liquefaction is an important part of this paper, because the design of the main hardware components (compressors, turbines, heat exchangers, pumps, adsorbers, etc.) provides the input for cost calculation, as well as enables to estimate the reliability of the plant and the maintenance expenses. All these numbers are necessary to calculate the economics of a low temperature application.

Basics of Low-temperature Refrigeration

Energy Technology Data Exchange (ETDEWEB)

Alekseev, A [Linde AG, Munich (Germany)

2014-07-01

This chapter gives an overview of the principles of low temperature refrigeration and the thermodynamics behind it. Basic cryogenic processes - Joule-Thomoson process, Brayton process as well as Claude process - are described and compared. A typical helium laboratory refrigerator based on Claude process is used as a typical example of a low-temperature refrigeration system. A description of the hardware components for helium liquefaction is an important part of this paper, because the design of the main hardware components (compressors, turbines, heat exchangers, pumps, adsorbers, etc.) provides the input for cost calculation, as well as enables to estimate the reliability of the plant and the maintenance expenses. All these numbers are necessary to calculate the economics of a low temperature application.

Embedded Sensors and Controls to Improve Component Performance and Reliability: Conceptual Design Report

Energy Technology Data Exchange (ETDEWEB)

Kisner, Roger A [ORNL; Melin, Alexander M [ORNL; Burress, Timothy A [ORNL; Fugate, David L [ORNL; Holcomb, David Eugene [ORNL; Wilgen, John B [ORNL; Miller, John M [ORNL; Wilson, Dane F [ORNL; Silva, Pamela C [ORNL; Whitlow, Lynsie J [ORNL; Peretz, Fred J [ORNL

2012-10-01

The overall project objective is to demonstrate improved reliability and increased performance made possible by deeply embedding instrumentation and controls (I&C) in nuclear power plant components. The project is employing a highly instrumented canned rotor, magnetic bearing, fluoride salt pump as its I&C technology demonstration vehicle. The project s focus is not primarily on pump design, but instead is on methods to deeply embed I&C within a pump system. However, because the I&C is intimately part of the basic millisecond-by-millisecond functioning of the pump, the I&C design cannot proceed in isolation from the other aspects of the pump. The pump will not function if the characteristics of the I&C are not embedded within the design because the I&C enables performance of the basic function rather than merely monitoring quasi-stable performance. Traditionally, I&C has been incorporated in nuclear power plant (NPP) components after their design is nearly complete; adequate performance was obtained through over-design. This report describes the progress and status of the project and provides a conceptual design overview for the embedded I&C pump.

Embedded DAQ System Design for Temperature and Humidity Measurement

Directory of Open Access Journals (Sweden)

Tarique Rafique Memon

2016-05-01

Full Text Available In this work, we have proposed a cost effective DAQ (Data Acquisition system design useful for local industries by using user friendly LABVIEW (Laboratory Virtual Instrumentation Electronic Workbench. The proposed system can measure and control different industrial parameters which can be presented in graphical icon format. The system design is proposed for 8-channels, whereas tested and recorded for two parameters i.e. temperature and RH (Relative Humidity. Both parameters are set as per upper and lower limits and controlled using relays. Embedded system is developed using standard microcontroller to acquire and process the analog data and plug-in for further processing using serial interface with PC using LABVIEW. The designed system is capable of monitoring and recording the corresponding linkage between temperature and humidity in industrial unit's and indicates the abnormalities within the process and control those abnormalities through relays

Finite-temperature symmetry restoration in the four-dimensional Φ4 model with four components

International Nuclear Information System (INIS)

Jansen, K.

1990-01-01

The finite-temperature symmetry restoration in the four-dimensional φ 4 theory with four components and with an infinite self-coupling is studied by means of Monte Carlo simulations on lattices with time extensions L t =4,5,6 and space extensions 12 3 -28 3 . The numerical calculations are done by means of the Wolff cluster algorithm which is very efficient for simulations near a phase transition. The numerical results are in good agreement with an improved one-loop expansion and with the 1/N-expansion, indicating that in the electroweak theory the symmetry restoration temperature T sr is about 350 GeV. (orig.)

Safety design requirements for safety systems and components of JSFR

International Nuclear Information System (INIS)

Kubo, Shigenobu; Shimakawa, Yoshio; Yamano, Hidemasa; Kotake, Shoji

2011-01-01

Safety design requirements for JSFR were summarized taking the development targets of the FaCT project and design feature of JSFR into account. The related safety principle and requirements for Monju, CRBRP, PRISM, SPX, LWRs, IAEA standards, goals of GIF, basic principle of INPRO etc. were also taken into account so that the safety design requirements can be a next-generation global standard. The development targets for safety and reliability are set based on those of FaCT, namely, ensuring safety and reliability equal to future LWR and related fuel cycle facilities. In order to achieve these targets, the defence-in-depth concept is used as the basic safety design principle. General features of the safety design requirements are 1) Achievement of higher reliability, 2) Achievement of higher inspectability and maintainability, 3) Introduction of passive safety features, 4) Reduction of operator action needs, 5) Design consideration against Beyond Design Basis Events, 6) In-Vessel Retention of degraded core materials, 7) Prevention and mitigation against sodium chemical reactions, and 8) Design against external events. The current specific requirements for each system and component are summarized taking the basic design concept of JSFR into account, which is an advanced loop-type large-output power plant with a mixed-oxide-fuelled core. (author)

Temperature field analysis of single layer TiO2 film components induced by long-pulse and short-pulse lasers.

