Test facility for fast gas injections into a vessel filled with water

International Nuclear Information System (INIS)

Wilhelm, D.; Kirstahler, M.

1987-11-01

The Fast Gas Injection Facility (SGI) was set up to study the hydrodynamics during the expansion of a gas bubble into a vessel filled with water. The gas stored in a pressure vessel expands against gravity through a circular duct into a large cylindrical vessel partly with water. This report covers the description of the test facility and the data acquisition. Results of the first test series are added. (orig.) [de

Analysis and Design of Cryogenic Pressure Vessels for Automotive Hydrogen Storage

Science.gov (United States)

Espinosa-Loza, Francisco Javier

Cryogenic pressure vessels maximize hydrogen storage density by combining the high pressure (350-700 bar) typical of today's composite pressure vessels with the cryogenic temperature (as low as 25 K) typical of low pressure liquid hydrogen vessels. Cryogenic pressure vessels comprise a high-pressure inner vessel made of carbon fiber-coated metal (similar to those used for storage of compressed gas), a vacuum space filled with numerous sheets of highly reflective metalized plastic (for high performance thermal insulation), and a metallic outer jacket. High density of hydrogen storage is key to practical hydrogen-fueled transportation by enabling (1) long-range (500+ km) transportation with high capacity vessels that fit within available spaces in the vehicle, and (2) reduced cost per kilogram of hydrogen stored through reduced need for expensive structural material (carbon fiber composite) necessary to make the vessel. Low temperature of storage also leads to reduced expansion energy (by an order of magnitude or more vs. ambient temperature compressed gas storage), potentially providing important safety advantages. All this is accomplished while simultaneously avoiding fuel venting typical of cryogenic vessels for all practical use scenarios. This dissertation describes the work necessary for developing and demonstrating successive generations of cryogenic pressure vessels demonstrated at Lawrence Livermore National Laboratory. The work included (1) conceptual design, (2) detailed system design (3) structural analysis of cryogenic pressure vessels, (4) thermal analysis of heat transfer through cryogenic supports and vacuum multilayer insulation, and (5) experimental demonstration. Aside from succeeding in demonstrating a hydrogen storage approach that has established all the world records for hydrogen storage on vehicles (longest driving range, maximum hydrogen storage density, and maximum containment of cryogenic hydrogen without venting), the work also

33 CFR 165.1151 - Security Zones; liquefied hazardous gas tank vessels, San Pedro Bay, California.

Science.gov (United States)

2010-07-01

... a tank vessel as liquefied petroleum gas, liquefied natural gas, or similar liquefied gas products... Eleventh Coast Guard District § 165.1151 Security Zones; liquefied hazardous gas tank vessels, San Pedro... the sea floor, within a 500 yard radius around any liquefied hazardous gas (LHG) tank vessel that is...

Application of radioisotope tracer techniques in evaluation of irradiation vessel of flue gas treatment system

International Nuclear Information System (INIS)

Joon-Ha Jin; Myun-Joo Lee; Sung-Hee Jung; Young-Chang Nho

1998-01-01

The proper design of the irradiation vessel of electron beam flue gases treatment plant and resultant optimum gas flow pattern is a very important factor to get a high removal efficiency of toxic materials from flue gases. Radioisotope tracer experiments were conducted to study the residence time distribution of gas flow in a cylindrical irradiation vessel. A few mCi of gaseous radioisotope tracer Ar-41 was injected to the upstream of the vessel and the input and output response were measured with two NaI scintillation detectors. The same experiment was conducted after the modification of the vessel by introducing 4 baffles. The experimental data were analyzed to calculate mean residence times and mixing characteristics of each system using the residence time distribution (RTD) analysis software. A method to estimate pollutant removal efficiencies of an irradiation vessel from the residence time distributions measured by radiotracer experiments was suggested. The analytical results were compared to evaluate the effect of the baffles on the removal efficiency of the plant

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)

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.)

Heating and cooling system for an on-board gas adsorbent storage vessel

Science.gov (United States)

Tamburello, David A.; Anton, Donald L.; Hardy, Bruce J.; Corgnale, Claudio

2017-06-20

In one aspect, a system for controlling the temperature within a gas adsorbent storage vessel of a vehicle may include an air conditioning system forming a continuous flow loop of heat exchange fluid that is cycled between a heated flow and a cooled flow. The system may also include at least one fluid by-pass line extending at least partially within the gas adsorbent storage vessel. The fluid by-pass line(s) may be configured to receive a by-pass flow including at least a portion of the heated flow or the cooled flow of the heat exchange fluid at one or more input locations and expel the by-pass flow back into the continuous flow loop at one or more output locations, wherein the by-pass flow is directed through the gas adsorbent storage vessel via the by-pass line(s) so as to adjust an internal temperature within the gas adsorbent storage vessel.

Safety Research Experiment Facility Project. Conceptual design report. Volume V. Reactor vessel and closure

International Nuclear Information System (INIS)

1975-12-01

The Prestressed Concrete Reactor Vessel (PCRV) will serve as the primary pressure retaining structure for the Safety Research Experiment Facility (SAREF) reactor. The reactor core, control rod drive room, primary heat exchangers, and gas circulators will be located in cavities within the PCRV. The orientation of these cavities, except for the control rod drive room, will be similar to the high-temperature gas-cooled reactor (HTGR) designs that are currently proposed or under design. Due to the nature of this type of structure, all biological and radiological shielding requirements are incorporated into the basic vessel design. At the midcore plane there are three radially oriented slots that will extend from the outside surface of the PCRV to the reactor core liner. These slots will accommodate each of the fuel motion monitoring systems which will be part of the observation apparatus used with the loop experiments

Equation of state of an ideal gas with nonergodic behavior in two connected vessels.

Science.gov (United States)

Naplekov, D M; Semynozhenko, V P; Yanovsky, V V

2014-01-01

We consider a two-dimensional collisionless ideal gas in the two vessels connected through a small hole. One of them is a well-behaved chaotic billiard, another one is known to be nonergodic. A significant part of the second vessel's phase space is occupied by an island of stability. In the works of Zaslavsky and coauthors, distribution of Poincaré recurrence times in similar systems was considered. We study the gas pressure in the vessels; it is uniform in the first vessel and not uniform in second one. An equation of the gas state in the first vessel is obtained. Despite the very different phase-space structure, behavior of the second vessel is found to be very close to the behavior of a good ergodic billiard but of different volume. The equation of state differs from the ordinary equation of ideal gas state by an amendment to the vessel's volume. Correlation of this amendment with a share of the phase space under remaining intact islands of stability is shown.

CFD simulation of gas-liquid floating particles mixing in an agitated vessel

Directory of Open Access Journals (Sweden)

Li Liangchao

2017-01-01

Full Text Available Gas dispersion and floating particles suspension in an agitated vessel were studied numerically by using computational fluid dynamics (CFD. The Eulerian multi-fluid model along with standard k-ε turbulence model was used in the simulation. A multiple reference frame (MRF approach was used to solve the impeller rotation. The velocity field, gas and floating particles holdup distributions in the vessel were first obtained, and then, the effects of operating conditions on gas dispersion and solid suspension were investigated. The simulation results show that velocity field of solid phase and gas phase are quite different in the agitated vessel. Floating particles are easy to accumulate in the center of the surface region and the increasing of superficial gas velocity is in favor of floating particles off-surface suspension. With increasing solids loading, the gas dispersion becomes worse, while relative solid holdup distribution changes little. The limitations of the present modeling are discussed and further research in the future is proposed.

Design of vacuum vessel for Indian Test Facility (INTF) for 100 keV neutral beams

International Nuclear Information System (INIS)

Joshi, Jaydeep; Yadav, Ashish; Gangadharan, Roopesh; Prasad, Rambilas; Ulahannan, Shino; Rotti, Chandramouli; Bandyopadhyay, Mainak; Chakraborty, Arun

2015-01-01

Highlights: • Thickness calculation and optimization for the main shell, ducts, Dishends and top lid on the main shell. • Nozzle and flange design for the port openings. • Support structure design for the main shell and ducts. • FEA validation of the INTF vessel for operational, seismic and lifting condition. - Abstract: The Indian Test Facility (INTF) vacuum vessel is designed to install a full-scale test set-up of Diagnostic Neutral Beam (DNB) [1] for the qualification of beam parameters and the behavior of beam-line components prior to installation and operation in ITER. Vacuum vessel is designed in cylindrical shape having length of ∼9 m with diameter of ∼4.5 m and has a detachable top-lid for mounting as well as removal of internal components during installation and maintenance phases. The Vessel has hemispherical dish-ends with large openings for high-voltage bushing on one side and duct on another side. Vessel is provided with openings for hydraulic, cryo, gas-feed and diagnostics. Vessel duct is composed of three segments with length ranges from 3 m to 5 m with diameter of ∼1.5 m and one vessel at the end to house the second calorimeter. The objective of this paper is to present the design and analysis of vacuum vessel, with respect to its functional and operational requirements. The design calculations are done as per ASME-BPVC SectionVIII-Div.1 and subsequently Finite Element Analysis (FEM) method has been adopted to verify the design.

Design of vacuum vessel for Indian Test Facility (INTF) for 100 keV neutral beams

Energy Technology Data Exchange (ETDEWEB)

Joshi, Jaydeep, E-mail: Jaydeep.joshi@iter-india.org [ITER-India, Institute for Plasma Research, A29, GIDC Electronics Estate, Gandhinagar 382016, Gujarat (India); Yadav, Ashish; Gangadharan, Roopesh [ITER-India, Institute for Plasma Research, A29, GIDC Electronics Estate, Gandhinagar 382016, Gujarat (India); Prasad, Rambilas [Madan Mohan Malaviya University of Technology, Gorakhpur, Uttar Pradesh 273001 (India); Ulahannan, Shino [Airframe Aerodesigns Pvt. Ltd., HAL Airport Exit Road, Old Airport Road, Bengaluru 17 (India); Rotti, Chandramouli; Bandyopadhyay, Mainak; Chakraborty, Arun [ITER-India, Institute for Plasma Research, A29, GIDC Electronics Estate, Gandhinagar 382016, Gujarat (India)

2015-10-15

Highlights: • Thickness calculation and optimization for the main shell, ducts, Dishends and top lid on the main shell. • Nozzle and flange design for the port openings. • Support structure design for the main shell and ducts. • FEA validation of the INTF vessel for operational, seismic and lifting condition. - Abstract: The Indian Test Facility (INTF) vacuum vessel is designed to install a full-scale test set-up of Diagnostic Neutral Beam (DNB) [1] for the qualification of beam parameters and the behavior of beam-line components prior to installation and operation in ITER. Vacuum vessel is designed in cylindrical shape having length of ∼9 m with diameter of ∼4.5 m and has a detachable top-lid for mounting as well as removal of internal components during installation and maintenance phases. The Vessel has hemispherical dish-ends with large openings for high-voltage bushing on one side and duct on another side. Vessel is provided with openings for hydraulic, cryo, gas-feed and diagnostics. Vessel duct is composed of three segments with length ranges from 3 m to 5 m with diameter of ∼1.5 m and one vessel at the end to house the second calorimeter. The objective of this paper is to present the design and analysis of vacuum vessel, with respect to its functional and operational requirements. The design calculations are done as per ASME-BPVC SectionVIII-Div.1 and subsequently Finite Element Analysis (FEM) method has been adopted to verify the design.

On the Adequacy of API 521 Relief-Valve Sizing Method for Gas-Filled Pressure Vessels Exposed to Fire

Directory of Open Access Journals (Sweden)

Anders Andreasen

2018-03-01

Full Text Available In this paper, the adequacy of the legacy API 521 guidance on pressure relief valve (PRV sizing for gas-filled vessels subjected to external fire is investigated. Multiple studies show that in many cases, the installation of a PRV offers little or no protection—therefore provides an unfounded sense of security. Often the vessel wall will be weakened by high temperatures, before the PRV relieving pressure is reached. In this article, a multiparameter study has been performed taking into consideration various vessel sizes, design pressures (implicitly vessel wall thickness, vessel operating pressure, fire type (pool fire or jet fire by applying the methodology presented in the Scandpower guideline. A transient thermomechanical response analysis has been carried out to accurately determine vessel rupture times. It is demonstrated that only vessels with relatively thick walls, as a result of high design pressures, benefit from the presence of a PRV, while for most cases no appreciable increase in the vessel survival time beyond the onset of relief is observed. For most of the cases studied, vessel rupture will occur before the relieving pressure of the PRV is reached.

Polar vessel hullform design based on the multi-objective optimization NSGA II

Directory of Open Access Journals (Sweden)

DUAN Fei

2017-12-01

Full Text Available [Objectives] With the increasing exploitation of the Arctic abundant oil and gas resources, a large number of ships which meet the polar navigational requirements are needed.[Methods] In this paper, the fast elitist Non-Dominated Sorting Genetic Algorithm (NSGA Ⅱ is applied to the hull optimization, and the multi-objective optimization method of polar vessel design is proposed. With the optimization goal of resistance and icebreaking resistance, filtering hull forms through the standard of polar vessel displacement and EEDI, fast ship hull optimization that satisfy the ice-ship dead weight and EEDI requirements has been achieved. Taking a 65 000 t shuttle tanker as an example, full parametric modeling method is adopted, the hull optimization of three different bow forms is conducted through the polar vessel multi-objective optimization method.[Results] The ship hull after optimization can satisfy the IA class navigation require, where the resistance in calm water decreases up to 12.94%, and the minimum propulsion power in ice field has a 27.36% reduction.[Conclusions] The feasibility and validity of the NSGA Ⅱ applying in polar vessel design is verified.

Standard practice for examination of Gas-Filled filament-wound composite pressure vessels using acoustic emission

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 This practice provides guidelines for acoustic emission (AE) examination of filament-wound composite pressure vessels, for example, the type used for fuel tanks in vehicles which use natural gas fuel. 1.2 This practice requires pressurization to a level equal to or greater than what is encountered in normal use. The tanks' pressurization history must be known in order to use this practice. Pressurization medium may be gas or liquid. 1.3 This practice is limited to vessels designed for less than 690 bar [10,000 psi] maximum allowable working pressure and water volume less than 1 m3 or 1000 L [35.4 ft3]. 1.4 AE measurements are used to detect emission sources. Other nondestructive examination (NDE) methods may be used to gain additional insight into the emission source. Procedures for other NDE methods are beyond the scope of this practice. 1.5 This practice applies to examination of new and in-service filament-wound composite pressure vessels. 1.6 This practice applies to examinations conducted at amb...

Computational analysis of transient gas release from a high pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Pedro, G.; Oshkai, P.; Djilali, N. [Victoria Univ., BC (Canada). Inst. for Integrated Energy Systems; Penau, F. [CERAM Euro-American Inst. of Technology, Sophia Antipolis (France)

2006-07-01

Gas jets exiting from compressed vessels can undergo several regimes as the pressure in the vessel decreases, and a greater understanding of the characteristics of gas jets is needed to determine safety requirements in the transport, distribution, and use of hydrogen. This paper provided a study of the bow shock waves that typically occur during the initial stage of a gas jet incident. The transient behaviour of an initiated jet was investigated using unsteady, compressible flow simulations. The gas was considered to be ideal, and the domain was considered to be axisymmetric. Tank pressure for the analysis was set at a value of 100 atm. Jet structure was examined, as well as the shock structures and separation due to adverse pressure gradients at the nozzle. Shock structure displacement was also characterized.

Reactor pressure vessel design

International Nuclear Information System (INIS)

Foehl, J.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 2, the general principles of reactor pressure vessel design are elaborated. Crack and fracture initiation and propagation are treated in some detail

The evolution and structural design of prestressed concrete pressure vessels

International Nuclear Information System (INIS)

Hannah, I.W.

1978-01-01

The introduction of the prestressed concrete pressure vessel to contain the main gas coolant circuit of nuclear reactors has marked a major step forward. This chapter traces the evolution and development of the PCPV, and lists the principal parameters adopted. Current design and loading standards are discussed in relation to the two main limit states of serviceability and safety. Prestressed concrete pressure vessel analysis has called for very extensive adaptation and expansion of conventional finite element and finite difference methods in order to deal with the elevated temperature of operation, together with extensive concrete testing at temperature and under multi-directional stressing. These new methods and extra data are being adopted in prestressed applications in other fields and may well prove to be of much wider significance than is presently appreciated. (author)

Hydrogen Gas Retention and Release from WTP Vessels: Summary of Preliminary Studies

Energy Technology Data Exchange (ETDEWEB)

Gauglitz, Phillip A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bontha, Jagannadha R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Daniel, Richard C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mahoney, Lenna A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rassat, Scot D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wells, Beric E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bao, Jie [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Boeringa, Gregory K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buchmiller, William C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Burns, Carolyn A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chun, Jaehun [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Karri, Naveen K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Li, Huidong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Tran, Diana N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-07-01

The Hanford Waste Treatment and Immobilization Plant (WTP) is currently being designed and constructed to pretreat and vitrify a large portion of the waste in the 177 underground waste storage tanks at the Hanford Site. A number of technical issues related to the design of the pretreatment facility (PTF) of the WTP have been identified. These issues must be resolved prior to the U.S. Department of Energy (DOE) Office of River Protection (ORP) reaching a decision to proceed with engineering, procurement, and construction activities for the PTF. One of the issues is Technical Issue T1 - Hydrogen Gas Release from Vessels (hereafter referred to as T1). The focus of T1 is identifying controls for hydrogen release and completing any testing required to close the technical issue. In advance of selecting specific controls for hydrogen gas safety, a number of preliminary technical studies were initiated to support anticipated future testing and to improve the understanding of hydrogen gas generation, retention, and release within PTF vessels. These activities supported the development of a plan defining an overall strategy and approach for addressing T1 and achieving technical endpoints identified for T1. Preliminary studies also supported the development of a test plan for conducting testing and analysis to support closing T1. Both of these plans were developed in advance of selecting specific controls, and in the course of working on T1 it was decided that the testing and analysis identified in the test plan were not immediately needed. However, planning activities and preliminary studies led to significant technical progress in a number of areas. This report summarizes the progress to date from the preliminary technical studies. The technical results in this report should not be used for WTP design or safety and hazards analyses and technical results are marked with the following statement: “Preliminary Technical Results for Planning – Not to be used for WTP Design

An Approach to Understanding Cohesive Slurry Settling, Mobilization, and Hydrogen Gas Retention in Pulsed Jet Mixed Vessels

Energy Technology Data Exchange (ETDEWEB)

Gauglitz, Phillip A.; Wells, Beric E.; Fort, James A.; Meyer, Perry A.

2009-05-22

The Hanford Waste Treatment and Immobilization Plant (WTP) is being designed and built to pretreat and vitrify a large portion of the waste in Hanford’s 177 underground waste storage tanks. Numerous process vessels will hold waste at various stages in the WTP. Some of these vessels have mixing-system requirements to maintain conditions where the accumulation of hydrogen gas stays below acceptable limits, and the mixing within the vessels is sufficient to release hydrogen gas under normal conditions and during off-normal events. Some of the WTP process streams are slurries of solid particles suspended in Newtonian fluids that behave as non-Newtonian slurries, such as Bingham yield-stress fluids. When these slurries are contained in the process vessels, the particles can settle and become progressively more concentrated toward the bottom of the vessels, depending on the effectiveness of the mixing system. One limiting behavior is a settled layer beneath a particle-free liquid layer. The settled layer, or any region with sufficiently high solids concentration, will exhibit non-Newtonian rheology where it is possible for the settled slurry to behave as a soft solid with a yield stress. In this report, these slurries are described as settling cohesive slurries.

Design and construction of Alborz tokamak vacuum vessel system

International Nuclear Information System (INIS)

Mardani, M.; Amrollahi, R.; Koohestani, S.

2012-01-01

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

H_2 production by the steam reforming of excess boil off gas on LNG vessels

International Nuclear Information System (INIS)

Fernández, Ignacio Arias; Gómez, Manuel Romero; Gómez, Javier Romero; López-González, Luis M.

2017-01-01

Highlights: • BOG excess in LNG vessels is burned in the GCU without energy use. • The gas management plants need to be improved to increase efficiency. • BOG excess in LNG vessels is used for H_2 production by steam reforming. • The availability of different fuels increases the versatility of the ship. - Abstract: The gas management system onboard LNG (Liquid Natural Gas) vessels is crucial, since the exploitation of the BOG (Boil Off Gas) produced is of utmost importance for the overall efficiency of the plant. At present, LNG ships with no reliquefaction plant consume the BOG generated in the engines, and the excess is burned in the GCU (Gas Combustion Unit) without any energy use. The need to improve the gas management system, therefore, is evident. This paper proposes hydrogen production through a steam reforming plant, using the excess BOG as raw material and thus avoiding it being burned in the GCU. To test the feasibility of integrating the plant, an actual study of the gas management process on an LNG vessel with 4SDF (4 Stroke Dual Fuel) propulsion and with no reliquefaction plant was conducted, along with a thermodynamic simulation of the reforming plant. With the proposed gas management system, the vessel disposes of different fuels, including H_2, a clean fuel with zero ozone-depleting emissions. The availability of H_2 on board in areas with strict anti-pollution regulations, such as ECAs (Emission Control Area), means that the vessel may be navigated without using fossil fuels which generate CO_2 and SO_X emissions. Moreover, while at port, Cold Ironing is avoided, which entails high costs. Thus it is demonstrated that the installation of a reforming plant is both energetically viable and provides greater versatility to the ship.

The high integrity design and manufacture of the Heysham II/Torness gas baffle

International Nuclear Information System (INIS)

Armor, J.; Day, B.V.; White, C.M.

1985-01-01

The AGR design used on the Heysham II and Torness power stations requires a gas baffle which is essentially a steel pressure vessel for which one can demonstrate a high degree of integrity. The design, analytical, manufacturing, erection and testing processes which were undertaken to achieve the standard required of the completed assembly are discussed. To this end the vessels were manufactured in purpose-made shops and transported to site, leaving a minimum amount of work to be undertaken at site. Subsequent evaluation has shown a very low probability of failure compared with conventional steel pressure vessels. (author)

Conceptual design of the test facility for the two-phase critical flow with non-condensable gas

Energy Technology Data Exchange (ETDEWEB)

Chang, Seok Kyu; Chung, Chang Hwan

2000-12-01

The two-phase critical flow test with non-condensible gas is for the simulation of the critical flow phenomena which can be occurred during SB-LOCA on SMART reactor. The requirements of the critical flow test are 7{approx}20mm pipe break dia., 7{approx}12MPa stagnation pressure, 0{approx}60 deg C subcooling degree and 0{approx}0.5kg/s N2 gas flow rate. For the satisfaction of these requirements on the test facility, critical flow rates were calculated with various models. With the selected reference pressure vessel(1.3m{sup 3}), the conceptual design of the test facility was performed. The important components of the test facility are the pressure vessel which has main circulation line, the test section attached to the bottom of the pressure vessel, suppression tank, the N2 gas supply tanks for maintaining the system pressure and N2 gas flow rate at test section and the auxiliary N2 gas converting system. For the measurements of the critical flow rate, flowmeter and level gauge is installed at the upstream of the test section and the pressure vessel, respectively. The realtime pressure control system is installed at the entrance of the pressure vessel for maintaining the system pressure and the N2 gas flow regulating system is also installed at the upstream of the test section. The design of the control and monitoring system for the operation of the test facility and the DAS for acquiring the test data were also performed. The conceptual operating process of the test facility was determined.

Conceptual design of the test facility for the two-phase critical flow with non-condensable gas

International Nuclear Information System (INIS)

Chang, Seok Kyu; Chung, Chang Hwan

2000-12-01

The two-phase critical flow test with non-condensible gas is for the simulation of the critical flow phenomena which can be occurred during SB-LOCA on SMART reactor. The requirements of the critical flow test are 7∼20mm pipe break dia., 7∼12MPa stagnation pressure, 0∼60 deg C subcooling degree and 0∼0.5kg/s N2 gas flow rate. For the satisfaction of these requirements on the test facility, critical flow rates were calculated with various models. With the selected reference pressure vessel(1.3m 3 ), the conceptual design of the test facility was performed. The important components of the test facility are the pressure vessel which has main circulation line, the test section attached to the bottom of the pressure vessel, suppression tank, the N2 gas supply tanks for maintaining the system pressure and N2 gas flow rate at test section and the auxiliary N2 gas converting system. For the measurements of the critical flow rate, flowmeter and level gauge is installed at the upstream of the test section and the pressure vessel, respectively. The realtime pressure control system is installed at the entrance of the pressure vessel for maintaining the system pressure and the N2 gas flow regulating system is also installed at the upstream of the test section. The design of the control and monitoring system for the operation of the test facility and the DAS for acquiring the test data were also performed. The conceptual operating process of the test facility was determined

Containment vessel design and practice

International Nuclear Information System (INIS)

Bangash, Y.

1983-01-01

The state of the art of analysis and design of the concrete containment vessels required for BWR and PWR is reviewed. A step-by-step critical appraisal of the existing work is given. Elastic, inelastic and cracking conditions under extreme loads are fully discussed. Problems associated with these structures are highlighted. A three-dimensional finite element analysis is included to cater for service, overload and dynamic cracking of such structures. Missile impact and seismic effects are included in this work. The second analysis is known as the limit state analysis, which is given to design such vessels for any kind of load. (U.K.)

Design of vessel baking system and thermal radiation shields for SST-1

International Nuclear Information System (INIS)

Kumar, E.R.; Nagabhushana, S.; Pathak, H.A.; Panigrahi, S.; Nath, T.R.; Babu, A.V.S; Gangradey, R.; Patel, R.J.; Saxena, Y.C.

1998-01-01

SST-1 is a Steady State Tokamak with a major radius of 1.1 m, minor radius of 0.2 m and toroidal field of 3.0 T. The toroidal and poloidal field coils of SST-1 are superconducting. One of the main objectives of SST-1 is to demonstrate steady state particle removal and active plasma density control which states the necessity of wall conditioning. The vacuum vessel will be baked up to 525 K by passing hot nitrogen gas through the U - channels welded on the inner surface of vacuum vessel. The required mass flow rate at 5 bar is 0.712 Kg/s to maintain 525 K wall temperature in steady state. Superconducting coils operating at 4.5 K will be protected against thermal radiation from hot surfaces using liquid nitrogen cooled panels operating at 87 K. Maximum 1200 litres/hour liquid nitrogen is required during vessel baking. The design of vacuum vessel baking system and thermal radiation shields and related flow analysis are presented here. (authors)

Design of vessel baking system and thermal radiation shields for SST-1

Energy Technology Data Exchange (ETDEWEB)

Kumar, E.R.; Nagabhushana, S.; Pathak, H.A.; Panigrahi, S.; Nath, T.R.; Babu, A.V.S; Gangradey, R.; Patel, R.J.; Saxena, Y.C. [Institute for Plasma Research, Gandhinagar (India)

1998-07-01

SST-1 is a Steady State Tokamak with a major radius of 1.1 m, minor radius of 0.2 m and toroidal field of 3.0 T. The toroidal and poloidal field coils of SST-1 are superconducting. One of the main objectives of SST-1 is to demonstrate steady state particle removal and active plasma density control which states the necessity of wall conditioning. The vacuum vessel will be baked up to 525 K by passing hot nitrogen gas through the U - channels welded on the inner surface of vacuum vessel. The required mass flow rate at 5 bar is 0.712 Kg/s to maintain 525 K wall temperature in steady state. Superconducting coils operating at 4.5 K will be protected against thermal radiation from hot surfaces using liquid nitrogen cooled panels operating at 87 K. Maximum 1200 litres/hour liquid nitrogen is required during vessel baking. The design of vacuum vessel baking system and thermal radiation shields and related flow analysis are presented here. (authors)

Design and development of in-vessel viewing periscope for ITER (International Thermonuclear Experimental Reactor)

International Nuclear Information System (INIS)

Obara, Kenjiro; Kakudate, Satoshi; Ito, Akira; Shibanuma, Kiyoshi; Tada, Eisuke

1999-02-01

An in-vessel viewing system is essential not only to detect and locate damage of components exposed to plasma, but also to monitor and assist in-vessel maintenance operation. In ITER, the in-vessel viewing system must be capable of operating at high temperature (200degC), under intense gamma radiation (30 kGy/h) and high vacuum or 1 bar inert gas. A periscope-type in-vessel viewing system has been chosen as a reference of the ITER in-vessel viewing system due to its wide viewing capability and durability for sever environments. According to the ITER research and development program, a full-scale radiation hard periscope with a length of 15 m has been successfully developed by the Japan Home Team. The performance tests have been shown sufficient capability at high temperature up to 250degC and radiation resistance over 100 MGy. This report describes the design and R and D results of the ITER in-vessel viewing periscope based on the development of 15-m-length radiation hard periscope. (author)

Conceptual design studies of in-vessel viewing equipment for ITER

International Nuclear Information System (INIS)

Obara, Kenjiro; Oka, Kiyoshi; Taguchi, Hiroshi; Itoh, Akira; Tada, Eisuke; Shibanuma, Kiyoshi

1996-03-01

In-vessel viewing systems are essential to inspect all surface of in-vessel components so as to detect and locate damages, and to assist in-vessel maintenance operations. The in-vessel viewing operations are categorized into the three cases, which are 1) rapid inspection just after off-normal events such as disruption, 2) scheduled inspection, and 3) supplementary inspection during maintenance operations. In case of the rapid inspection, the viewing systems have to be operated in vacuum (ca. 10 -5 Pa) and high temperature (ca. 300degC) under a gamma ray dose rate of 10 7 R/h. On the other hand, the latter two cases are anticipated to be under atmospheric inert gas, 150degC and 3x10 6 R/h. Accordingly, the in-vessel viewing systems are required to have sufficient durability under those conditions of all cases as well as precision of the vision to all of in-vessel surface. Based on those requirements, scoping studies on various viewing concepts have been performed and the applicability to the ITER conditions have been assessed. As a result, two types of viewing systems have been chosen, which are a periscope type viewing system and a image fiber type viewing system with a multi-joint manipulator. Both systems are based on radiation hard optical elements which are being developed. In this report, the design features of both viewing systems are described, including technical issues for ITER application. Finally, a periscope type viewing system is recommended as a primary system and the following specifications/conditions are proposed for the further engineering design. (1) Unified type periscope with a movable mirror at the tip (2) Integrated lighting device into the periscope (3) Accessed from top vertical ports located at 7.3m from the machine center (4) Proposed configuration with a total length of around 27m and a diameter of 200mm. (author)

Automatic design of prestressed concrete vessels

International Nuclear Information System (INIS)

Sotomura, Kentaro; Murazumi, Yasuyuki

1984-01-01

Prestressed concrete appeared after high strnegth steel had been produced, therefore it has the history of only 40 years even in Europe where it was developed. High compressive force is given to concrete beforehand by high strength steel to resist tensile force. It is superior to ordinary steel in strength, economy, rust prevention, fire protection and workability, and it competes with ordinary steel in the fields of bridges, towers, water tanks, water pipes, barges, LPG and LNG tanks, reactor pressure vessels, reactor containment vessels and so on. The design of prestressed concrete containment vessels (PCCV) being constructed in Japan adopts the form of mounting a semi-spherical dome on a cylindrical wall of 43m inside diameter and about 1.5m thickness, and the steel pipe sheaths for inserting tendons are arranged in the wall. The Taisei Construction Co. has developed the PC-ADE system which enables the optimum design of PCCVs. The outline of the automatic design system, the design of tendon arrangement, the preparation of the data on the load for stress analysis, the stress analysis by axisymmetric finite element method and the calculation of cross sections are explained. Design is a creative activity, and in the design of PCCVs also, the intention of designers should be materialized when this program is utilized. (Kako, I.)

Organic Iodine Adsorption by AgZ under Prototypical Vessel Off-Gas Conditions

International Nuclear Information System (INIS)

Bruffey, Stephanie H.; Jubin, Robert Thomas; Jordan, J. A.

2016-01-01

U.S. regulations will require the removal of 129 I from the off-gas streams of any used nuclear fuel (UNF) reprocessing plant prior to discharge of the off-gas to the environment. Multiple off-gas streams within a UNF reprocessing plant combine prior to release, and each of these streams contains some amount of iodine. For an aqueous UNF reprocessing plant, these streams include the dissolver off-gas, the cell off-gas, the vessel off-gas (VOG), the waste off-gas and the shear off-gas. To achieve regulatory compliance, treatment of multiple off-gas streams within the plant must be performed. Preliminary studies have been completed on the adsorption of I 2 onto silver mordenite (AgZ) from prototypical VOG streams. The study reported that AgZ did adsorb I 2 from a prototypical VOG stream, but process upsets resulted in an uneven feed stream concentration. The experiments described in this document both improve the characterization of I 2 adsorption by AgZ from dilute gas streams and further extend it to include characterization of the adsorption of organic iodides (in the form of CH 3 I) onto AgZ under prototypical VOG conditions. The design of this extended duration testing was such that information about the rate of adsorption, the penetration of the iodine species, and the effect of sorbent aging on iodine removal in VOG conditions could be inferred.

Organic Iodine Adsorption by AgZ under Prototypical Vessel Off-Gas Conditions

Energy Technology Data Exchange (ETDEWEB)

Bruffey, Stephanie H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jordan, J. A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-09-30

U.S. regulations will require the removal of ¹²⁹I from the off-gas streams of any used nuclear fuel (UNF) reprocessing plant prior to discharge of the off-gas to the environment. Multiple off-gas streams within a UNF reprocessing plant combine prior to release, and each of these streams contains some amount of iodine. For an aqueous UNF reprocessing plant, these streams include the dissolver off-gas, the cell off-gas, the vessel off-gas (VOG), the waste off-gas and the shear off-gas. To achieve regulatory compliance, treatment of multiple off-gas streams within the plant must be performed. Preliminary studies have been completed on the adsorption of I₂ onto silver mordenite (AgZ) from prototypical VOG streams. The study reported that AgZ did adsorb I₂ from a prototypical VOG stream, but process upsets resulted in an uneven feed stream concentration. The experiments described in this document both improve the characterization of I₂ adsorption by AgZ from dilute gas streams and further extend it to include characterization of the adsorption of organic iodides (in the form of CH₃I) onto AgZ under prototypical VOG conditions. The design of this extended duration testing was such that information about the rate of adsorption, the penetration of the iodine species, and the effect of sorbent aging on iodine removal in VOG conditions could be inferred.

Confinement Vessel Assay System: Design and Implementation Report

International Nuclear Information System (INIS)

Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Mayo, Douglas R.; Gomez, Cipriano D.; Miko, David K.; Salazar, William R.; Stange, Sy; Vigil, Georgiana M.