Science.gov (United States)

Wang, Bin; Zhang, Hongchao; Qin, Yuan; Wang, Xi; Ni, Xiaowu; Shen, Zhonghua; Lu, Jian

2011-07-10

To study the differences between the damaging of thin film components induced by long-pulse and short-pulse lasers, a model of single layer TiO(2) film components with platinum high-absorptance inclusions was established. The temperature rises of TiO(2) films with inclusions of different sizes and different depths induced by a 1 ms long-pulse and a 10 ns short-pulse lasers were analyzed based on temperature field theory. The results show that there is a radius range of inclusions that corresponds to high temperature rises. Short-pulse lasers are more sensitive to high-absorptance inclusions and long-pulse lasers are more easily damage the substrate. The first-damage decision method is drawn from calculations. © 2011 Optical Society of America

Component Cooling Heat Exchanger Heat Transfer Capability Operability Monitoring

International Nuclear Information System (INIS)

Mihalina, M.; Djetelic, N.

2010-01-01

The ultimate heat sink (UHS) is of highest importance for nuclear power plant safe and reliable operation. The most important component in line from safety-related heat sources to the ultimate heat sink water body is a component cooling heat exchanger (CC Heat Exchanger). The Component Cooling Heat Exchanger has a safety-related function to transfer the heat from the Component Cooling (CC) water system to the Service Water (SW) system. SW systems throughout the world have been the root of many plant problems because the water source, usually river, lake, sea or cooling pond, are conductive to corrosion, erosion, biofouling, debris intrusion, silt, sediment deposits, etc. At Krsko NPP, these problems usually cumulate in the summer period from July to August, with higher Sava River (service water system) temperatures. Therefore it was necessary to continuously evaluate the CC Heat Exchanger operation and confirm that the system would perform its intended function in accordance with the plant's design basis, given as a minimum heat transfer rate in the heat exchanger design specification sheet. The Essential Service Water system at Krsko NPP is an open cycle cooling system which transfers heat from safety and non-safety-related systems and components to the ultimate heat sink the Sava River. The system is continuously in operation in all modes of plant operation, including plant shutdown and refueling. However, due to the Sava River impurities and our limited abilities of the water treatment, the system is subject to fouling, sedimentation buildup, corrosion and scale formation, which could negatively impact its performance being unable to satisfy its safety related post accident heat removal function. Low temperature difference and high fluid flows make it difficult to evaluate the CC Heat Exchanger due to its specific design. The important effects noted are measurement uncertainties, nonspecific construction, high heat transfer capacity, and operational specifics (e

Computational Design of Multi-component Bio-Inspired Bilayer Membranes

Directory of Open Access Journals (Sweden)

Evan Koufos

2014-04-01

Full Text Available Our investigation is motivated by the need to design bilayer membranes with tunable interfacial and mechanical properties for use in a range of applications, such as targeted drug delivery, sensing and imaging. We draw inspiration from biological cell membranes and focus on their principal constituents. In this paper, we present our results on the role of molecular architecture on the interfacial, structural and dynamical properties of bio-inspired membranes. We focus on four lipid architectures with variations in the head group shape and the hydrocarbon tail length. Each lipid species is composed of a hydrophilic head group and two hydrophobic tails. In addition, we study a model of the Cholesterol molecule to understand the interfacial properties of a bilayer membrane composed of rigid, single-tail molecular species. We demonstrate the properties of the bilayer membranes to be determined by the molecular architecture and rigidity of the constituent species. Finally, we demonstrate the formation of a stable mixed bilayer membrane composed of Cholesterol and one of the phospholipid species. Our approach can be adopted to design multi-component bilayer membranes with tunable interfacial and mechanical properties. We use a Molecular Dynamics-based mesoscopic simulation technique called Dissipative Particle Dynamics that resolves the molecular details of the components through soft-sphere coarse-grained models and reproduces the hydrodynamic behavior of the system over extended time scales.