2012-01-01

Los Alamos National Laboratory has a number of spherical confinement vessels remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1- to 2-inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the vessels. We have developed a neutron assay system for the purposes of Materials Control and Accountability (MC and A) measurements of the vessel prior to and after cleanout. We present our approach to confronting the challenges in designing, building, and testing such a system. The system was designed to meet a set of functional and operational requirements. A Monte Carlo model was developed to aid in optimizing the detector design as well as to predict the systematic uncertainty associated with confinement vessel measurements. Initial testing was performed to optimize and determine various measurement parameters, and then the system was characterized using 252 Cf placed a various locations throughout the measurement system. Measurements were also performed with a 252 Cf source placed inside of small steel and HDPE shells to study the effect of moderation. These measurements compare favorably with their MCNPX model equivalent, making us confident that we can rely on the Monte Carlo simulation to predict the systematic uncertainty due to variations in response to material that may be localized at different points within a vessel.

Device for removing hydrogen gas from the safety containment vessel of a nuclear reactor

International Nuclear Information System (INIS)

Stiefel, M.

1983-01-01

The safe processing of all concentrations of gas mixtures should be possible with such a device using a thermal recombiner of compact construction. A recombiner consisting of a metal case and diverter sheets situated in it is heated by induction. The incoming pipe for the gas mixture enriched with hydrogen and the outgoing pipe for the gas mixture with low hydrogen content are connected together by a three way valve. The third connection to the safety valve takes the larger port of the gas mixture with low hydrogen content back to the safety containment vessel. Sufficient amount of the gas mixture with low hydrogen content is taken via the three way valve to the safety containment vessel to ensure that the hydrogen content of the gas mixture taken to the recombiner remains below the 4% by volume limit. (orig./PW)

Design and analysis of concrete reactor vessels: New developments, problems and trends

International Nuclear Information System (INIS)

Bazant, Z.P.

1984-01-01

This lecture reviews new developments in analysis and design of prestressed concrete reactor vessels (PCRV). After a brief assessment of the current status and experience, the advantages, disadvantages, and especially the safety features of PCRV, are discussed. Attention is then focused on the design of penetrations and openings, and on the design for high-temperature resistance - areas in which further developments are needed. Various possible designs for high-temperature exposure of concrete in a hypothetical accident are analyzed. Considered are not only PCRVs for gas-cooled reactors (GCR), but also guard vessels for liquid metal fast breeder reactors (LMFBR), for which designs mitigating the adverse effects of molten sodium, molten steel, and core melt are surveyed. Realistic analysis of the problems requires further development in the knowledge of material behavior and its mathematical modeling. Recent advances in the modeling of high-temperature response of concrete, including pore water transfer, pore pressure, creep and shrinkage are outlined. This is followed by a discussion of new developments in the analysis of cracking of concrete, where the need of switching from stress criteria to energy criteria for fracture is emphasized. The lecture concludes with a brief discussion of long-time behavior, the effect of aging, and probabilistic analysis of creep. (orig.)

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

Design and development of the CRBRP ex-vessel transfer machine

International Nuclear Information System (INIS)

Jones, C.E. Jr.

1977-01-01

The Reactor Refueling System (RRS) for the Clinch River Breeder Reactor Project (CRBRP) uses the Ex-Vessel Transfer Machine (EVTM) for transferring core assemblies outside the reactor vessel. The design of the Ex-Vessel Transfer Machine (EVTM) and its gantry-trolly for the CRBRP is discussed. The development tests required for the design are presented, in conjunction with the impact of the test results on the design. The impact of the increased seismic requirements on the design are also presented

37 CFR 212.5 - Recordation of distinctive identification of vessel hull designer.

Science.gov (United States)

2010-07-01

... identification of vessel hull designer. 212.5 Section 212.5 Patents, Trademarks, and Copyrights COPYRIGHT OFFICE, LIBRARY OF CONGRESS COPYRIGHT OFFICE AND PROCEDURES PROTECTION OF VESSEL HULL DESIGNS § 212.5 Recordation of distinctive identification of vessel hull designer. (a) General. Any owner of a vessel hull may...

Thermal radiation from fireballs on failure of liquefied petroleum gas storage vessels

Energy Technology Data Exchange (ETDEWEB)

Roberts, T.; Hawksworth, S. [Health and Safety Executive, Health and Safety Lab., Buxton (United Kingdom); Gosse, A. [BG Technology, Loughborough (United Kingdom)

2000-05-01

Fire impingement on vessels containing pressure liquefied gases can result in catastrophic failure of the vessel leading to a Boiling Liquid Expanding Vapour Explosion (BLEVE). If the gas is flammable, this can result in the formation of very large fireballs. In safety assessments where catastrophic vessel failure is identified as a real possibility, the risk of death from a fireball tends to be higher than that from missiles or blast. Since many of the physical processes which take place in a BLEVE are scale dependent, a series of tests were undertaken at a large scale where 2 tonne propane vessels were taken to failure in a jet fire and the vessel response, mode of failure and consequence of failure characterised. The measurements taken by the Health and Safety Laboratory and BG Technology relating to fireball formation are described. (Author)

Basic design of the test facility for the two-phase critical flow with non-condensable gas

International Nuclear Information System (INIS)

Chang, Seok Kyu; Kim, Chang Hwe; Chung, Chang Hwan

2000-12-01

The two-phase critical flow test with non-condensible gas is for the simulation of the critical flow phenomena which can be occurred during SB-LOCA on SMART reactor. The basic design of the test facility for the actual installation is performed from the basis of the previous conceptual design according to the test requirements. The 1.3m 3 pressure vessel has the circulation pipeline which contains pump(5m 3 /hr), main heater(150KW) and cooler for heating the working fluid to the test temperature within 6 hours. The N2 gas, water supply line are attached to the upper part and test section, flowmeter and various sensors are installed at the lower part of the pressure vessel. The suppression tank is for the storage and cooling of the discharged water. The N2 gas storage tank provides the system pressure to the pressure vessel during the test. The 0.7m 3 N2 gas injection tank supplies the required N2 gas to the entrance of the test section. Since these N2 supply systems require much amount of gas during short period, multistage valve systems and optimal control logics are needed and applied. For the filling of the N2 gas to the N2 storage tank, 5m 3 LN2 tank and related gas converting system were designed. The operating mode of the test facility can be classified to the starting, steady, main test and cooling modes and the proper monitoring and control logics are developed for each operating mode. The operation of the test facility is performed through the PLC and the acquisition of the test data is done with DAS

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.)

The velocity of missiles generated by the disintegration of gas-pressurized vessels and pipes

International Nuclear Information System (INIS)

Baum, M.R.

1984-01-01

A theoretical model is developed to describe the velocity of fragments generated when a gas-pressurized vessel disintegrates. The predictions are compared with new and existing experimental data for spherical and cylindrical vessels and are shown to be an improvement over the widely used empirical correlation developed by Moore. It is also shown that, by an appropriate definition of the energy available for doing work on the fragments, the velocity of the fragments from the disintegration of a section of gas pipeline may be predicted by the same model

The velocity of missiles generated by the disintegration of gas-pressurised vessels and pipes

International Nuclear Information System (INIS)

Baum, M.R.

1983-03-01

A theoretical model is developed to describe the velocity of fragments generated when a gas-pressurised vessel disintegrates. The predictions are compared with new and existing experimental data for spherical and cylindrical vessels and are shown to be an improvement over the widely used empirical correlation developed by Moore. It is also shown that, by an appropriate definition of the energy available for doing work on the fragments, the velocity of the fragments from the disintegration of a section of gas pipeline may be predicted by the same model. (author)

Advanced in-vessel retention design for next generation risk management

Energy Technology Data Exchange (ETDEWEB)

Suh, Kune Y; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

1998-12-31

In the TMI-2 accident, approximately twenty (20) tons of molten core material drained into the lower plenum. Early advanced light water reactor (LWR) designs assumed a lower head failure and incorporated various measures for ex-vessel accident mitigation. However,one of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100 deg C for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel. Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident/risk management strategies. As an advanced in-vessel design concept, this work presents the COrium Attack Syndrome Immunization Structures (COASIS) that are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in-vessel (COASISI) and ex-vessel (COASISO) are demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the Korean Standard Nuclear Power Plant (KSNPP) reactor. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options. 15 refs., 5 figs., 1 tab. (Author)

Advanced in-vessel retention design for next generation risk management

Energy Technology Data Exchange (ETDEWEB)

Suh, Kune Y.; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

1997-12-31

In the TMI-2 accident, approximately twenty (20) tons of molten core material drained into the lower plenum. Early advanced light water reactor (LWR) designs assumed a lower head failure and incorporated various measures for ex-vessel accident mitigation. However,one of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100 deg C for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel. Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident/risk management strategies. As an advanced in-vessel design concept, this work presents the COrium Attack Syndrome Immunization Structures (COASIS) that are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in-vessel (COASISI) and ex-vessel (COASISO) are demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the Korean Standard Nuclear Power Plant (KSNPP) reactor. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options. 15 refs., 5 figs., 1 tab. (Author)

Design study on evaluation for power conversion system concepts in high temperature gas cooled reactor with gas turbine

International Nuclear Information System (INIS)

Minatsuki, Isao; Mizokami, Yorikata

2007-01-01

The design studies on High Temperature Gas Cooled Reactor with Gas Turbine (HTGR-GT) have been performed, which were mainly promoted by Japan Atomic Energy Agency (JAEA) and supported by fabricators in Japan. HTGR-GT plant feature is almost determined by selection of power conversion system concepts. Therefore, plant design philosophy is observed characteristically in selection of them. This paper describes the evaluation and analysis of the essential concepts of the HTGR-GT power conversion system through the investigations based on our experiences and engineering knowledge as a fabricator. As a result, the following concepts were evaluated that have advantages against other competitive one, such as the horizontal turbo machine rotor, the turbo machine in an individual vessel, the turbo machine with single shaft, the generator inside the power conversion vessel, and the power conversion system cycle with an intercooler. The results of the study can contribute as reference data when the concepts will be selected. Furthermore, we addressed reasonableness about the concept selection of the Gas Turbine High Temperature Reactor GTHTR300 power conversion system, which has been promoted by JAEA. As a conclusion, we recognized the GTHTR300 would be one of the most promising concepts for commercialization in near future. (author)

Design characteristics of pantograph type in vessel fuel handling system in SFR

International Nuclear Information System (INIS)

Kim, S. H.; Koo, G. H.

2012-01-01

The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied

Design characteristics of pantograph type in vessel fuel handling system in SFR

Energy Technology Data Exchange (ETDEWEB)

Kim, S. H.; Koo, G. H. [KAERI, Daejeon (Korea, Republic of)

2012-10-15

The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied.

Design Procedure on Stud Bolt for Reactor Vessel Assembly

International Nuclear Information System (INIS)

Kim, Jong-Wook; Lee, Gyu-Mahn; Jeoung, Kyeong-Hoon; Kim, Tae-Wan; Park, Keun-Bae; Kim, Keung-Koo

2008-10-01

The reactor pressure vessel flange is welded to the upper part of reactor pressure vessel, and there are stud holes to mount the closure head with stud bolts. The surface mating the closure head is compressed with O-ring, which acts as a sealing gasket to prevent coolant leakage. Bolted flange connections perform a very important structural role in the design of a reactor pressure vessel. Their importance stems from two important functions: (a) maintenance of the structural integrity of the connection itself, and (b) prevention of leakage through the O-ring preloaded by stud bolts. In the present study, an evaluation procedure for the design of stud bolt is developed to meet ASME code requirements. The developed design procedure could provide typical references in the development of advanced reactor design in the future

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

Thermo-mechanical behaviour of FBTR reactor vessel due to natural convection in cover gas space

International Nuclear Information System (INIS)

Srinivasan, G.; Varadarajan, S.; Kapoor, R.P.

1988-01-01

Fast Breeder Test Reactor is a 40 MW(t), loop type sodium cooled reactor, similar in design to Rapsodie. The Reactor Assembly, which is the heart of FBTR, comprises the Reactor Vessel (RV) housed in a safety vessel within a concrete cell (A1 Cell). The RV which supports the core is shielded at the top by two rotatable plugs which are stacked with layers of borated graphite and steel. The smaller plug (SRP), is mounted excentric to the larger one (LRP). A nominal annular gap of 16 mm is provided between RV and LRP and between LRP and SRP to enable free rotation of the plugs. Stainless Steel insulation is fixed inside the steel vessel, to avoid overheating of the A1 Cell concrete. The core is supported by the Grid Plate (GP), bolted to the RV. During preheating, sodium charging and isothermal runs upto 350 0 C, temperature asymmetries were noticed in the reactor vessel wall in the cover gas space. This was attributable to convection currents in the annulus between RV and LRP. The asymmetries also resulted in a lateral shift of the grid plate. This paper discusses our experience in suppressing these convection currents, and minimising the grid plate shift

Design of pressure vessels. Part 1

International Nuclear Information System (INIS)

Grandemange, J.M.

2008-01-01

The equipments and loops of PWR reactors are basically pressure vessels. Their specificities concern the integrity warranties that must be implemented considering their importance for the reactors safety. Thus, stress is put on the exhaustiveness of the prevention of in-service degradation and on the safety scenarios considered. The second specificity concerns the possibility of activation of wear and corrosion products during their flow inside the reactor core. This second aspect leads to some constraints on the choice of the materials used and on the surface coating of the inside wall of big components of the primary circuit. The aim of this document is to develop the general approach adopted for the design of the pressure vessels of PWR fluid loops, and to stress more particularly on the nuclear particularities of these equipments. Some extensions of these rules to high temperature resistant materials (FBR-type reactors) are also evoked. Content: General considerations: design basis of pressure vessels, risk analysis and design conditions, ruining paths and safety coefficients; 2 - damage prevention for excessive deformation: definitions, criteria; 3 - prevention of the plastic instability damage: definition, criteria; 4 - buckling prevention: definition and mechanisms, rules and criteria; 5 - prevention of progressive deformation damage: definitions, plastic adaptation, plastic accommodation, progressive deformation; 6 - prevention of fatigue damage: definitions, general prevention approach, design fatigue curves, analytic approach, particular aspects, analysis of zones with geometrical singularity; 7 - prevention of sudden rupture damage: fragile rupture and ductile tear, general approach, analytic criteria, irradiation and aging effects; 8 - other potential damages; 9 - conclusion. (J.S.)

Design description of the vacuum vessel for the Advanced Toroidal Facility

International Nuclear Information System (INIS)

Chipley, K.K.; Nelson, B.E.; Vinyard, L.M.; Williamson, D.F.

1983-01-01

The Advanced Toroidal Facility (ATF) will be a stellarator experiment to investigate improvements in toroidal confinement. The vacuum vessel for this facility will provide the appropriate evacuated region for plasma containment within the helical field (HF) coils. The vessel is designed to provide the maximum reasonable volume inside the HF coils and to provide the maximum reasonable access for future diagnostics. The vacuum vessel design is at an early phase and all of the details have not been completed. The heat transfer analysis and stress analysis completed during the conceptual design indicate that the vessel will not change drastically

Mark III Containment vessel/annulus concrete design

International Nuclear Information System (INIS)

Chang, P.S.; Moussa, M.M.

1981-01-01

Recently, engineers have been considering the significant dynamic impact of safety/relief valve (S/RV) discharge loads on the containment structures, safety equipment, and piping systems in BWR type reactors. For a plant in the construction stage, extensive modifications will be made to qualify these new loads. The lower portion of the containment vessel serves as a suppression pool pressure boundary and is designed to sustain the effects of postulated loss of coolant accidents, seismic occurrences, S/RV discharge loads, and other effects. Extremely high spectral peak accelerations of the free-standing steel containment vessel can be obtained during the air dearing process of the S/RV discharge. Parametric studies indicated that a substantial reduction in response can be obtained by increasing the stiffness of the steel containment vessel in the lover area. A concrete backing configuration in the suppression pool area of Mark III Containment is proposed in this paper. A composite action is assumed between the steel containment vessel shell and the concrete section. The system is physically separated from the shield building. This approach warrants an early erection of the shield building and a late installation of piping systems in the containment vessel suppression pool area. Finite element analyses are performed by using ASHSD2 and EASE2 computer codes. The results of the analyses have shown the proposed stress criteria are satisfied. The approach pressented is justified to be a workable system for a new plant design. (orig./HP)

Structural considerations in design of lightweight glass-fiber composite pressure vessels

Science.gov (United States)

Faddoul, J. R.

1973-01-01

The design concepts used for metal-lined glass-fiber composite pressure vessels are described, comparing the structural characteristics of the composite designs with each other and with homogeneous metal pressure vessels. Specific design techniques and available design data are identified. The discussion centers around two distinctly different design concepts, which provide the basis for defining metal lined composite vessels as either (1) thin-metal lined, or (2) glass fiber reinforced (GFR). Both concepts are described and associated development problems are identified and discussed. Relevant fabrication and testing experience from a series of NASA-Lewis Research Center development efforts is presented.