Application of Box-Behnken Design and Response Surface Methodology for Surface Roughness Prediction Model of CP-Ti Powder Metallurgy Components Through WEDM

Science.gov (United States)

Das, Arunangsu; Sarkar, Susenjit; Karanjai, Malobika; Sutradhar, Goutam

2018-04-01

The present work was undertaken to investigate and characterize the machining parameters (such as surface roughness, etc.) of uni-axially pressed commercially pure titanium sintered powder metallurgy components. Powder was uni-axially pressed at designated pressure of 840 MPa to form cylindrical samples and the green compacts were sintered at 0.001 mbar for about 4 h with sintering temperature varying from 1350 to 1450 °C. The influence of the sintering temperature, pulse-on and pulse-off time at wire-EDM on the surface roughness of the preforms has been investigated thoroughly. Experiments were conducted under different machining parameters in a CNC operated wire-cut EDM. The surface roughness of the machined surface was measured and critically analysed. The optimum surface roughness was achieved under the conditions of 6 μs pulse-on time, 9 μs pulse-off time and at sintering temperature of 1450 °C.

Application of Box-Behnken Design and Response Surface Methodology for Surface Roughness Prediction Model of CP-Ti Powder Metallurgy Components Through WEDM

Science.gov (United States)

Das, Arunangsu; Sarkar, Susenjit; Karanjai, Malobika; Sutradhar, Goutam

2017-06-01

The present work was undertaken to investigate and characterize the machining parameters (such as surface roughness, etc.) of uni-axially pressed commercially pure titanium sintered powder metallurgy components. Powder was uni-axially pressed at designated pressure of 840 MPa to form cylindrical samples and the green compacts were sintered at 0.001 mbar for about 4 h with sintering temperature varying from 1350 to 1450 °C. The influence of the sintering temperature, pulse-on and pulse-off time at wire-EDM on the surface roughness of the preforms has been investigated thoroughly. Experiments were conducted under different machining parameters in a CNC operated wire-cut EDM. The surface roughness of the machined surface was measured and critically analysed. The optimum surface roughness was achieved under the conditions of 6 μs pulse-on time, 9 μs pulse-off time and at sintering temperature of 1450 °C.

Status of irradiation capsule design

International Nuclear Information System (INIS)

Nagata, Hiroshi; Yamaura, Takayuki; Nagao, Yoshiharu

2013-01-01

For the irradiation test after the restart of JMTR, further precise temperature control and temperature prediction are required. In the design of irradiation capsule, particularly sophisticated irradiation temperature prediction and evaluation are urged. Under such circumstance, among the conventional design techniques of irradiation capsule, the authors reviewed the evaluation method of irradiation temperature. In addition, for the improvement of use convenience, this study examined and improved FINAS/STAR code in order to adopt the new calculation code that enables a variety of analyses. In addition, the study on the common use of the components for radiation capsule enabled the shortening of design period. After the restart, the authors will apply this improved calculation code to the design of irradiation capsule. (A.O.)

Design of a low-cost system for electrical conductivity measurements of high temperature

Science.gov (United States)

Singh, Yadunath

2018-05-01

It is always a curiosity and interest among researchers working in the field of material science to know the impact of high temperature on the physical and transport properties of the materials. In this paper, we report on the design and working of a system for the measurements of electrical resistivity with high temperature. It was designed at our place and successively used for these measurements in the temperature range from room temperature to 500 ˚C.

Influence of the design temperature on long-term safety of a salt dome repository

International Nuclear Information System (INIS)

Buhmann, D.; Brenner, J.; Storck, R.

1993-03-01

All studies made so far within the framwork of the mixed concept system analysis proceeded from a design temperature of the mine structure of 200 C. The concept based on a design temperature of 150 C was aimed at studying whether it made sense to maintain lower temperatures, if necessary. Deterministic and probabilistic calculations were made in order to determine the influence of the lower design temperature on long-term safety. The calculations were based on concept A of Joint Borehole and Gallery Storage. Assuming reference values of the input parameters, the deterministic calculations do not produce any radionuclide release from the mine structure. If, however, one assumes a lower rate for rock convergence, radionuclides are released at maximum dose rates of about 3.10 -5 Sv/a. Even a larger volume of limited brine inclusions may lead to radionuclide releases, in that case with dose commitments of the order of magnitude of 1.10 -5 Sv/a. The probabilistic calculations show that a design temperature of 150 C for long-term safety is less favourable than a higher design temperature. The share of simulations in the probabilistic calculations with a radionuclide release, and the expected value of dose commitment, are almost double as high as in the concept based on 200 C design temperature. Thus a higher design temperature is preferable with regard to the long-term safety of a salt repository. The most important parameters concerning dose commitment are the volume of limited brine inclusions, the convergence rate, and the permeability of barriers and backfilling rock. (orig./HP) [de

Design of a low temperature district heating network with supply recirculation

DEFF Research Database (Denmark)

Li, Hongwei; Dalla Rosa, Alessandro; Svendsen, Svend

2010-01-01

The focus on continuing improving building energy efficiency and reducing building energy consumption brings the key impetus for the development of the new generation district heating (DH) system. In the new generation DH network, the supply and return temperature are designed low in order to sig...... calculates the heat loss in the twin pipe as that in the single pipe. The influence of this simplification on the supply/return water temperature prediction was analyzed by solving the coupled differential energy equations.......-pass system starts to function. The aim of this paper is to investigate the influence of by-pass water on the network return temperature and introduce the concept of supply water recirculation into the network design so that the traditional by-pass system can be avoided. Instead of mixing the by-pass water......The focus on continuing improving building energy efficiency and reducing building energy consumption brings the key impetus for the development of the new generation district heating (DH) system. In the new generation DH network, the supply and return temperature are designed low in order...

Design and R&D for manufacturing the beamline components of MITICA and ITER HNBs

Energy Technology Data Exchange (ETDEWEB)

Dalla Palma, M., E-mail: mauro.dallapalma@igi.cnr.it [Consorzio RFX, Padova (Italy); Sartori, E. [Consorzio RFX, Padova (Italy); Blatchford, P.; Chuilon, B. [CCFE, Culham Science Centre, Oxfordshire (United Kingdom); Graceffa, J. [ITER Organization, St Paul Lez Durance (France); Hanke, S. [KIT, Institute for Technical Physics, Eggenstein-Leopoldshafen (Germany); Hardie, C. [CCFE, Culham Science Centre, Oxfordshire (United Kingdom); Masiello, A. [F4E, Barcelona (Spain); Muraro, A. [Consorzio RFX, Padova (Italy); Ochoa, S. [KIT, Institute for Technical Physics, Eggenstein-Leopoldshafen (Germany); Shah, D. [ITER Organization, St Paul Lez Durance (France); Veltri, P.; Zaccaria, P.; Zaupa, M. [Consorzio RFX, Padova (Italy)

2015-10-15

Highlights: • Particle beam-component interaction was analysed developing and applying numerical codes. • Gas density distribution was calculated with AVOCADO code and applied for electrical analyses. • High heat flux components were designed, analysed with subcooled boiling, verified for fatigue. • Fracture behaviour of ceramics was analysed by finite element modelling and was verified. • R&D supports the design of the beamline components, especially for water-vacuum barriers. - Abstract: The design of the beamline components of MITICA, the full prototype of the ITER heating neutral beam injectors, is almost finalised and technical specifications for the procurement are under preparation. These components are the gas neutraliser, the electrostatic residual ion dump, and the calorimeter. Electron dump panels are foreseen each side of the upstream end of the neutraliser to protect the cryo-panels from electrons, created by stripping and other processes, that exit the 1 MeV accelerator. As the design of the components must fulfil requirements on the beam physics, insight on physical processes is required to identify performance trade-offs and constraints. The spatial gas distribution was simulated to verify the pumping requirements with electron dump panels and local conditions for breakdown voltage. Electrostatic analyses were carried out for the insulating elements of the RID to verify the limits of the electric field intensity. Different criteria were approached to investigate the fracture behaviour of ceramics considering the manufacturing implications and extrapolating the conditions for proof testing. Severe heating conditions will be applied steadily, as the maximum pulse duration is 1 h, and cyclically so requiring to fulfil fatigue and ratcheting verifications. High heat fluxes, up to 13 MW/m{sup 2} on the calorimeter, with enhanced heat transfer in subcooled boiling conditions will occur in the actively cooled CuCr1Zr panel elements provided with

Development of the interactive model between Component Cooling Water System and Containment Cooling System using GOTHIC

International Nuclear Information System (INIS)

Byun, Choong Sup; Song, Dong Soo; Jun, Hwang Yong

2006-01-01

In a design point of view, component cooling water (CCW) system is not full-interactively designed with its heat loads. Heat loads are calculated from the CCW design flow and temperature condition which is determined with conservatism. Then the CCW heat exchanger is sized by using total maximized heat loads from above calculation. This approach does not give the optimized performance results and the exact trends of CCW system and the loads during transient. Therefore a combined model for performance analysis of containment and the component cooling water(CCW) system is developed by using GOTHIC software code. The model is verified by using the design parameters of component cooling water heat exchanger and the heat loads during the recirculation mode of loss of coolant accident scenario. This model may be used for calculating the realistic containment response and CCW performance, and increasing the ultimate heat sink temperature limits

Regulatory Guide 1.122: Development of floor design response spectra for seismic design of floor-supported equipment or components

International Nuclear Information System (INIS)

Anon.

1993-01-01

''Reactor Site Criteria,'' requires, in part, that safety-related structures, systems, and components remain functional in the event of a Safe Shutdown Earthquake (SSE). It specifies the use of a suitable dynamic analysis as one method of ensuring that the structures, systems, and components can withstand the seismic loads. Similarly, paragraph (a)(2) of Section VI of the same appendix requires, in part, that the structures, systems, and components necessary for continued operation without undue risk to the health and safety of the public remain functional in the event of an Operating Basis Earthquake (OBE). Again, the use of suitable dynamic analysis is specified as one method of ensuring that the structures, systems, and components can withstand the seismic loads. This guide describes methods acceptable to the NRC staff for developing two horizontal and one vertical floor design response spectra at various floors or other equipment-support locations of interest from the time-history motions resulting from the dynamic analysis of the supporting structure. These floor design response spectra are needed for the dynamic analysis of the systems or equipment supported at various locations of the supporting structure