ITER cryostat main chamber and vacuum vessel pressure suppression system design

International Nuclear Information System (INIS)

Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke; Nakashima, Yoshitane; Ueno, Osamu

1999-03-01

Design of Cryostat Main Chamber and Vacuum Vessel Pressure Suppression System (VVPS) of International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat is a cylindrical vessel that includes in-vessel component such as vacuum vessel, superconducting toroidal coils and poloidal coils. This cryostat provides the adiabatic vacuum about 10 -4 Pa for the superconducting coils operating at 4 K and forms the second confinement barrier to tritium. The adiabatic vacuum is to reduce thermal loads applied to the superconducting coils and their supports so as to keep their temperature 4 K. The VVPS consists of a suppression tank located under the lower bio-shield and 4 relief pipes to connect the vacuum vessel and the suppression tank. The VVPS is to keep the maximum pressure rise of the vacuum vessel below the design value of 0.5 MPa in case of the in-vessel LOCA (water spillage from in-vessel component). The spilled water and steam are lead to the suppression tank through the relief pipes when the internal pressure of vacuum vessel is over 0.2 MPa, and then the internal pressure is kept below 0.5 MPa. This report summarizes the structural design of the cryostat main chamber and pressure suppression system, together with their fabrication and installation. (author)

ITER cryostat main chamber and vacuum vessel pressure suppression system design

Energy Technology Data Exchange (ETDEWEB)

Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nakashima, Yoshitane; Ueno, Osamu

1999-03-01

Design of Cryostat Main Chamber and Vacuum Vessel Pressure Suppression System (VVPS) of International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat is a cylindrical vessel that includes in-vessel component such as vacuum vessel, superconducting toroidal coils and poloidal coils. This cryostat provides the adiabatic vacuum about 10{sup -4} Pa for the superconducting coils operating at 4 K and forms the second confinement barrier to tritium. The adiabatic vacuum is to reduce thermal loads applied to the superconducting coils and their supports so as to keep their temperature 4 K. The VVPS consists of a suppression tank located under the lower bio-shield and 4 relief pipes to connect the vacuum vessel and the suppression tank. The VVPS is to keep the maximum pressure rise of the vacuum vessel below the design value of 0.5 MPa in case of the in-vessel LOCA (water spillage from in-vessel component). The spilled water and steam are lead to the suppression tank through the relief pipes when the internal pressure of vacuum vessel is over 0.2 MPa, and then the internal pressure is kept below 0.5 MPa. This report summarizes the structural design of the cryostat main chamber and pressure suppression system, together with their fabrication and installation. (author)

Design concept for vessels and heat exchangers

International Nuclear Information System (INIS)

Elfmann, W.; Ferrari, L.D.B.

1981-01-01

A design concept for vessels and heat exchangers against internal and external loads resulting from normal operation and accident is shown. A definition and explanation of the operating conditions and stress levels are given. A description of the type of analysis (stress, fatigue, deformation, stability, earthquake and vibration) is presented in detail, also including technical guidelines which are used for the vessels and heat exchangers and their individual structure parts. (Author) [pt

Arctic research vessel design would expand science prospects

Science.gov (United States)

Elsner, Robert; Kristensen, Dirk

The U.S. polar marine science community has long declared the need for an arctic research vessel dedicated to advancing the study of northern ice-dominated seas. Planning for such a vessel began 2 decades ago, but competition for funding has prevented construction. A new design program is underway, and it shows promise of opening up exciting possibilities for new research initiatives in arctic marine science.With its latest design, the Arctic Research Vessel (ARV) has grown to a size and capability that will make it the first U.S. academic research vessel able to provide access to the Arctic Ocean. This ship would open a vast arena for new studies in the least known of the world's seas. These studies promise to rank high in national priority because of the importance of the Arctic Ocean as a source of data relating to global climate change. Other issues that demand attention in the Arctic include its contributions to the world's heat budget, the climate history buried in its sediments, pollution monitoring, and the influence of arctic conditions on marine renewable resources.

Prestressed concrete pressure vessels for nuclear reactors - 1973

Energy Technology Data Exchange (ETDEWEB)

1977-01-01

This standard deals with the design, construction, inspection and testing of prestressed concrete pressure vessels for nuclear reactors. Such pressure vessels serve the dual purpose of shielding and containing gas cooled nuclear reactors and are a form of civil engineering structure requiring particularly high integrity, and ensured leak tightness. (Metric)

Structural analysis and evaluation for the design of pressure vessel

International Nuclear Information System (INIS)

Arai, K.; Uragami, K.; Funada, T.; Baba, K.; Kira, T.

1977-01-01

For the design of pressure vessel, the detailed structural analysis such as the fatigue analysis under operating conditions is required by ASME Code or Japanese regulation. Accordingly, it should be verified by the analysis that the design of the pressure vessel is in compliance with the stress limitation defined in the Code or the regulation. However, it was apparent that the analysis is very complicated and takes a lot of time to evaluate in accordance with the Code requirements. Thereupon we developed the computer program by which we can perform the stress analysis with correctness and comparatively in a short period of design work reflecting the calculation results on detailed drawings to be used for fabrication. The computer program is controlled in combination with the system of the design work and out put list of the program can be directly used for the stress analysis report which is issued to customers. In addition to the above computer program, we developed the specific three dimensional finite element computer program to make sure of the structural integrity of the vessel head and flanges which are most complex for the analysis compared with the stress distribution measured by strain gauges on the vessel head and flange. Besides the structural analysis, the fracture mechanics analysis for the purpose of preventing the pressure vessel from the brittle fracture during heat-up and cool-down operation is also important and thereby we showed herein that the pressure vessel is in safety against the brittle fracture for the specified operating conditions. As a result of the above-mentioned analysis, the pressure vessel is designed with safety from the stand-points of the structural intensity and the fracture mechanics. (auth.)

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)

Starting procedure for internal combustion vessels

Science.gov (United States)

Harris, Harry A.

1978-09-26

A vertical vessel, having a low bed of broken material, having included combustible material, is initially ignited by a plurality of ignitors spaced over the surface of the bed, by adding fresh, broken material onto the bed to buildup the bed to its operating depth and then passing a combustible mixture of gas upwardly through the material, at a rate to prevent back-firing of the gas, while air and recycled gas is passed through the bed to thereby heat the material and commence the desired laterally uniform combustion in the bed. The procedure permits precise control of the air and gaseous fuel mixtures and material rates, and permits the use of the process equipment designed for continuous operation of the vessel.

Ultimate load design and testing of a cylindrical prestressed concrete vessel

International Nuclear Information System (INIS)

Stefanou, G.D.

1982-01-01

The object of this research was to design, construct and test to failure a prestressed concrete pressure vessel model that could be used to investigate the behavior of a full scale structure underworking and ultimate load. The properties and the design of the model was based generally on full scale vessels already constructed to house the nuclear reactors used in atomic power stations. To design the model the ultimate load approach was adopted throughout. All load factors associated with the prestressing have been defined and kept to a minimum in order that the vessel's behavior may be predicted. The tests on the vessel were carried out first on the elastic range to observe its behavior at working load and then at the ultimate range to observe the modes of failure and compare the actual results in both cases with the predicted values. Although full agreement between observed results and predicted values was not obtained, the conclusions drawn from the study were useful for the design of full scale vessels. (author)

The pressure vessel for the NSF tandem

International Nuclear Information System (INIS)

Jones, C.W.

1979-04-01

The pressure vessel is a major component of the 30 MV tandem Van de Graaff electrostatic accelerator to be used in nuclear structure research at Daresbury Laboratory. The accelerator will be capable of accelerating the full range of ions in the form of a beam. Acceleration takes place in a vertical evacuated tube (beam tube) by means of a high potential on a terminal at the central position, the terminal and beam tube assembly being supported by an insulated stack structure within the pressure vessel. Under operating conditions the vessel is filled with sulphur hexafluoride gas (SF 6 ) at high pressure which acts as an insulating medium between the centre terminal and the vessel wall. The vessel is situated inside a concrete tower which besides supporting the injector room above the vessel also acts as radiation shielding around the accelerator. The report covers: functional requirements; fundamental considerations with regard to the design and procurement; detail design; materials; manufacture; acceptance test; surface treatment; final leak test. (U.K.)

Prestressed reactor vessel for nuclear power plants

International Nuclear Information System (INIS)

Schoening, J.; Schwiers, H.G.

1982-01-01

With usual pressure vessels for nuclear reactor plants, especially for gas-cooled nuclear reactors, the load occurring due to the inner overpressure, especially the tensile load affecting the vessel top and/or bottom, their axis of inertia being horizontal, shall be compensated without a supplementary modification in design of the top and/or the bottom. This is attained by choosing an appropriate prestressing system of the vessel wall in the field the top and/or the bottom, so that the top and/or the bottom form a tension vault directed towards the interior of the vessel. (orig.) [de

Preliminary design analysis of hot gas ducts and a intermediate heat exchanger for the nuclear hydrogen reactor

International Nuclear Information System (INIS)

Song, K. N.; Kim, Y. W.

2008-01-01

Korea Atomic Energy Research Institute (KAERI) is in the process of carrying out a nuclear hydrogen system by considering the indirect cycle gas cooled reactors that produce heat at temperatures in the order of 950 .deg. C. Primary and secondary hot gas ducts with coaxial double tubes and are key components connecting a reactor pressure vessel and a intermediate heat exchanger for the nuclear hydrogen system. In this study, preliminary design analyses on the hot gas ducts and the intermediate heat exchanger were carried out. These preliminary design activities include a preliminary design on the geometric dimensions, a preliminary strength evaluation, thermal sizing, and an appropriate material selection

Optimization of Helium Vessel Design for ILC Cavities

Energy Technology Data Exchange (ETDEWEB)

Fratangelo, Enrico [Univ. of Pisa (Italy)

2009-01-01

certify the compliance of the Helium vessel and the cavity to the ASME code standard. After briefly recalling to the main contents of the the ASME Code (Sections II and Vlll - Division ll), the procedure used for finding all relevant stresses and comparing the obtained results with the maximum values allowed are explained. This part also includes the buckling verification of the cavity. In Chapter 5 the manufacturing process of the cavity end-caps, whose function is to link the Helium vessel with the cavity, is studied. The present configuration of the dies is described and the manufacturing process is simulated in order to explain the origin of some defects fol.llld on real parts. Finally a new design of the dies is proposed and the resulting deformed piece is compared with the design requirements. Chapter 6 describes a finite elements analysis to assess the efficiency and the stiffness of the Helium vessel. Furthermore the results of the optimization of the Helium vessel (in order to increase the value of the efficiency) are reported. The same stiffness analysis is used in Chapter 7 for the Blade-Tuner study. After a description of this tuner and of its function, the preliminary analyses done to confirm the results provided by the vendor are described and then its limiting load conditions are found. Chapter 8 shows a study of the resistance of all the welds present in between the cavity and the end-cap and between the end-caps and the He vessel for a smaller superconducting cavity operating at 3.9 GHz. Finally Chapter 9 briefly describes some R&D activities in progress at INFN (Section of Pisa) and Fermilab that could produce significant cost reductions of the Helium vessel design. All the finite elements analyses contained and described in this thesis made possible the certification of the whole superconducting cavity-Helium vessel assembly at Fermilab. Furthermore they gave several useful indications to the Fermilab staff to improve the performance of the Helium

A prototype knowledge based system for pressure vessel design

Energy Technology Data Exchange (ETDEWEB)

Gunnarsson, L.

1991-11-22

The usage of expert system techniques in the area of mechanical engineering design has been studied. A prototype expert system for pressure vessel design has been developed. The work has been carried out in two steps. Firstly, a pre-processor for the finite element system PCFEMP, named INFEMP, was developed. Secondly, an expert supported system for pressure vessel design, named PVES, was developed. Both INFEMP and PVES are integrated to the AutoCAD system, and AutoCAD`s language AutoLISP has been used. A practical example has been investigated to demonstrate the principal ideas of the prototype. (au).

A prototype knowledge based system for pressure vessel design

Energy Technology Data Exchange (ETDEWEB)

Gunnarsson, L.

1991-11-22

The usage of expert system techniques in the area of mechanical engineering design has been studied. A prototype expert system for pressure vessel design has been developed. The work has been carried out in two steps. Firstly, a pre-processor for the finite element system PCFEMP, named INFEMP, was developed. Secondly, an expert supported system for pressure vessel design, named PVES, was developed. Both INFEMP and PVES are integrated to the AutoCAD system, and AutoCAD's language AutoLISP has been used. A practical example has been investigated to demonstrate the principal ideas of the prototype. (au).

A prototype knowledge based system for pressure vessel design

International Nuclear Information System (INIS)

Gunnarsson, L.

1991-01-01

The usage of expert system techniques in the area of mechanical engineering design has been studied. A prototype expert system for pressure vessel design has been developed. The work has been carried out in two steps. Firstly, a pre-processor for the finite element system PCFEMP, named INFEMP, was developed. Secondly, an expert supported system for pressure vessel design, named PVES, was developed. Both INFEMP and PVES are integrated to the AutoCAD system, and AutoCAD's language AutoLISP has been used. A practical example has been investigated to demonstrate the principal ideas of the prototype. (au)

Structural design and manufacturing of TPE-RX vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Sago, H.; Orita, J.; Kaguchi, H.; Ishigami, Y. [Mitsubishi Heavy Ind. Ltd., Kobe (Japan); Urata, K.; Kudough, F. [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan); Hasegawa, M.; Oyabu, I. [Mitsubishi Electric Co., Tokyo (Japan); Yagi, Y.; Sekine, S.; Shimada, T.; Hirano, Y.; Sakakita, H.; Koguchi, H. [Electrotechnical Laboratory, Tsukuba (Japan)

1999-10-01

TPE-RX is a newly constructed, large-sized reversed field pinch (RFP) machine installed at the Electrotechnical Laboratory of the Ministry of International Trade and Industry in Japan. This is the third largest RFP in the world. Major and minor radii of the plasma are 1.72 and 0.45 m, respectively. TPE-RX aims to optimize plasma confinement up to 1 MA. RFP plasma configuration was successfully obtained in March 1998. This paper reports the structural design and manufacturing of the vacuum vessel of TPE-RX. The supporting system on the bellows sections of the vessel was designed based on a detailed finite element method. The integrity of the vacuum vessel against a plasma disruption has been confirmed using dynamic inelastic analyses. (orig.)

Structural design and manufacturing of TPE-RX vacuum vessel

International Nuclear Information System (INIS)

Sago, H.; Orita, J.; Kaguchi, H.; Ishigami, Y.; Urata, K.; Kudough, F.; Hasegawa, M.; Oyabu, I.; Yagi, Y.; Sekine, S.; Shimada, T.; Hirano, Y.; Sakakita, H.; Koguchi, H.

1999-01-01

TPE-RX is a newly constructed, large-sized reversed field pinch (RFP) machine installed at the Electrotechnical Laboratory of the Ministry of International Trade and Industry in Japan. This is the third largest RFP in the world. Major and minor radii of the plasma are 1.72 and 0.45 m, respectively. TPE-RX aims to optimize plasma confinement up to 1 MA. RFP plasma configuration was successfully obtained in March 1998. This paper reports the structural design and manufacturing of the vacuum vessel of TPE-RX. The supporting system on the bellows sections of the vessel was designed based on a detailed finite element method. The integrity of the vacuum vessel against a plasma disruption has been confirmed using dynamic inelastic analyses. (orig.)

Ex-vessel core catcher design requirements and preliminary concepts evaluation

International Nuclear Information System (INIS)

Friedland, A.J.; Tilbrook, R.W.

1974-01-01

As part of the overall study of the consequences of a hypothetical failure to scram following loss of pumping power, design requirements and preliminary concepts evaluation of an ex-vessel core catcher (EVCC) were performed. EVCC is the term applied to a class of devices whose primary objective is to provide a stable subcritical and coolable configuration within containment following a postulated accident in which it is assumed that core debris has penetrated the Reactor Vessel and Guard Vessel. Under these assumed conditions a set of functional requirements were developed for an EVCC and several concepts were evaluated. The studies were specifically directed toward the FFTF design considering the restraints imposed by the physical design and construction of the FFTF plant

Experimental and Numerical Study of Effect of Thermal Management on Storage Capacity of the Adsorbed Natural Gas Vessel

KAUST Repository

Ybyraiymkul, Doskhan

2017-07-08

One of the main challenges in the adsorbed natural gas (ANG) storage system is the thermal effect of adsorption, which significantly lowers storage capacity. These challenges can be solved by efficient thermal management system. In this paper, influence of thermal management on storage capacity of the ANG vessel was studied experimentally and numerically. 3D numerical model was considered in order to understand heat transfer phenomena and analyze influence of thermal control comprehensively. In addition, a detailed 2D axisymmetric unit cell model of adsorbent layer with heat exchanger was developed, followed by optimization of heat exchanging device design to minimize volume occupied by fins and tubes. Heat transfer, mass transfer and adsorption kinetics, which occur in ANG vessel during charging process, are accounted for in models. Nelder-Mead method is implemented to obtain the geometrical parameters, which lead to the optimal characteristics of heat exchange. A new optimized configuration of ANG vessel was developed with compact heat exchanger. Results show that storage capacity of the ANG vessel increased significantly due to lowering of heat exchanger volume for 3 times from 13.5% to 4.3% and effective temperature control.

Conceptual design of EAST flexible in-vessel inspection system

International Nuclear Information System (INIS)

Peng, X.B.; Song, Y.T.; Li, C.C.; Lei, M.Z.; Li, G.