Advances in Human-Computer Interaction: Graphics and Animation Components for Interface Design

Science.gov (United States)

Cipolla Ficarra, Francisco V.; Nicol, Emma; Cipolla-Ficarra, Miguel; Richardson, Lucy

We present an analysis of communicability methodology in graphics and animation components for interface design, called CAN (Communicability, Acceptability and Novelty). This methodology has been under development between 2005 and 2010, obtaining excellent results in cultural heritage, education and microcomputing contexts. In studies where there is a bi-directional interrelation between ergonomics, usability, user-centered design, software quality and the human-computer interaction. We also present the heuristic results about iconography and layout design in blogs and websites of the following countries: Spain, Italy, Portugal and France.

Design and Implementation of High Precision Temperature Measurement Unit

Science.gov (United States)

Zeng, Xianzhen; Yu, Weiyu; Zhang, Zhijian; Liu, Hancheng

2018-03-01

Large-scale neutrino detector requires calibration of photomultiplier tubes (PMT) and electronic system in the detector, performed by plotting the calibration source with a group of designated coordinates in the acrylic sphere. Where the calibration source positioning is based on the principle of ultrasonic ranging, the transmission speed of ultrasonic in liquid scintillator of acrylic sphere is related to temperature. This paper presents a temperature measurement unit based on STM32L031 and single-line bus digital temperature sensor TSic506. The measurement data of the temperature measurement unit can help the ultrasonic ranging to be more accurate. The test results show that the temperature measurement error is within ±0.1°C, which satisfies the requirement of calibration source positioning. Take energy-saving measures, with 3.7V/50mAH lithium battery-powered, the temperature measurement unit can work continuously more than 24 hours.

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-03-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-01-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

A Designed Room Temperature Multilayered Magnetic Semiconductor

Science.gov (United States)

Bouma, Dinah Simone; Charilaou, Michalis; Bordel, Catherine; Duchin, Ryan; Barriga, Alexander; Farmer, Adam; Hellman, Frances; Materials Science Division, Lawrence Berkeley National Lab Team

2015-03-01

A room temperature magnetic semiconductor has been designed and fabricated by using an epitaxial antiferromagnet (NiO) grown in the (111) orientation, which gives surface uncompensated magnetism for an odd number of planes, layered with the lightly doped semiconductor Al-doped ZnO (AZO). Magnetization and Hall effect measurements of multilayers of NiO and AZO are presented for varying thickness of each. The magnetic properties vary as a function of the number of Ni planes in each NiO layer; an odd number of Ni planes yields on each NiO layer an uncompensated moment which is RKKY-coupled to the moments on adjacent NiO layers via the carriers in the AZO. This RKKY coupling oscillates with the AZO layer thickness, and it disappears entirely in samples where the AZO is replaced with undoped ZnO. The anomalous Hall effect data indicate that the carriers in the AZO are spin-polarized according to the direction of the applied field at both low temperature and room temperature. NiO/AZO multilayers are therefore a promising candidate for spintronic applications demanding a room-temperature semiconductor.

Engineering design and thermal hydraulics of plasma facing components of SST-1

International Nuclear Information System (INIS)

Pragash, N. Ravi; Chaudhuri, P.; Santra, P.; Chenna Reddy, D.; Khirwadkar, S.; Saxena, Y.C.

2001-01-01

SST-1 is a medium size tokamak with super conducting magnetic field coils. All the subsystems of SST-1 are designed for quasi steady state (∼1000 s) operation. Plasma Facing Components (PFCs) of SST-1 consisting of divertors, passive stabilizers, baffles and poloidal limiters are also designed to be compatible for steady state operation. As SST-1 is designed to run double null divertor plasmas, these components also have up-down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. All the PFC are made of copper alloys (CuCrZr and CuZr) on which graphite tiles are mechanically attached. These copper alloy back plates are actively cooled with water flowing in the channels grooved on them with the main consideration in the design of PFCs as the steady state heat removal of about 1.0 MW/m 2 . In addition to be able to remove high heat fluxes, the PFCs are also designed to be compatible for baking at 350 degree sign C. Extensive studies, involving different flow parameters and various cooling layouts, have been done to select the final cooling parameters and layout. Thermal response of the PFCs and vacuum vessel during baking, has been calculated using a FORTRAN code and a 2-D finite element analysis. The PFCs and their supports are also designed to withstand large electro-magnetic forces. Finite element analysis using ANSYS software package is used in this and other PFCs design. The engineering design including thermal hydraulics for cooling and baking of all the PFCs is completed. Poloidal limiters are being fabricated. The remaining PFCs, viz. divertors, stabilizers and baffles are likely to go for fabrication in the next few months. The detailed engineering design, the finite element calculations in the structural and thermal designs are presented in this paper

Design and analysis on fume exhaust system of blackbody cavity sensor for continuously measuring molten steel temperature

Directory of Open Access Journals (Sweden)

Guohui Mei

2017-03-01

Full Text Available Fume exhaust system is the main component of the novel blackbody cavity sensor with a single layer tube, which removes the fume by gas flow along the exhaust pipe to keep the light path clean. However, the gas flow may break the conditions of blackbody cavity and results in the poor measurement accuracy. In this paper, we analyzed the influence of the gas flow on the temperature distribution of the measuring cavity, and then calculated the integrated effective emissivity of the non-isothermal cavity based on Monte-Carlo method, accordingly evaluated the sensor measurement accuracy, finally obtained the maximum allowable flow rate for various length of the exhaust pipe to meet the measurement accuracy. These results will help optimize the novel blackbody cavity sensor design and use it better for measuring the temperature of molten steel.

Advancements in the design of safety-related systems and components of the MARS nuclear plant

International Nuclear Information System (INIS)

Caira, M.; Caruso, G.; Naviglio, A.; Sorabella, L.; Farello, C.E.

1992-01-01

In the paper, the advancements in the design of safety-related systems and components of the MARS nuclear plant, equipped with a 600 MW th PWR, are described. These advancements are due to the special safety features of this plant, which relies completely on inherent and passive safety. In particular, the new steps of the design of the innovative, completely passive, and with an unlimited autonomy Emergency core Cooling System are described, together with the characteristics of the last version of the steam generator, developed in a new design involving disconnecting components, for a fast erection and an easy maintenance. (author)

Tests of qualification of national components of nuclear power plants under design basis accident

International Nuclear Information System (INIS)

Mesquita, A.Z.

1990-01-01

With the purpose of qualifying national components of nuclear power plants, whose working must be maintained during and after an accident, the Thermohydraulic Division of CDTN have done tests to check the equipment stability, under Design Basis Accident conditions. Until this moment, the following components were tested: electrical junction boxes (connectors); coating systems for wall, inside cover and steel containment; hydraulics components of personnel and equipment airlock. This work describes the test instalation, the tests performed and its results. The components tested, in a general way, fulfil the specified requirements. (author) [pt

Design characteristics of a three-component AEOI Neutriran Albedo Neutron Personnel Dosimeter

International Nuclear Information System (INIS)

Sohrabi, M.; Katouzi, M.

1991-01-01

In establishing a national personnel neutron dosimetry service in Iran, different parameters of the AEOI Neutriran Albedo Neutron Personnel Dosimeter (NANPD) have been optimized. A NANPD was designed with three dosimetry components to measure (a) direct thermal neutrons, (b) direct fast neutrons and (C) direct neutrons by the detection of the albedo neutrons reflected from the body. The dosimeter consists of one or more Lexan polycarbonate and/or CR-39 foils and two 10 B (n,α) 7 Li converters in a cadmium cover so arranged as to efficiently measure the three neutron dose components separately. The boron converter thickness, its position relative to the beam direction and its distance from the PC foil were studied and the results were incorporated into the design. The dose response of the dosimeter, its lower detection limit as well as the correction factors related to the field neutrons and albedo neutrons were also determined for a 238 Pu-Be, an 241 Am-Be and a 252 Cf sources. In this paper, the dosimeter design and its dosimetric characteristics are presented and discussed. (author)

Oxidation damage evaluation by non-destructive method for graphite components in high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Shibata, Taiju; Tada, Tatsuya; Sumita, Junya; Sawa, Kazuhiro

2008-01-01

To develop non-destructive evaluation methods for oxidation damage on graphite components in High Temperature Gas-cooled Reactors (HTGRs), the applicability of ultrasonic wave and micro-indentation methods were investigated. Candidate graphites, IG-110 and IG-430, for core components of Very High Temperature Reactor (VHTR) were used in this study. These graphites were oxidized uniformly by air at 500degC. The following results were obtained from this study. (1) Ultrasonic wave velocities with 1 MHz can be expressed empirically by exponential formulas to burn-off, oxidation weight loss. (2) The porous condition of the oxidized graphite could be evaluated with wave propagation analysis with a wave-pore interaction model. It is important to consider the non-uniformity of oxidized porous condition. (3) Micro-indentation method is expected to determine the local oxidation damage. It is necessary to assess the variation of the test data. (author)

Safety design philosophy of gas turbine high temperature reactor (GTHTR300)

International Nuclear Information System (INIS)