2010-01-01

Remote handling technology, especially the flexible in-vessel inspection system (FIVIS) without breaking the working condition of the vacuum vessel, has been identified as one major challenge on the maintenance for the future tokamak fusion reactor. The FIVIS introduced here is specially developed for EAST superconducting tokamak that has actively cooled plasma facing components (PFCs). It aims flexible close-up inspection of EAST PFCs to help the understanding of operation issues that could occur in the vacuum vessel. This paper resumes the preliminary work of the FIVIS project, including the requirement analysis and the development of the conceptual design. The FIVIS consists out of a long reach multi-articulated manipulator and a process tool. The manipulator has a modular design for its subsystems and can reach all areas of the first wall in the distance of 15 mm and in the range of ±90 o along toroidal direction. It will be folded and hidden in the designated horizontal port during plasma discharge period.

Preliminary structural evaluations of the STAR-LM reactor vessel and the support design

International Nuclear Information System (INIS)

Koo, Gyeong-Hoi; Sienicki, James J.; Moisseytsev, Anton

2007-01-01

In this paper, preliminary structural evaluations of the reactor vessel and support design of the STAR-LM (The Secure, Transportable, Autonomous Reactor - Liquid Metal variant), which is a lead-cooled reactor, are carried out with respect to an elevated temperature design and seismic design. For an elevated temperature design, the structural integrity of a direct coolant contact to the reactor vessel is investigated by using a detail structural analysis and the ASME-NH code rules. From the results of the structural analyses and the integrity evaluations, it was found that the design concept of a direct coolant contact to the reactor vessel cannot satisfy the ASME-NH rules for a given design condition. Therefore, a design modification with regards to the thermal barrier is introduced in the STAR-LM design. For a seismic design, detailed seismic time history response analyses for a reactor vessel with a consideration of a fluid-structure interaction are carried out for both a top support type and a bottom support type. And from the results of the hydrodynamic pressure responses, an investigation of the minimum thickness design of the reactor vessel is tentatively carried out by using the ASME design rules

Iter in vessel viewing system design and assessment activities

Energy Technology Data Exchange (ETDEWEB)

Neri, C., E-mail: carlo.neri@enea.it [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Costa, P.; Ferri De Collibus, M.; Florean, M.; Mugnaini, G.; Pillon, M.; Pollastrone, F.; Rossi, P. [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy)

2011-10-15

The In Vessel Viewing System (IVVS) is fundamental remote handling equipment, which will be used to make a survey of the status of the blanket first wall and divertor plasma facing components. A prototype of a laser In Vessel Viewing and ranging System was developed and tested at ENEA laboratories in Frascati under EFDA task agreements, it is able to perform sub-millimetric bi-dimensional and three-dimensional images inside ITER during maintenance procedure allowing the evaluation of the state and damages of the in-vessel surface. The present prototype has been designed to operate under room conditions and starting from springtime 2009 a Grant with F4E is in progress for the design and the assessment of the IVVS system for ITER, keeping in account all the environmental conditions and constraints.

French administrative practice and design codes for nuclear vessels

International Nuclear Information System (INIS)

Roche, R.L.

1987-07-01

French regulations on boilers and pressure vessels have prevailed for a very long time, the first measure having been promulgated on 29 October 1823. Restraining the attention to nuclear pressure vessels it must be pointed out regulations and enforcement by public authorities are more stringent than they are for conventional pressure vessels. The first part of this paper will be devoted to regulations with a special attention to the decree of 26 February 1974 and to the practice of public authorities in this field with special attention given to the Bureau de Controle de la Construction Nucleaire (BCCN = Bureau of Inspection of Nuclear Design and Manufacturing). The second part of this paper will deal with the French construction codes for nuclear components RCC-M (water reactors) and RCC-MR (elevated temperature design)

Pressure vessel for nuclear reactor plant consisting of several pre-stressed cast pressure vessels

International Nuclear Information System (INIS)

Bodmann, E.

1984-01-01

Several cylindrical pressure vessel components made of pressure castings are arranged on a sector of a circle around the cylindrical cast pressure vessel for accommodating the helium cooled HTR. Each component pressure vessel is connected to the reactor vessel by a horizontal gas duct. The contact surfaces between reactor and component pressure vessel are in one plane. In the spaces between the individual component pressure vessels, there are supporting blocks made of cast iron, which are hollow and also have flat surfaces. With the reactor vessel and the component pressure vessels they form a disc-shaped connecting part below and above the gas ducts. (orig./PW)

Analysis of stress in reactor core vessel under effect of pressure lose shock wave

International Nuclear Information System (INIS)

Li Yong; Liu Baoting

2001-01-01

High Temperature gas cooled Reactor (HTR-10) is a modular High Temperature gas cooled Reactor of the new generation. In order to analyze the safety characteristics of its core vessel in case of large rupture accident, the transient performance of its core vessel under the effect of pressure lose shock wave is studied, and the transient pressure difference between the two sides of the core vessel and the transient stresses in the core vessel is presented in this paper, these results can be used in the safety analysis and safety design of the core vessel of HTR-10. (author)

Design and fabrication of the vacuum vessel for the Advanced Toroidal Facility

International Nuclear Information System (INIS)

Chipley, K.K.; Frey, G.N.

1985-01-01

The vacuum vessel for the Advanced Toroidal Facility (ATF) is a heavily contoured and very complex formed vessel that is specifically designed to allow for maximum plasma volume in a pure stellarator arrangement. The design of the facility incorporates an internal vessel that is closely fitted to the two helical field coils following the winding law theta = 1/6phi. Metallic seals have been incorporated throughout the system to minimize impurities. The vessel has been fabricated utilizing a comprehensive set of tooling fixtures specifically designed for the task of forming 6-mm stainless steel plate to the complex shape. Computer programs were used to develop a series of ribs that essentially form an internal mold of the vessel. Plates were press-formed with multiple compound curves, fitted to the fixture, and joined with full-penetration welds. 7 refs., 8 figs

FINAL DESIGN REVIEW REPORT Subcritical Experiments Gen 2, 3-ft Confinement Vessel Weldment

Energy Technology Data Exchange (ETDEWEB)

Romero, Christopher [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-09-28

A Final Design Review (FDR) of the Subcritical Experiments (SCE) Gen 2, 3-ft. Confinement Vessel Weldment was held at Los Alamos National Laboratory (LANL) on September 14, 2017. The review was a focused review on changes only to the confinement vessel weldment (versus a system design review). The changes resulted from lessons-learned in fabricating and inspecting the current set of confinement vessels used for the SCE Program. The baseline 3-ft. confinement vessel weldment design has successfully been used (to date) for three (3) high explosive (HE) over-tests, two (2) fragment tests, and five (5) integral HE experiments. The design team applied lessons learned from fabrication and inspection of these vessel weldments to enhance fit-up, weldability, inspection, and fitness for service evaluations. The review team consisted of five (5) independent subject matter experts with engineering design, analysis, testing, fabrication, and inspection experience. The

Heat exchanger design considerations for high temperature gas-cooled reactor (HTGR) plants

International Nuclear Information System (INIS)

McDonald, C.F.; Vrable, D.L.; Van Hagan, T.H.; King, J.H.; Spring, A.H.

1980-02-01

Various aspects of the high-temperature heat exchanger conceptual designs for the gas turbine (HTGR-GT) and process heat (HTGR-PH) plants are discussed. Topics include technology background, heat exchanger types, surface geometry, thermal sizing, performance, material selection, mechanical design, fabrication, and the systems-related impact of installation and integration of the units in the prestressed concrete reactor vessel. The impact of future technology developments, such as the utilization of nonmetallic materials and advanced heat exchanger surface geometries and methods of construction, is also discussed

Limit analysis and design of containment vessels

International Nuclear Information System (INIS)

Save, M.

1984-01-01

In the introduction, the theory of plastic analysis of shells is briefly recalled. Minimum-volume design for assigned load factor at plastic collapse is then considered and optimality criteria are derived for plates and shells of continuously varying or piecewise-constant thickness. In the first part, containers made of metal are examined. Analytical and numerical limit analysis solutions and corresponding experimental results are considered for various types of vessels, including intersecting shells. Attention is given to experimental post-yield behavior. Some tests up to fracture are discussed. New theoretical and experimental results of limit analysis of stiffened cylindrical vessels are presented, in which reinforcing rings are treated as discrete structural element (no smearing out) and due account is taken of their strong curvature. Cases of collapse by instability under internal pressure are pointed out. Minimum-volume design of circular plates and cylindrical shells is then formulated and various examples are presented of sandwich and solid metal structures. Containers of piecewise-constant thickness are given particular attention. Available experimental evidence on minimum-volume design of plates and shells is reviewed and commented upon. The second part deals with reinforced concrete vessels. Cylindrical containers are studied, from both points of view of limit analysis and of limit design with minimum volume of reinforcement. The practical use of the latter solutions is discussed. A third part reviews other loading cases (including cyclic and impact loads) and gives indications on corresponding theories, formulations and solution methods. The last part is devoted to a discussion of the limitations of the methods presented, within the frame of the 'limit states' design philosophy, which is first briefly recalled. Considerations on further research in the field conclude the paper. (orig.)

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.

Conceptual design of the handling and storage system for spent target vessel

Energy Technology Data Exchange (ETDEWEB)

Adachi, Junichi; Sasaki, Shinobu; Kaminaga, Masanori; Hino, Ryutaro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-03-01

A conceptual design of a handling and storage system for spent target vessels has been carried out, in order to establish spent target technology for the neutron scattering facility. The spent target vessels must be treated remotely with high reliability and safety, since they are highly activated and contain the poisonous mercury. The system is composed of a target exchange trolley to exchange the target vessel, remote handling equipment such as manipulators, airtight casks for the spent target vessel, storage pits and so on. This report presents the results of conceptual design study on a basic plan, a handling procedure, main devices and their arrangement of a handling and storage system for the spent target vessels. (author)

Proof testing of an explosion containment vessel

Energy Technology Data Exchange (ETDEWEB)

Esparza, E.D. [Esparza (Edward D.), San Antonio, TX (United States); Stacy, H.; Wackerle, J. [Los Alamos National Lab., NM (United States)

1996-10-01

A steel containment vessel was fabricated and proof tested for use by the Los Alamos National Laboratory at their M-9 facility. The HY-100 steel vessel was designed to provide total containment for high explosives tests up to 22 lb (10 kg) of TNT equivalent. The vessel was fabricated from an 11.5-ft diameter cylindrical shell, 1.5 in thick, and 2:1 elliptical ends, 2 in thick. Prior to delivery and acceptance, three types of tests were required for proof testing the vessel: a hydrostatic pressure test, air leak tests, and two full design charge explosion tests. The hydrostatic pressure test provided an initial static check on the capacity of the vessel and functioning of the strain instrumentation. The pneumatic air leak tests were performed before, in between, and after the explosion tests. After three smaller preliminary charge tests, the full design charge weight explosion tests demonstrated that no yielding occurred in the vessel at its rated capacity. The blast pressures generated by the explosions and the dynamic response of the vessel were measured and recorded with 33 strain channels, 4 blast pressure channels, 2 gas pressure channels, and 3 displacement channels. This paper presents an overview of the test program, a short summary of the methodology used to predict the design blast loads, a brief description of the transducer locations and measurement systems, some of the hydrostatic test strain and stress results, examples of the explosion pressure and dynamic strain data, and some comparisons of the measured data with the design loads and stresses on the vessel.

Design of Hemispherical Downward-Facing Vessel for Critical Heat Flux Experiment

International Nuclear Information System (INIS)

Hwang, J. S.; Suh, K. Y.

2009-01-01

The in-vessel retention (IVR) is one of major severe accident management strategies adopted by some operating nuclear power plants during a severe accident. The recent Shin-Gori Units 3 and 4 of the Advanced Power Reactor 1400 MWe (APR1400) have adopted the external reactor vessel cooling (ERVC) by reactor cavity flooding as major severe accident management strategy. The ERVC in the APR1400 design resorts to active flooding system using thermal insulator. The Corium Attack Stopper Apparatus Spherical Channel (CASA SC) tests are conducted to measure the critical power and critical heat flux (CHF) on a downward hemispherical vessel scaled down from the APR1400 lower head by 1/10 on a linear scale. CASA is designed through scaling and thermal analysis to simulate the APR1400 vessel and thermal insulator. The heated vessel of CASA SC represents the external surface of a hemisphere submerged vessel in water. The heated vessel plays an important role in the ERVC experiment depending on the configuration of oxide pool and metallic layer. Hand calculation and computational analysis are performed to produce high heat flux from the downward facing hemisphere in excess of 1 MW/m 2

Simulant Basis for the Standard High Solids Vessel Design

Energy Technology Data Exchange (ETDEWEB)

Peterson, Reid A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fiskum, Sandra K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Suffield, Sarah R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Daniel, Richard C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gauglitz, Phillip A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wells, Beric E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2017-09-30

The Waste Treatment and Immobilization Plant (WTP) is working to develop a Standard High Solids Vessel Design (SHSVD) process vessel. To support testing of this new design, WTP engineering staff requested that a Newtonian simulant and a non-Newtonian simulant be developed that would represent the Most Adverse Design Conditions (in development) with respect to mixing performance as specified by WTP. The majority of the simulant requirements are specified in 24590-PTF-RPT-PE-16-001, Rev. 0. The first step in this process is to develop the basis for these simulants. This document describes the basis for the properties of these two simulant types. The simulant recipes that meet this basis will be provided in a subsequent document.

Problems in Pressure Vessel Design and Manufacture

Energy Technology Data Exchange (ETDEWEB)

Hellstroem, O [Uddeholms AB, Degerfors (Sweden); Nilson, Ragnar [AB Atomenergi, Nykoeping (Sweden)

1963-05-15

The general desire by the power reactor process makers to increase power rating and their efforts to involve more advanced thermal behaviour and fuel handling facilities within the reactor vessels are accompanied by an increase in both pressure vessel dimensions and various difficulties in giving practical solutions of design materials and fabrication problems. In any section of this report it is emphasized that difficulties and problems already met with will meet again in the future vessels but then in modified forms and in many cases more pertinent than before. As for the increase in geometrical size it can be postulated that with use of better materials and adjusted fabrication methods the size problems can be taken proper care of. It seems likely that vessels of sufficient large diameter and height for the largest power output, which is judged as interesting in the next ten year period, can be built without developing totally new site fabrication technique. It is, however, supposed that such a fabrication technique will be feasible though at higher specific costs for the same quality requirements as obtained in shop fabrication. By the postulated use of more efficient vessel material with principally the same good features of easy fabrication in different stages such as preparation, welding, heat treatment etc as ordinary or slightly modified carbon steels the increase in wall thickness might be kept low. There exists, however, a development work to be done for low-alloy steels to prove their justified use in large reactor pressure vessels.

Problems in Pressure Vessel Design and Manufacture

International Nuclear Information System (INIS)

Hellstroem, O.; Nilson, Ragnar

1963-05-01

The general desire by the power reactor process makers to increase power rating and their efforts to involve more advanced thermal behaviour and fuel handling facilities within the reactor vessels are accompanied by an increase in both pressure vessel dimensions and various difficulties in giving practical solutions of design materials and fabrication problems. In any section of this report it is emphasized that difficulties and problems already met with will meet again in the future vessels but then in modified forms and in many cases more pertinent than before. As for the increase in geometrical size it can be postulated that with use of better materials and adjusted fabrication methods the size problems can be taken proper care of. It seems likely that vessels of sufficient large diameter and height for the largest power output, which is judged as interesting in the next ten year period, can be built without developing totally new site fabrication technique. It is, however, supposed that such a fabrication technique will be feasible though at higher specific costs for the same quality requirements as obtained in shop fabrication. By the postulated use of more efficient vessel material with principally the same good features of easy fabrication in different stages such as preparation, welding, heat treatment etc as ordinary or slightly modified carbon steels the increase in wall thickness might be kept low. There exists, however, a development work to be done for low-alloy steels to prove their justified use in large reactor pressure vessels

Preliminary study of an expert system for mechanical design of a pressure vessel

International Nuclear Information System (INIS)

Kasmuri, N.H.; Md Som, A.

2006-01-01

This paper describes a preliminary study of an expert system for mechanical design of a pressure vessel. The system supports the framework for the conceptual mechanical design from the initial stages within the design procedures. ASME Boiler and Pressure Vessel Code Section VIII Division 1 were applied as a design rule. The proposed methodology facilitates the development of knowledge base acquisition, knowledge base construction and the prototype implementation. This study characterizes a knowledge base (procedure) of mechanical design of a pressure vessel subjected to internal pressure including all design parameters; i.e. temperature, shell thickness, selection of materials of constructions, stress analysis procedure, support and ancillary items. The rationalization of the mechanical design is shown in the form of a schematic flow diagram. A Kappa PC expert system shell is used as a tool to develop the prototype software. It provides graphical representation for creating objects, hierarchies and rules for knowledge base used in pressure vessel design. (Author)

Gas supply device

International Nuclear Information System (INIS)

Yamamoto, Tokihiro.

1986-01-01

Purpose: When injecting gas during plasma discharge, to prevent erronous operation of the piezoelectric valves due to vibrations and to inject gas into the vacuum vessel by accurately controlling gas flow. Constitution: The piezoelectric valve designed to control the flow of gas when charging vacuum vessels is installed between the vacuum vessel and the gas cylinder and a pulse voltage generator to control the valve (open/close) and the reverse voltage generator to apply a reverse current to the valve element are also provided. When a voltage is applied to the piezoelectric element, the piezoelectric element bends. However, when the polarity is changed, the direction of the bend is also reversed. The seat portion is pressed in proportion to the degree of bending. When gas is injected while plasma is being discharged, the gap created, due to vibrations, between the piezoelectric element and the sealing material is prevented by the application of a reverse current. This allows precise control of the gas injection volume as well as the injection interval. This assures maintenance of precise test conditions when testing thermonuclear devices. (Horiuchi, T.)