Katanishi, Shoji; Kunitomi, Kazuhiko

2003-01-01

Japan Atomic Energy Research Institute (JAERI) has been developing design studies of the Gas Turbine High Temperature Reactor (GTHTR300). The original safety design philosophy has also been discussed and fixed for the GTHTR300 based on the experience of the High Temperature Engineering Test Reactor (HTTR) of JAERI which is the first High Temperature Gas-cooled Reactor (HTGR) in Japan. One of the unique feature of the safety philosophy of the GTHTR300 is that a depressurization accident induced by a large pipe break is postulated as a design basis accident in order to show the high level of safety characteristics, though its probability of occurrence is lower than the probability range of design basis accident. Another feature of safety design is to adopt a double confinement that is one of the original concepts for the GTHTR300. By using a double confinement, a feasibility of safety design without containment vessel was clarified even in case of the depressurization accident. The safety design philosophies for passive cooling system, reactor shutdown system, and so on were determined. The methodology for the safety evaluation, such as safety criteria and selection of events to be evaluated by using estimation of probability of occurrence, were also discussed and determined. This article describes the safety design philosophy and some results of preliminary evaluations which were conducted in order to clarify the feasibility of original safety design of the GTHTR300. The present study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

Implementation of a design and configuration management platform for fusion components on the Tore Supra WEST Project

Energy Technology Data Exchange (ETDEWEB)

Benoît, Fabrice, E-mail: fabrice-2.benoit@cea.fr [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Allegretti, Ludovic [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Aumeunier, Marie-Hélène [OPTIS, ZE de La Farlède, F-83078 Toulon Cedex 9 (France); Bucalossi, Jérôme; Doceul, Louis; Faïsse, Frederic; Firdaouss, Medhi; Geynet, Michel; Houtte, Didier van; Larroque, Sébastien; Magaud, Philippe; Maini, Patrick; Missirlian, Marc; Parrat, Hélène [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Robert, Julien [SOFYNE, F-69800 Saint Priest (France)

2014-10-15

Highlights: •A design and configuration management platform is under development for managing fusion components lifecycle at CEA. •Design platform ensures an efficient sharing of the data and provides connections between the different software and databases involved in fusion components design. •Design platform rollout on WEST project is ongoing as part of change control and configuration management implementation. -- Abstract: This paper presents the technical solutions and methodologies that are used and under development for managing the design lifecycle of the WEST project (W – for tungsten – Environment in Steady-state Tokamak, upgrade of Tore Supra's with actively cooled tungsten plasma facing components) fusion components and explains the interfaces that are implemented or in construction to connect together the different tools like documents management system, CAD modeler, or simulation codes around the data management backbone. It describes the methodologies used on the WEST project to optimize the design process by managing the engineering data workflow and ensuring the consistency between the different 3D representations for design or analysis as well as the specification or interfaces documents. Finally it explains how this platform contributes to reach the project targets in terms of performance, cost and schedule.

Implementation of a design and configuration management platform for fusion components on the Tore Supra WEST Project

International Nuclear Information System (INIS)

Benoît, Fabrice; Allegretti, Ludovic; Aumeunier, Marie-Hélène; Bucalossi, Jérôme; Doceul, Louis; Faïsse, Frederic; Firdaouss, Medhi; Geynet, Michel; Houtte, Didier van; Larroque, Sébastien; Magaud, Philippe; Maini, Patrick; Missirlian, Marc; Parrat, Hélène; Robert, Julien

2014-01-01

Highlights: •A design and configuration management platform is under development for managing fusion components lifecycle at CEA. •Design platform ensures an efficient sharing of the data and provides connections between the different software and databases involved in fusion components design. •Design platform rollout on WEST project is ongoing as part of change control and configuration management implementation. -- Abstract: This paper presents the technical solutions and methodologies that are used and under development for managing the design lifecycle of the WEST project (W – for tungsten – Environment in Steady-state Tokamak, upgrade of Tore Supra's with actively cooled tungsten plasma facing components) fusion components and explains the interfaces that are implemented or in construction to connect together the different tools like documents management system, CAD modeler, or simulation codes around the data management backbone. It describes the methodologies used on the WEST project to optimize the design process by managing the engineering data workflow and ensuring the consistency between the different 3D representations for design or analysis as well as the specification or interfaces documents. Finally it explains how this platform contributes to reach the project targets in terms of performance, cost and schedule

Structural mechanics research and development for main components of chinese 300 MWe PWR NPPs: from design to life management

International Nuclear Information System (INIS)

Yao Weida; Dou Yikang; Xie Yongcheng; He Yinbiao; Zhang Ming; Liang Xingyun

2005-01-01

Qinshan Nuclear Power Plant (Unit I), is a 300 MWe prototype PWR independently developed by Chinese own efforts, from design, manufacture, construction, installation, commissioning, to operation, inspection, maintenance, ageing management and lifetime assessment. Shanghai Nuclear Engineering Research and Design Institute (SNERDI) has taken up with and involved in deeply the R and D to tackle problems of this type of reactor since very beginning in early 1970s. Structural mechanics is one of the important aspects to ensure the safety and reliability for NPP components. This paper makes a summary on role of structural mechanics for component safety and reliability assessment in different stages of design, commissioning, operation, as well as lifetime assessment on this type PWR NPPs, including Qinshan-I and Chashma-I, a sister plant in Pakistan designed by SNERDI. The main contents of the paper cover design by analysis for key components of NSSS; mechanical problems relating to safety analysis; special problems relating to pressure retaining components, such as fracture mechanics, sealing analysis and its test verifications, etc.; experimental research on flow-induced vibration; seismic qualification for components; component failure diagnosis and root cause analysis; vibration qualification and diagnosis technique; component online monitoring technique; development of defect assessment; methodology of aging management and lifetime assessment for key components of NPPs, etc. (authors)

Development of low temperature solid state joining technology of dissimilar for nuclear heat exchanger tube components

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-08-15

By conventional fusion welding process (TIG), a realization of reliable and sound joints for the nuclear heat exchanger components is very difficult, especially for the parts comprising of the dissimilar metal couples (Ti-STS, Ti-Cu alloy etc.). This is mainly attributed to the formation of brittle intermetallics (Ti{sub x}Cu{sub y}, Ti{sub x}Fe{sub y}, Ti{sub x}Ni{sub y} etc.) and wide difference in physical properties. Moreover, it usually employs very high thermal input, so making it difficult to obtain sound joints due to generations of high residual stresses and degradation of the adjacent base metals, even for similar metal combinations. In this project, the low temperature solid-state joining technology was established by developing new alloy fillers, e.g. the multi-component eutectic based alloys or amorphous alloys, and thereby lowering the joining temperature down to {approx}800 .deg. C without affecting the structural properties of base metals. Based on a low temperature joining, the interlayer engineering technology was then developed to be able to eliminate the brittleness of the joints for strong Ti-STS dissimilar joints, and the diffusion brazing technology of Ti-Ti with a superior joining strength and corrosion-resistance comparable to those of base metal were developed. By using those developed technologies, the joining procedures feasible for the heat exchanger components were finally established for the dissimilar metal joints including Ti tube sheet to super STS tube, Ti tube sheet to super STS tube sheet, and the joints of the Ti tube to Ti tube sheet

Design of Plasma Facing Components for Superconducting Modification of JT-60

International Nuclear Information System (INIS)

Shinji Sakurai; Kei Masaki; Yusuke-Kudo Shibama; Hiroshi Tamai; Makoto Matsukawa; Cordier, J.J.

2006-01-01

JT-60 is planning to modify the machine as a fully superconducting coil tokamak (JT-60 Super Advanced, the former JT-60SC and NCT) to establish scientific and technological bases for an economically and environmentally attractive DEMO reactor. It will be also a satellite tokamak in a part of broader approach for ITER. It is designed for high beta (betaN = 3.5-5.5) and steady-state research in a break-even class DD plasma for 100 s or longer. Nominal plasma parameters are I p =5.5 MA, B t =2.7 T, R=3.01 m, a=1.14 m with double-null configuration. An ITER-like single-null configuration with I p =3.5 MA, B t =2.6 T can be also operated. In order to study the ITER-relevant high confinement plasma with high density, designed plasma heating power was enhanced from 25 MW to 41 MW for 100 s through the design review with EU and Japan. The heat flux onto outer divertor target exceeds 10 MW/m 2 with moderate radiative fraction of 50-60% in single-null configuration. Therefore, the ITER-like mono-block CFC target will be adopted to aim at power handling of 15 MW/m 2 . A cooling water system should be reinforced 3 times from original design for double null divertor with high coolant flow velocity of ∼10 m/s. The peak heat flux onto the neutral beam armor for perpendicular injected positive NB is evaluated to be 2 MW/m 2 , which needs to be actively water-cooled. A bolt-fixed CFC tile was tested at the heat flux of 1-3 MW/m 2 and will be applied to the NB armor. In order to improve plasma beta value by enhancing wall stabilization effect, passive-stabilizing plates, which are electrically and mechanically connected in poloidal and toroidal direction, will be installed near the plasma surface (r wall /a=1.1-1.3) at the outboard side. Stabilizing plate has double-wall ribbed structure and can be operated at 573 K with heating nitrogen gas instead of cooling water between double walls. It has crank-type support legs to allow thermal expansion at high temperature operation. The

Conceptual benefits of passive nuclear power plants and their effect on component design

International Nuclear Information System (INIS)

DeVine, J.C. Jr.

1996-01-01

Today, nearly ten years after the advanced light water reactor (ALWR) Program was conceived by US utility leaders, and a decade and a half since a new nuclear power plant was ordered in the US, the ALWR passive plant is coming into its own. This design concept, a midsized simplified light water reactor, features extremely reliable passive systems for accident prevention and mitigation and combines proven experience with state-of-the-art engineering and human factors. It is now emerging as the front runner to become the next generation reactor in the US and perhaps around the world. Although simple and straightforward in concept, the passive plant is in many respects a significant departure from previous trends in reactor engineering. Successful implementation of this concept presents numerous challenges to the designers of passive plant systems and components. This paper provides a brief history of the ALWR program, it outlines the ALWR passive plant design objectives and principles, and it summarizes with examples their implications on component design. (orig.)