Design system for in-vessel mainipulator of fusion reactor 'DESIM'

International Nuclear Information System (INIS)

Adachi, Junihci; Kobayashi, Takeshi; Ise, Hideo; Sato, Keisuke; Matsuda, Hirotsugu

1989-01-01

A computer aided design system 'DESIM' for the in-vessel manipulators of nuclear fusion reactors has been developed to design the manipulators efficiently. The DESIM consists of the following subsystems: (1) the design system for arm mechanisms to realize optimum manipulation performance in the specified workspace; (2) the robot simulator to study manipulator movement, postures and interference problems; (3) the CAD system which is used to define the structure object data for robots, and the interface system for the data conversion from the CAD system to the robot simulator. The DESIM has been used to design the in-vessel manipulator for the Fusion Experimental Reactor (FER) to confirm the effectiveness. (author)

[Conjunct changes in the resistance and engorgement of the cerebral vessels in shifts in the blood gas composition].

Science.gov (United States)

Krasil'nikov, V G; Artem'eva, A I

1982-08-01

In anesthetized cats, under perfusion and with constant volume of the hemodynamically isolated brain, hypercapnia and hypoxia led to a decrease of cerebral vessels resistance and to a reduction of the brain blood flow, whereas a decrease in the PCO2 and an increase in the PO2 in the blood exerted on opposite effect. The different responses of the vessels had some similar features in respect to threshold changes of the PCO2 and PO2, to potentiation of effects of both parts of the brain vascular system on increased shifts of the blood gas tension, to greater sensitivity of both parts to PCO2 changes, to effect of the blood gas tension on reactivity of both parts to noradrenaline. The authors suggest a possibility of alterations of the filter-absorption interrelationships in the brain due to different responses of arterial and venous vessels to changes of the blood gas tension.

The characteristics of the prestressed concrete reactor vessel of the HHT demonstration plant

International Nuclear Information System (INIS)

Schoening, J.; Schwiers, H.G.

1979-01-01

The paper concentrates on the design studies of the HTGR prestressed concrete reactor vessel (PCRV) for the HHT Demonstration Plant. The multi-cavity reactor pressure vessel accommodates all components carrying primary gas, including heat exchangers and gas turbine. For reasons of economics and availability of the reactor plant, generic requirements are made for the PCRV. A short description of the power plant is also presented

Design features of the KSTAR in-vessel control coils

Energy Technology Data Exchange (ETDEWEB)

Kim, H.K. [National Fusion Research Institute (NFRI), 52 Yeoeun-dong, Yusung-ku, Daejeon, 305-333 (Korea, Republic of)], E-mail: hkkim@nfri.re.kr; Yang, H.L.; Kim, G.H.; Kim, Jin-Yong; Jhang, Hogun; Bak, J.S.; Lee, G.S. [National Fusion Research Institute (NFRI), 52 Yeoeun-dong, Yusung-ku, Daejeon, 305-333 (Korea, Republic of)

2009-06-15

In-vessel control coils (IVCCs) are to be used for the fast plasma position control, field error correction (FEC), and resistive wall mode (RWM) stabilization for the Korea Superconducting Tokamak Advanced Research (KSTAR) device. The IVCC system comprises 16 segments to be unified into a single set to achieve following remarkable engineering advantages; (1) enhancement of the coil system reliability with no welding or brazing works inside the vacuum vessel, (2) simplification in fabrication and installation owing to coils being fabricated outside the vacuum vessel and installed after device assembly, and (3) easy repair and maintenance of the coil system. Each segment is designed in 8 turns coil of 32 mm x 15 mm rectangular oxygen free high conductive copper with a 7 mm diameter internal coolant hole. The conductors are enclosed in 2 mm thick Inconel 625 rectangular welded vacuum jacket with epoxy/glass insulation. Structural analyses were implemented to evaluate structural safety against electromagnetic loads acting on the IVCC for the various operation scenarios using finite element analysis. This paper describes the design features and structural analysis results of the KSTAR in-vessel control coils.

In-Vessel Coil Material Failure Rate Estimates for ITER Design Use

Energy Technology Data Exchange (ETDEWEB)

L. C. Cadwallader

2013-01-01

The ITER international project design teams are working to produce an engineering design for construction of this large tokamak fusion experiment. One of the design issues is ensuring proper control of the fusion plasma. In-vessel magnet coils may be needed for plasma control, especially the control of edge localized modes (ELMs) and plasma vertical stabilization (VS). These coils will be lifetime components that reside inside the ITER vacuum vessel behind the blanket modules. As such, their reliability is an important design issue since access will be time consuming if any type of repair were necessary. The following chapters give the research results and estimates of failure rates for the coil conductor and jacket materials to be used for the in-vessel coils. Copper and CuCrZr conductors, and stainless steel and Inconel jackets are examined.

Advanced dependent pressure vessel (DPV) nickel-hydrogen spacecraft battery design

Energy Technology Data Exchange (ETDEWEB)

Coates, D.K.; Grindstaff, B.; Swaim, O.; Fox, C. [Eagle-Picher Industries, Inc., Joplin, MO (United States). Advanced Systems Operation

1995-12-31

The dependent pressure vessel (DPV) nickel-hydrogen (NiH{sub 2}) battery is being developed as a potential spacecraft battery design for both military and commercial satellites. The limitations of standard NiH{sub 2} individual pressure vessel (IPV) flight battery technology are primarily related to the internal cell design and the battery packaging issues associated with grouping multiple cylindrical cells. The DPV cell design offers higher energy density and reduced cost, while retaining the established IPV technology flight heritage and database. The advanced cell design offers a more efficient mechanical, electrical and thermal cell configuration and a reduced parts count. The geometry of the DPV cell promotes compact, minimum volume packaging and weight efficiency. The DPV battery design offers significant cost and weight savings advantages while providing minimal design risks.

ITER in-vessel system design and performance

Science.gov (United States)

Parker, R. R.

2000-03-01

The article reviews the design and performance of the in-vessel components of ITER as developed for the Engineering Design Activities (EDA) Final Design Report. The double walled vacuum vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g. the most intense vertical displacement events (VDEs) and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature non-uniformities. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor concept is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m2 are expected on the target. These are accommodated by HHF technology developed during the EDA. Disruptions and VDEs can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowable ranges for all postulated disruption and seismic events.

ITER in-vessel system design and performance

International Nuclear Information System (INIS)

Parker, R.R.

2000-01-01

The article reviews the design and performance of the in-vessel components of ITER as developed for the Engineering Design Activities (EDA) Final Design Report. The double walled vacuum vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g. the most intense vertical displacement events (VDEs) and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature non-uniformities. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor concept is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m 2 are expected on the target. These are accommodated by HHF technology developed during the EDA. Disruptions and VDEs can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowable ranges for all postulated disruption and seismic events. (author)

Design and deployment of autoclave pressure vessels for the portable deep-sea drill rig MeBo (Meeresboden-Bohrgerät)

Science.gov (United States)

Pape, Thomas; Hohnberg, Hans-Jürgen; Wunsch, David; Anders, Erik; Freudenthal, Tim; Huhn, Katrin; Bohrmann, Gerhard

2017-11-01

Pressure barrels for sampling and preservation of submarine sediments under in situ pressure with the robotic sea-floor drill rig MeBo (Meeresboden-Bohrgerät) housed at the MARUM (Bremen, Germany) were developed. Deployments of the so-called MDP (MeBo pressure vessel) during two offshore expeditions off New Zealand and off Spitsbergen, Norway, resulted in the recovery of sediment cores with pressure stages equaling in situ hydrostatic pressure. While initially designed for the quantification of gas and gas-hydrate contents in submarine sediments, the MDP also allows for analysis of the sediments under in situ pressure with methods typically applied by researchers from other scientific fields (geotechnics, sedimentology, microbiology, etc.). Here we report on the design and operational procedure of the MDP and demonstrate full functionality by presenting the first results from pressure-core degassing and molecular gas analysis.

Design, fabrication and quality assurance of pressure vessels

International Nuclear Information System (INIS)

Kimura, Ichiro; Miki, Masao; Yamazaki, Tsuneji; Tanaka, Yoshikazu; Sato, Misao

1978-01-01

The production facilities, design and manufacturing technologies, and quality assurance in the Toyo Works, Ehime Manufactory, Sumitomo Heavy Industries, Ltd., which manufactures pressure vessels, are described, and especially the actual example of non-destructive tests is shown. The Toyo Works was completed in April, 1973, to manufacture large structures such as pressure vessels, offshore structures and bridges. The total area of the site is 535,000 m 2 , that of factory buildings is 33,600 m 2 , and the outdoor assembling yard is 114,800 m 2 . The large dry dock and main installations such as 12,000 tf hydraulic press, an annealing furnace, a heat treating furnace, a quenching tank, a horizontal boring machine, 6 m vertical lathe, various welding machines, 8 MeV X-ray apparatus, sand blasting and pickling facilities, and two 160 t cranes for shipment are arranged so as to enable smooth flow of production. The standards for chemical pressure vessels in various countries are compared, and considerably high allowable stress is adopted in Europe. The design and stress analysis of pressure vessels are carried out in accordance with ASME Section 8, Div. 1 or Div. 2. As for the materials, attention must be paid to the change of properties due to heat and strain, temper brittleness, low temperature toughness and so on. The quality assurance system must be established to observe the requirements of standards. (Kako, I.)

Finite element analyses for design evaluation of biodegradable magnesium alloy stents in arterial vessels

Energy Technology Data Exchange (ETDEWEB)

Wu Wei [Laboratory of Biological Structure Mechanics, Structural Engineering Department, Politecnico di Milano, Piazza Leonardo da Vinci, 32, 20133 Milan (Italy); Gastaldi, Dario, E-mail: dario.gastaldi@polimi.it [Laboratory of Biological Structure Mechanics, Structural Engineering Department, Politecnico di Milano, Piazza Leonardo da Vinci, 32, 20133 Milan (Italy); Yang Ke; Tan Lili [Division of Specialized Materials and Devices, Institute of Metal Research, Chinese Academy of Sciences, Shenyang (China); Petrini, Lorenza; Migliavacca, Francesco [Laboratory of Biological Structure Mechanics, Structural Engineering Department, Politecnico di Milano, Piazza Leonardo da Vinci, 32, 20133 Milan (Italy)

2011-12-15

Biodegradable magnesium alloy stents (MAS) can provide a great benefit for diseased vessels and avoid the long-term incompatible interactions between vessels and permanent stent platforms. However, the existing MAS showed insufficient scaffolding to the target vessels due to short degradation time. In this study, a three dimensional finite element model combined with a degradable material model of AZ31 (Al 0.03, Zn 0.01, Mn 0.002 and Mg balance, mass percentage) was applied to three different MAS designs including an already implanted stent (Stent A), an optimized design (Stent B) and a patented stent design (Stent C). One ring of each design was implanted through a simulation in a vessel model then degraded with the changing interaction between outer stent surface and the vessel. Results showed that a proper stent design (Stent B) can lead to an increase of nearly 120% in half normalized recoil time of the vessel compared to the Stent A; moreover, the expectation that the MAS design, with more mass and optimized mechanical properties, can increase scaffolding time was verified numerically. The Stent C has more materials than Stent B; however, it only increased the half normalized recoil time of the vessel by nearly 50% compared to the Stent A because of much higher stress concentration than that of Stent B. The 3D model can provide a convenient design and testing tool for novel magnesium alloy stents.

Finite element analyses for design evaluation of biodegradable magnesium alloy stents in arterial vessels

International Nuclear Information System (INIS)

Wu Wei; Gastaldi, Dario; Yang Ke; Tan Lili; Petrini, Lorenza; Migliavacca, Francesco

2011-01-01

Biodegradable magnesium alloy stents (MAS) can provide a great benefit for diseased vessels and avoid the long-term incompatible interactions between vessels and permanent stent platforms. However, the existing MAS showed insufficient scaffolding to the target vessels due to short degradation time. In this study, a three dimensional finite element model combined with a degradable material model of AZ31 (Al 0.03, Zn 0.01, Mn 0.002 and Mg balance, mass percentage) was applied to three different MAS designs including an already implanted stent (Stent A), an optimized design (Stent B) and a patented stent design (Stent C). One ring of each design was implanted through a simulation in a vessel model then degraded with the changing interaction between outer stent surface and the vessel. Results showed that a proper stent design (Stent B) can lead to an increase of nearly 120% in half normalized recoil time of the vessel compared to the Stent A; moreover, the expectation that the MAS design, with more mass and optimized mechanical properties, can increase scaffolding time was verified numerically. The Stent C has more materials than Stent B; however, it only increased the half normalized recoil time of the vessel by nearly 50% compared to the Stent A because of much higher stress concentration than that of Stent B. The 3D model can provide a convenient design and testing tool for novel magnesium alloy stents.

ITER in-vessel system design and performance

International Nuclear Information System (INIS)

Parker, R.R.

1999-01-01

This paper reviews the design and performance of the in-vessel components of ITER as developed for the EDA Final Design Report (FDR). The double-wall vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g., the most intense VDE's and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature differences. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m 2 are expected and these are accommodated by HHF technology developed during the EDA. Disruptions and VDE's can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowables for all postulated disruption and seismic events. (author)

ITER in-vessel system design and performance

International Nuclear Information System (INIS)

Parker, R.R.

2001-01-01

This paper reviews the design and performance of the in-vessel components of ITER as developed for the EDA Final Design Report (FDR). The double-wall vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g., the most intense VDE's and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature differences. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m 2 are expected and these are accommodated by HHF technology developed during the EDA. Disruptions and VDE's can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowables for all postulated disruption and seismic events. (author)

Innovations in prestressed concrete pressure vessel design

International Nuclear Information System (INIS)

Chow, P.Y.; Ngo, D.; Lin, T.Y.

1979-01-01

The study explored a new approach to the design of a high-pressure PCPV that accepts tension and tension cracks in the outer region of the PCPV. It examined the possibility of incorporating artificially-introduced preformed separations that pre-determined crack locations in the design as a method of controlling high tensile stresses generated by internal temperature and pressure. The results showed that the PCPV so designed was, in the extreme case of the DSV, approximately 70% cheaper than the 18 steel vessels of equivalent capacity it replaces. (orig.)

Design and Optimization of Filament Wound Composite Pressure Vessels

NARCIS (Netherlands)

Zu, L.

2012-01-01

One of the most important issues for the design of filament-wound pressure vessels reflects on the determination of the most efficient meridian profiles and related fiber architectures, leading to optimal structural performance. To better understand the design and optimization of filament-wound

Design optimization of a thin walled pressure vessel

International Nuclear Information System (INIS)

Sadiq, S.

2001-01-01

Design evaluation of a pressure vessel is not only to build confidence on its integrity but also to reduce structural weight and enhance the performance of the structure. Pressure vessel, e.g., a rocket motor not only has to withstand the high operating temperatures but it must also be able to survive the internal pressures and external aerodynamic forces and bending stresses during its operation in flight. A research program was devised to study the stresses, which are generated in a thin walled pressure vessel during actual operation and its simulation with cold testing technique, i.e., by means of hydrostatic testing employing electrical resistance strain gauges on the external surface of the cylinder. The objective of the research was to uphold the performance of the vessel by reducing its thickness from 6.09 to 5.5 mm (which of course reduces the safety factor margin from 1.8 to 1.5); thereby curtailing the overall structural weight and maintaining the efficiency of the vessel itself during its live operation. The techniques employed were hydrostatic testing, data acquisition system for obtaining data on strains from the electrical resistance strain gauges and later employing V on Mises yield criterion empirical relation to computer the stresses in hoop and longitudinal directions. (author)

Structural analysis of the KSTAR vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

In, Sang Ryul; Yoon, Byeong Joo [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-10-01

Structure analysis of the vacuum vessel for the KSTAR tokamak which, is in the end phase of the conceptual design have been performed. Mechanical stresses and deformations of the vessel produced by constant forces due to atmospheric pressure, dead weight, fluid pressure, etc and various transient electromagnetic forces induced during tokamak operations were calculated as well as modal characteristics and buckling properties were investigated. Influences of the temperature gradient and the constraint condition of the support on the thermal stress and deformation of the vessel were analyzed. The thermal stress due to the temperature distribution on the vessel as supplying the N{sub 2} gas of 400 deg C through poloidal channels according to the recent baking concept were calculated. No severe problem in the robustness of the vessel was found when applying the constant pressures on the vessel. However the mechanical stress due to the EM force induced by halo currents flowing on the vessel and the plasma facing components (PFCs) far exceeded the allowable limit. Some reinforcing components should be added on the boundary of the PFC support and the vessel, and that of the vessel support and the vessel. A steep temperature gradient in the vicinity of the inlet and oulet of the heating gas produced a thermal stress much higher than allowable. It is necessary to make the temperature of the vessel as uniform as possible and to develop a new support concept which is flexible enough to accommodate a thermal expansion of a few cm while sufficiently strong to resist mechanical impacts. (author). 5 refs., 41 figs., 9 tabs.

The design, fabrication, and testing of WETF high-quality, long-term-storage, secondary containment vessels

International Nuclear Information System (INIS)

Fisher, Kane J.

2000-01-01

Los Alamos National Laboratory's Weapons Engineering Tritium Facility (WETF) requires secondary containment vessels to store primary tritium containment vessels. The primary containment vessel provides the first boundary for tritium containment. The primary containment vessel is stored within a secondary containment vessel that provides the secondary boundary for tritium containment. WETF requires high-quality, long-term-storage, secondary tritium containment vessels that fit within a Mound-designed calorimeter. In order to qualify the WETF high-quality, long-term-storage, secondary containment vessels for use at WETF, steps have been taken to ensure the appropriate design, adequate testing, quality in fabrication, and acceptable documentation

Development of advanced design features for KNGR reactor vessel and internals

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Kyun; Ru, Bong; Lee, Jae Han; Lee, Hyung Yeon; Kim, Jong Bum; Ku, Kyung Heoy; Lee, Ki Young; Lee, Jun; Kim, Young In [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-12-01

Developments of KNGR design require to enhance the design to implement the design requirements, such as plant life time from 40 years to 60 years, safety, performance and structure and components design. The designs used for existing nuclear power plants should be modified or improved to meet the requirements in KNGR design. The purpose of the task is to develop the Advanced Design Features (ADF) related to mechanical and structural design for KNGR reactor vessel and reactor internals. The structural integrity for the System 80+ reactor vessel, of which design life is 60 years, was reviewed. EPRI-URD, CESSAR-DC, and the present design status and characteristics of System 80+ reactor vessel were comparatively studied and the improvement of reactor vessel surveillance program was investigated. The performance and aseismic characteristics of the CE-type CEDM, which will be used in System 80+, are investigated. The driving cycles of CEDM are evaluated for the load follow operation(LFO), of which Mode K is being developed by KAERI. The position of the USNRC, EPRI, ABB-CE, and industries on the elimination of OBE are reviewed, and especially ABB-CE System 80+ FSER is reviewed in detail. For the pre-stage of the verification of the OBE elimination from the design, the review of the seismic responses, i.e.. shear forces and moments, of YGN 3/4 RI was performed and the ratio of OBE response to SSE response was analysed. The screening criteria were reviewed to evaluate the integrity against pressurized thermal shock (PTS) for RV belt-line of System 80+. The evaluation methods for fracture integrity when screening criteria are not met were reviewed. The structural characteristics of IRWST spargers of System 80+ were investigated and the effect of hydrodynamic loads on NSSS was reviewed. 18 figs., 9 tabs., 40 refs. (Author) .new.

Development of advanced design features for KNGR reactor vessel and internals

International Nuclear Information System (INIS)

Park, Jong Kyun; Ru, Bong; Lee, Jae Han; Lee, Hyung Yeon; Kim, Jong Bum; Ku, Kyung Heoy; Lee, Ki Young; Lee, Jun; Kim, Young In

1995-12-01

Developments of KNGR design require to enhance the design to implement the design requirements, such as plant life time from 40 years to 60 years, safety, performance and structure and components design. The designs used for existing nuclear power plants should be modified or improved to meet the requirements in KNGR design. The purpose of the task is to develop the Advanced Design Features (ADF) related to mechanical and structural design for KNGR reactor vessel and reactor internals. The structural integrity for the System 80+ reactor vessel, of which design life is 60 years, was reviewed. EPRI-URD, CESSAR-DC, and the present design status and characteristics of System 80+ reactor vessel were comparatively studied and the improvement of reactor vessel surveillance program was investigated. The performance and aseismic characteristics of the CE-type CEDM, which will be used in System 80+, are investigated. The driving cycles of CEDM are evaluated for the load follow operation(LFO), of which Mode K is being developed by KAERI. The position of the USNRC, EPRI, ABB-CE, and industries on the elimination of OBE are reviewed, and especially ABB-CE System 80+ FSER is reviewed in detail. For the pre-stage of the verification of the OBE elimination from the design, the review of the seismic responses, i.e.. shear forces and moments, of YGN 3/4 RI was performed and the ratio of OBE response to SSE response was analysed. The screening criteria were reviewed to evaluate the integrity against pressurized thermal shock (PTS) for RV belt-line of System 80+. The evaluation methods for fracture integrity when screening criteria are not met were reviewed. The structural characteristics of IRWST spargers of System 80+ were investigated and the effect of hydrodynamic loads on NSSS was reviewed. 18 figs., 9 tabs., 40 refs. (Author) .new

Design of pressure vessels. Part 2

International Nuclear Information System (INIS)

Grandemange, J.M.

2008-01-01

This document deals with the classification of stresses, necessary for the implementation of the mechanical code criteria defined for the pressure vessels of PWR-type reactors. It describes the general approach of design, analysis, and in-service monitoring, the regulatory tests and the modalities of equivalence between industrial construction codes. Content: 1 - damage modes and stresses classification: context, general approach, example of application; 2 - from the design stage to the in-service monitoring: liabilities, design conditions, materials choice and dimensioning, analysis, particular case of pipes and valve parts, in-service monitoring; 3 - regulatory tests: context, tests prescribed by the design and construction rules of PWR mechanical components (RCC-M); 4 - equivalence possibilities between codes: codes for nuclear reactor equipments, convergence between industrial codes and standards; 5 - conclusion. (J.S.)

Pressure vessel design

International Nuclear Information System (INIS)

Annaratone, D.

2007-01-01

This book guides through general and fundamental problems of pressure vessel design. It moreover considers problems which seem to be of lower importance but which turn out to be crucial in the design phase. The basic approach is rigorously scientific with a complete theoretical development of the topics treated, but the analysis is always pushed so far as to offer concrete and precise calculation criteria that can be immediately applied to actual designs. This is accomplished through appropriate algorithms that lead to final equations or to characteristic parameters defined through mathematical equations. The first chapter describes how to achieve verification criteria, the second analyzes a few general problems, such as stresses of the membrane in revolution solids and edge effects. The third chapter deals with cylinders under pressure from the inside, while the fourth focuses on cylinders under pressure from the outside. The fifth chapter covers spheres, and the sixth is about all types of heads. Chapter seven discusses different components of particular shape as well as pipes, with special attention to flanges. The eighth chapter discusses the influence of holes, while the ninth is devoted to the influence of supports. Finally, chapter ten illustrates the fundamental criteria regarding fatigue analysis. Besides the unique approach to the entire work, original contributions can be found in most chapters, thanks to the author's numerous publications on the topic and to studies performed ad hoc for this book. (orig.)

Damage-tolerant design and inspection philosophy for nuclear and other pressure vessels

International Nuclear Information System (INIS)

Adams, N.J.I.

1980-01-01

Statistical analyses of pressure vessel failure rates indicate that, to date, the record is very good. However, the public hazard and environmental consequences of failure in certain industrial processes now give cause for much greater concern. With the exception of an Appendix in ASME III, the current design codes and requirements for new vessels are all based on the assumption that they are free from cracklike defects, but engineers recognize tht such perfect vessels cannot be manufactured. Taking into account failure mechanisms, material properties, pre- and in-service inspection, proof testing, failure statistics and probabilistic methods, views are put forward on how a damage-tolerant design and inspection philosophy may be developed to reduce further the possibility of ''rogue'' vessel failure. 21 refs

Optimized design of an ex-vessel cooling thermosyphon for decay heat removal in SFR

International Nuclear Information System (INIS)

Choi, Jae Young; Jeong, Yong Hoon; Song, Sub Lee; Chang, Soon Heung

2017-01-01

Passive decay heat removal and sodium fire are two major key issues of nuclear safety in sodium-cooled fast reactor (SFR). Several decay heat removal systems (DHR) were suggested for SFR around the world so far. Those DHRS mainly classified into two concepts: Direct reactor cooling system and ex-vessel cooling system. Direct reactor cooling method represented by PDHRS from PGSFR has disadvantages on its additional in-vessel structure and potential sodium fire risk due to the sodium-filled heat exchanger exposed to air. Contrastively, ex-vessel cooling method represented by RVACS from PRISM has low decay heat removal performance, which cannot be applicable to large scale reactors, generally over 1000 MWth. No passive DHRSs which can solve both side of disadvantages has been suggested yet. The goal of this study was to propose ex-vessel cooling system using two-phase closed thermosyphon to compensate the disadvantages of the past DHRSs. Reference reactor was Innovative SFR (iSFR), a pool-type SFR designed by KAIST and featured by extended core lifetime and increased thermal efficiency. Proposed ex-vessel cooling system consisted of 4 trains of thermosyphons and designed to remove 1% of thermal power with 10% of margin. The scopes of this study were design of proposed passive DHRS, validation of system analysis and optimization of system design. Mercury was selected as working fluid to design ex-vessel thermosyphon in consideration of system geometry, operating temperature and required heat flux. SUS 316 with chrome coated liner was selected as case material to resist against high corrosivity of mercury. Thermosyphon evaporator was covered on the surface of reactor vessel as the geometry of hollow shell filled with mercury. Condenser was consisted of finned tube bundles and was located in isolated water pool, the ultimate heat sink. Operation limits and thermal resistance was estimated to guarantee whether the design was adequate. System analysis was conducted by in

LECOTELO - conceptual design, testings and realisation of the main vessel

International Nuclear Information System (INIS)

Ioan, M.; Hororoi, M.

2013-01-01

Lead Corrosion Testing Loop (LECOTELO) facility was conceived to assure all conditions requested by corrosion/erosion tests in pure hot lead for different materials. The main vessel will receive at least 36 different material samples; each of them must be swept on both sides by a lead flow at a very well known speed. Taking into account that the inner system of this vessel is rather complex, it is very important to know the behavior of the vessel at different speeds of the lead flow around the samples. After many simulations of different configurations of the inner components, it was obtained the best inner geometry of the flow which provides the minimum pressure loss between inlet and outlet vessel. Consequently, the design of vessel components was changed in accordance with these new results of simulations and in this moment they are in the manufacturing process. (authors)

Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

Reactor Pressure Vessel (RPV) Acquisition Strategy

Energy Technology Data Exchange (ETDEWEB)

Mizia, Ronald Eugene [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2008-04-01

The Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, TRISO-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. The purpose of this report is to address the acquisition strategy for the NGNP Reactor Pressure Vessel (RPV). This component will be larger than any nuclear reactor pressure vessel presently in service in the United States. The RPV will be taller, larger in diameter, thicker walled, heavier and most likely fabricated at the Idaho National Laboratory (INL) site of multiple subcomponent pieces. The pressure vessel steel can either be a conventional materials already used in the nuclear industry such as listed within ASME A508/A533 specifications or it will be fabricated from newer pressure vessel materials never before used for a nuclear reactor in the US. Each of these characteristics will present a

The JET gas baking plant for DT operation and analysis of tritium permeation and baking gas activation in DTE1

Energy Technology Data Exchange (ETDEWEB)

Pearce, R.J.H.; Andrew, P.; Bryan, S.; Hemmrich, J.L. [JET Joint Undertaking, Abingdon, Oxon (United Kingdom)

1998-07-01

The JET gas baking plant allows the vacuum vessel to be heated for conditioning and plasma operations. The vessel was maintained at 320 deg. C for the JET DT experiments (DTE 1). The design of the plant is outlined with particular reference to the features to provide compatibility with tritium operations. The experience of baking gas activation and tritium permeation into the plant are given, Developmentsto reduce the tritium permeation out of the vessel are considered. (authors)

Design and manufacturing of vacuum vessel of TPE-RX

Energy Technology Data Exchange (ETDEWEB)

Sago, H.; Kaguchi, H.; Orita, J.; Ishigami, Y. [Mitsubishi Heavy Industries Ltd., Kobe (Japan); Urata, K. [Mitsubishi Heavy Industries Ltd. (Japan). Nuclear Energy Systems Engineering Center; Hasegawa, M. [Mitsubishi Electric Co. (Japan). Nuclear Fusion Development; Yagi, Y.; Hirano, Y.; Shimada, T.; Sekine, S.; Sakakita, H. [Electrotechnical Lab. (Japan)

1998-07-01

Construction of a new, large reversed field pinch (RFP) machine called TPE-RX was complete at the end of 1997 as a successor of the previous TPE-1RM20 machine at the Electrotechnical Laboratory (ETL). RFP configuration has been successfully obtained in March 1998. This paper introduces structural design and manufacturing of the vacuum vessel of TPE-RX. The support positions were decided by structural analyses. The structural integrity of the vacuum vessel was evaluated by inelastic analyses. (author)

Design and manufacturing of vacuum vessel of TPE-RX

International Nuclear Information System (INIS)

Sago, H.; Kaguchi, H.; Orita, J.; Ishigami, Y.; Urata, K.

1998-01-01

Construction of a new, large reversed field pinch (RFP) machine called TPE-RX was complete at the end of 1997 as a successor of the previous TPE-1RM20 machine at the Electrotechnical Laboratory (ETL). RFP configuration has been successfully obtained in March 1998. This paper introduces structural design and manufacturing of the vacuum vessel of TPE-RX. The support positions were decided by structural analyses. The structural integrity of the vacuum vessel was evaluated by inelastic analyses. (author)

Design and implementation of visual inspection system handed in tokamak flexible in-vessel robot

International Nuclear Information System (INIS)

Wang, Hesheng; Xu, Lifei; Chen, Weidong

2016-01-01

In-vessel viewing system (IVVS) is a fundamental tool among the remote handling systems for ITER, which is used to providing information on the status of the in-vessel components. The basic functional requirement of in-vessel visual inspection system is to perform a fast intervention with adequate optical resolution. In this paper, we present the software and hardware solution, which is designed and implemented for tokamak in-vessel viewing system that installed on end-effector of flexible in-vessel robot working under vacuum and high temperature. The characteristic of our in-vessel viewing system consists of two parts: binocular heterogeneous vision inspection tool and first wall scene emersion based augment virtuality. The former protected with water-cooled shield is designed to satisfy the basic functional requirement of visual inspection system, which has the capacity of large field of view and high-resolution for detection precision. The latter, achieved by overlaying first wall tiles images onto virtual first wall scene model in 3D virtual reality simulation system, is designed for convenient, intuitive and realistic-looking visual inspection instead of viewing the status of first wall only by real-time monitoring or off-line images sequences. We present the modular division of system, each of them in smaller detail, and go through some of the design choices according to requirements of in-vessel visual inspection task.

Design and implementation of visual inspection system handed in tokamak flexible in-vessel robot

Energy Technology Data Exchange (ETDEWEB)

Wang, Hesheng; Xu, Lifei [Department of Automation, Shanghai Jiao Tong University, Shanghai 200240 (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China (China); Chen, Weidong, E-mail: wdchen@sjtu.edu.cn [Department of Automation, Shanghai Jiao Tong University, Shanghai 200240 (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China (China)

2016-05-15

In-vessel viewing system (IVVS) is a fundamental tool among the remote handling systems for ITER, which is used to providing information on the status of the in-vessel components. The basic functional requirement of in-vessel visual inspection system is to perform a fast intervention with adequate optical resolution. In this paper, we present the software and hardware solution, which is designed and implemented for tokamak in-vessel viewing system that installed on end-effector of flexible in-vessel robot working under vacuum and high temperature. The characteristic of our in-vessel viewing system consists of two parts: binocular heterogeneous vision inspection tool and first wall scene emersion based augment virtuality. The former protected with water-cooled shield is designed to satisfy the basic functional requirement of visual inspection system, which has the capacity of large field of view and high-resolution for detection precision. The latter, achieved by overlaying first wall tiles images onto virtual first wall scene model in 3D virtual reality simulation system, is designed for convenient, intuitive and realistic-looking visual inspection instead of viewing the status of first wall only by real-time monitoring or off-line images sequences. We present the modular division of system, each of them in smaller detail, and go through some of the design choices according to requirements of in-vessel visual inspection task.

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 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. One of the unique feature of the safety philosophy of the GTHTR300 is that a depressurization accident is postulated as a design basis accident in order to show the high level of safety characteristics, though its probability of occurrence is much 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 a depressurization accident. 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. (author)

Design Improvement of Double Pressure Vessel in the In-pile Test Section

Energy Technology Data Exchange (ETDEWEB)

Hong, Jintae; Heo, Sung-Ho; Joung, Chang-Young; Kim, Ka-Hye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

To carry out an irradiation test of nuclear fuels, a nuclear fuel test rig should be fabricated and installed in the in-pile test section (IPS), which is installed in the reactor hall. While carrying out an irradiation test, sealing out coolant which passes through the test rig is one of the most important issues. In particular, although the double pressure vessel is assembled with the IPS head by two o-rings and six bolts, 15.5 MPa of highly pressurized coolant leaks through the gap between the vessel and IPS head. Because the temperature of the coolant in the test loop is 300 .deg. C , and the pool of HANARO is 40 .deg. C, the double pressure vessel is necessary to insulate them. Therefore, a new design to prevent the leakage of coolant needs to be developed. In this study, EB welding technique is considered to assemble the double pressure vessel and the IPS head, and their mechanical design is modified to enable the welding process. In this study, an improved design for sealing out the coolant at the pressure boundary between the double pressure vessel and the IPS head has been developed. An EB weld is applied to seal out the pressure boundary, and its sealing performance is verified by NDE, a cross section test, and a hydraulic pressure test. From the verification test results, the improved design can be used in fabricating the IPS for a nuclear fuel irradiation test.

Gas Tank for Cars

Directory of Open Access Journals (Sweden)

Peter Lorenz

2011-09-01

Full Text Available In this work the development of a highly efficient pressure vessel for liquid petroleum gas (LPG in integral design is described. The pressure vessel can be customized in an optimal available installation space and thus means that the suitable for everyday use of existing modified cars or trucks can be increased.

Design and fabrication methods of FW/blanket and vessel for ITER-FEAT

Energy Technology Data Exchange (ETDEWEB)

Ioki, K. E-mail: iokik@itereu.de; Barabash, V.; Cardella, A.; Elio, F.; Kalinin, G.; Miki, N.; Onozuka, M.; Osaki, T.; Rozov, V.; Sannazzaro, G.; Utin, Y.; Yamada, M.; Yoshimura, H

2001-11-01

Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R and D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R and D performed so far during EDA are still applicable.

Design and fabrication methods of FW/blanket and vessel for ITER-FEAT

International Nuclear Information System (INIS)

Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Kalinin, G.; Miki, N.; Onozuka, M.; Osaki, T.; Rozov, V.; Sannazzaro, G.; Utin, Y.; Yamada, M.; Yoshimura, H.

2001-01-01

Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R and D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R and D performed so far during EDA are still applicable

Offshore support vessel developments for deep water oil and gas E and P

Energy Technology Data Exchange (ETDEWEB)

Dielen, Baldo A.M. [SMIT, Rotterdam (Netherlands)

2008-07-01

The worldwide trend to move towards more exposed locations and deeper waters for O and G exploration and production activities resulted in an increased need for larger and more powerful tugs and offshore support vessels. These vessels must meet higher operational requirements under higher wind and sea-state conditions. This market-driven need, together with technological developments, is leading towards a new generation of powerful and sophisticated offshore support vessels (OSV's). This paper will describe the actual and future trends in OSV design for deep water offshore use. (author)

Rebound coefficient of collisionless gas in a rigid vessel. A model of reflection of field-reversed configuration

International Nuclear Information System (INIS)

Takaku, Yuichi; Hamada, Shigeo

1996-01-01

A system of collisionless neutral gas contained in a rigid vessel is considered as a simple model of reflection of field-reversed configuration (FRC) plasma by a magnetic mirror. The rebound coefficient of the system is calculated as a function of the incident speed of the vessel normalized by the thermal velocity of the gas before reflection. The coefficient is compared with experimental data of FIX (Osaka U.) and FRX-C/T(Los Alamos N.L.). Agreement is good for this simple model. Interesting is that the rebound coefficient takes the smallest value (∼0.365) as the incident speed tends to zero and approaches unity as it tends to infinity. This behavior is reverse to that expected for a system with collision dominated fluid instead of collisionless gas. By examining the rebound coefficient, therefore, it could be successfully inferred whether the ion mean free path in each experiment was longer or shorter than the plasma length. (author)

Cavitation damage prediction for the JSNS mercury target vessel

Energy Technology Data Exchange (ETDEWEB)

Naoe, Takashi, E-mail: naoe.takashi@jaea.go.jp; Kogawa, Hiroyuki; Wakui, Takashi; Haga, Katsuhiro; Teshigawara, Makoto; Kinoshita, Hidetaka; Takada, Hiroshi; Futakawa, Masatoshi

2016-01-15

The liquid mercury target system for the Japan Spallation Neutron Source (JSNS) at the Materials and Life science experimental Facility (MLF) in the Japan Proton Accelerator Research Complex (J-PARC) is designed to produce pulsed neutrons. The mercury target vessel in this system, which is made of type 316L stainless steel, is damaged by pressure wave-induced cavitation due to proton beam bombardment. Currently, cavitation damage is considered to be the dominant factor influencing the service life of the target vessel rather than radiation damage. In this study, cavitation damage to the interior surface of the target vessel was predicted on the basis of accumulated damage data from off-beam and on-beam experiments. The predicted damage was compared with the damage observed in a used target vessel. Furthermore, the effect of injecting gas microbubbles on cavitation damage was predicted through the measurement of the acoustic vibration of the target vessel. It was shown that the predicted depth of cavitation damage is reasonably coincident with the observed results. Moreover, it was confirmed that the injection of gas microbubbles had an effect on cavitation damage.

FFTF thermal-hydraulic testing results affecting piping and vessel component design in LMFBR's

International Nuclear Information System (INIS)

Stover, R.L.; Beaver, T.R.; Chang, S.C.

1983-01-01

The Fast Flux Test Facility completed four years of pre-operational testing in April 1982. This paper describes thermal-hydraulic testing results from this period which impact piping and vessel component design in LMFBRs. Data discussed are piping flow oscillations, piping thermal stratification and vessel upper plenum stratification. Results from testing verified that plant design limits were met

ITER vacuum vessel design and electromagnetic analysis on in-vessel components

International Nuclear Information System (INIS)

Ioki, K.; Johnson, G.; Shimizu, K.; Williamson, D.; Iizuka, T.

1995-01-01

Major functional requirements for the vacuum vessel are to provide the first safety barrier and to support electromagnetic loads due to plasma disruptions and vertical displacement events, and to withstand plausible accidents without losing confinement. A double wall structure concept has been developed for the vacuum vessel due to its beneficial characteristics from the viewpoints of structural integrity and electrical continuity. An electromagnetic analysis of the blanket modules and the vacuum vessel has been performed to investigate force distributions on in-vessel components. According to the vertical displacement events (VDE) scenario, which assumes a critical q-value of 1.5, the total downward vertical force, induced by coupling between the eddy current and external fields, is about 110 MN. We have performed a stress analysis for the vacuum vessel using the VDE disruption forces acting on the blankets, and a maximum stress intensity of 112 MPa was obtained in the vicinity of the lower support of the vessel. (orig.)

Conceptual Design of Electrical Propulsion System for Nuclear Operated Vessel Adventurer

International Nuclear Information System (INIS)

Halimi, B.; Suh, K. Y.

2009-01-01

A design concept of the electric propulsion system for the Nuclear Operated Vessel Adventure (NOVA) is presented. NOVA employs Battery Omnibus Reactor Integral System (BORIS), a liquid metal cooled small fast integral reactor, and Modular Optimized Brayton Integral System (MOBIS), a supercritical CO 2 (SCO 2 ) Brayton cycle as power converter to Naval Application Vessel Integral System (NAVIS)

Design, fabrication and operating experience of Monju ex-vessel fuel storage tank

International Nuclear Information System (INIS)

Yokota, Yoshio; Yamagishi, Yoshiaki; Kuroha, Mitsuo; Inoue, Tatsuya

1995-01-01

In FBRs there are two methods of storing and cooling the spent fuel - the in-vessel storage and the ex-vessel storage. Because of the sodium leaks through the tank at the beginning of pre-operation, the utilization of the ex-vessel fuel storage tank (EVST) of some FBR plant has been changed from the ex-vessel fuel storage to the interim fuel transfer tank. This led to reactor designers focusing on the material, structure and fabrication of the carbon steel sodium storage tanks worldwide. The Monju EVST was at the final stage of the design, when the leaks occurred. The lesson learned from that experience and the domestic fabrication technology are reflected to the design and fabrication of the Monju EVST. This paper describes the design, fabrication and R and D results for the tank, and operating experience in functional test. The items to be examined are as follows: (1) Overall structure of the tank and design philosophy on the function, (2) Structure of the cover shielding plug and its design philosophy, (3) Structures of the rotating rack and its bearings, and their design philosophy, (4) Cooling method and its design philosophy, (5) Structure and fabrication of the cooling coil support inside EVST with comparison of leaked case, (6) R and D effort for items above. The fabrication of the Monju EVST started in August 1986 and it was shipped to the site in March 1990. Installation was completed in November 1990, and sodium fill after pre-heating started in 1991. The operation has been continued since September 1992. In 1996 when the first spent fuel is stored, its total functions will be examined. (author)

Containment vessel drain system

Science.gov (United States)

Harris, Scott G.

2018-01-30

A system for draining a containment vessel may include a drain inlet located in a lower portion of the containment vessel. The containment vessel may be at least partially filled with a liquid, and the drain inlet may be located below a surface of the liquid. The system may further comprise an inlet located in an upper portion of the containment vessel. The inlet may be configured to insert pressurized gas into the containment vessel to form a pressurized region above the surface of the liquid, and the pressurized region may operate to apply a surface pressure that forces the liquid into the drain inlet. Additionally, a fluid separation device may be operatively connected to the drain inlet. The fluid separation device may be configured to separate the liquid from the pressurized gas that enters the drain inlet after the surface of the liquid falls below the drain inlet.

Design of pressure vessels using shape optimization: An integrated approach

Energy Technology Data Exchange (ETDEWEB)

Carbonari, R.C., E-mail: ronny@usp.br [Department of Mechatronic Engineering, Escola Politecnica da Universidade de Sao Paulo, Av. Prof. Mello Moraes, 2231 Sao Paulo, SP 05508-900 (Brazil); Munoz-Rojas, P.A., E-mail: pablo@joinville.udesc.br [Department of Mechanical Engineering, Universidade do Estado de Santa Catarina, Bom Retiro, Joinville, SC 89223-100 (Brazil); Andrade, E.Q., E-mail: edmundoq@petrobras.com.br [CENPES, PDP/Metodos Cientificos, Petrobras (Brazil); Paulino, G.H., E-mail: paulino@uiuc.edu [Newmark Laboratory, Department of Civil and Environmental Engineering, University of Illinois at Urbana-Champaign, 205 North Mathews Av., Urbana, IL 61801 (United States); Department of Mechanical Science and Engineering, University of Illinois at Urbana-Champaign, 158 Mechanical Engineering Building, 1206 West Green Street, Urbana, IL 61801-2906 (United States); Nishimoto, K., E-mail: knishimo@usp.br [Department of Naval Architecture and Ocean Engineering, Escola Politecnica da Universidade de Sao Paulo, Av. Prof. Mello Moraes, 2231 Sao Paulo, SP 05508-900 (Brazil); Silva, E.C.N., E-mail: ecnsilva@usp.br [Department of Mechatronic Engineering, Escola Politecnica da Universidade de Sao Paulo, Av. Prof. Mello Moraes, 2231 Sao Paulo, SP 05508-900 (Brazil)

2011-05-15

Previous papers related to the optimization of pressure vessels have considered the optimization of the nozzle independently from the dished end. This approach generates problems such as thickness variation from nozzle to dished end (coupling cylindrical region) and, as a consequence, it reduces the optimality of the final result which may also be influenced by the boundary conditions. Thus, this work discusses shape optimization of axisymmetric pressure vessels considering an integrated approach in which the entire pressure vessel model is used in conjunction with a multi-objective function that aims to minimize the von-Mises mechanical stress from nozzle to head. Representative examples are examined and solutions obtained for the entire vessel considering temperature and pressure loading. It is noteworthy that different shapes from the usual ones are obtained. Even though such different shapes may not be profitable considering present manufacturing processes, they may be competitive for future manufacturing technologies, and contribute to a better understanding of the actual influence of shape in the behavior of pressure vessels. - Highlights: > Shape optimization of entire pressure vessel considering an integrated approach. > By increasing the number of spline knots, the convergence stability is improved. > The null angle condition gives lower stress values resulting in a better design. > The cylinder stresses are very sensitive to the cylinder length. > The shape optimization of the entire vessel must be considered for cylinder length.

Scoping calculations for design and analysis of large reactor vessels for liquid-metal fast breeder reactor (LMFBR) plants

International Nuclear Information System (INIS)

Fiala, C.; Kulak, R.F.; Ma, D.C.; Pan, Y.C.; Seidensticker, R.W.; Wang, C.Y.; Zeuch, W.R.

1982-01-01

Reactor vessels for commercial-sized LMFBR plants are quite large - ranging 40 to 70 ft in diameter and 50 to 70 ft in overall depth. These stainless steel vessels contain liquid sodium at relatively low pressures, but at high temperatures. The resulting thin-walled vessels present the structural designer and analyst with special problems, particularly in providing a balanced design to accommodate seismic loads, design basis accident loads, and thermal loadings. A comprehensive set of scoping calculations - though preliminary in detail and depth of design - provides substantial guidance to the vessel designer for subsequent design iterations. Emphasis is placed on the analysis of the large-diameter top closure of the vessel - the deck structure

On the Adequacy of API 521 Relief-Valve Sizing Method for Gas-Filled Pressure Vessels Exposed to Fire

DEFF Research Database (Denmark)

Andreasen, Anders; Nieto, Marcos Zan; Borroni, Filippo

2018-01-01

sense of security. Often the vessel wall will be weakened by high temperatures, before the PRV relieving pressure is reached. In this article, a multiparameter study has been performed taking into consideration various vessel sizes, design pressures (implicitly vessel wall thickness), vessel operating...

GT-MHR design, performance, and safety

International Nuclear Information System (INIS)

Neylan, A.J.; Shenoy, A.; Silady, F.A.; Dunn, T.D.

1994-11-01

The Gas Turbine-Modular Helium Reactor (GT-MHR) is the result of coupling the evolution of a low power density passively safe modular reactor with key technology developments in the U.S. during the last decade: large industrial gas turbines; large active magnetic bearings; and compact, highly effective plate-fin heat exchangers. This is accomplished through the unique use of the Brayton cycle to produce electricity with the helium as primary coolant from the reactor directly driving the gas turbine electrical generator. This cycle can achieve a high net efficiency in the range of 45% to 48%. In the design of the GT-MHR the desirable inherent characteristics of the inert helium coolant, graphite core, and the coated fuel particles are supplemented with specific design features such as passive heat removal to achieve the safety objective of not disturbing the normal day-to-day activities of the public even for beyond design basis rare accidents. Each GT-MHR plant consists of four modules. The GT-MHR module components are contained within steel pressure vessels: a reactor vessel, a power conversion vessel, and a connecting cross vessel. All vessels are sited underground in a concrete silo, which serves as an independent vented low pressure containment structure. By capitalizing on industrial and aerospace gas turbine development, highly effective heat exchanger designs, and inherent gas cooled reactor temperature characteristics, the passively safe GT-MHR provides a sound technical, monetary, and environmental basis for new nuclear power generating capacity. This paper provides an update on the status of the design, which has been under development on the US-DOE program since February 1993. An assessment of plant performance and safety is also included

To the problem of reinforced concrete reactor vessel design and calculation

International Nuclear Information System (INIS)

Kirillov, A.P.; Artem'ev, V.P.; Bogopol'skij, V.G.; Nikolaev, Yu.B.; Paushkin, A.G.

1980-01-01

Modern methods for calculating reactor vessels of prestressed reinforced concrete are analyzed. It is shown that during the stage of technical and economical substantiation of reactor vessel structure for determining its stressed-deformed state engineering methods of calculation must be used, in particular, fragmentation method, method of rings and plates, and during the stages of contract and detail designs - method of finite elements and dynamic relaxation method. It is concluded that when solving cyclic symmetrical problems as well as asymmetrical problems, calculational algorithms for axis-symmetrical distributions of stresses in the vessel with provision for elastic properties of structural material may be used

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

Gas Hydrate Storage of Natural Gas

Energy Technology Data Exchange (ETDEWEB)

Rudy Rogers; John Etheridge

2006-03-31

Environmental and economic benefits could accrue from a safe, above-ground, natural-gas storage process allowing electric power plants to utilize natural gas for peak load demands; numerous other applications of a gas storage process exist. A laboratory study conducted in 1999 to determine the feasibility of a gas-hydrates storage process looked promising. The subsequent scale-up of the process was designed to preserve important features of the laboratory apparatus: (1) symmetry of hydrate accumulation, (2) favorable surface area to volume ratio, (3) heat exchanger surfaces serving as hydrate adsorption surfaces, (4) refrigeration system to remove heat liberated from bulk hydrate formation, (5) rapid hydrate formation in a non-stirred system, (6) hydrate self-packing, and (7) heat-exchanger/adsorption plates serving dual purposes to add or extract energy for hydrate formation or decomposition. The hydrate formation/storage/decomposition Proof-of-Concept (POC) pressure vessel and supporting equipment were designed, constructed, and tested. This final report details the design of the scaled POC gas-hydrate storage process, some comments on its fabrication and installation, checkout of the equipment, procedures for conducting the experimental tests, and the test results. The design, construction, and installation of the equipment were on budget target, as was the tests that were subsequently conducted. The budget proposed was met. The primary goal of storing 5000-scf of natural gas in the gas hydrates was exceeded in the final test, as 5289-scf of gas storage was achieved in 54.33 hours. After this 54.33-hour period, as pressure in the formation vessel declined, additional gas went into the hydrates until equilibrium pressure/temperature was reached, so that ultimately more than the 5289-scf storage was achieved. The time required to store the 5000-scf (48.1 hours of operating time) was longer than designed. The lower gas hydrate formation rate is attributed to a

Comprehending the structure of a vacuum vessel and in-vessel components of fusion machines. 1. Comprehending the vacuum vessel structure

International Nuclear Information System (INIS)

Onozuka, Masanori; Nakahira, Masataka

2006-01-01

The functions, conditions and structure of vacuum vessel using tokamak fusion machines are explained. The structural standard and code of vacuum vessel, process of vacuum vessel design, and design of ITER vacuum vessel are described. Production and maintenance of ultra high vacuum, confinement of radioactive materials, support of machines in vessel and electromagnetic force, radiation shield, plasma vertical stability, one-turn electric resistance, high temperature baking heat and remove of nuclear heat, reduce of troidal ripple, structural standard, features of safety of nuclear fusion machines, subjects of structural standard of fusion vacuum vessel, design flow of vacuum vessel, establishment of radial build, selections of materials, baking and cooling method, basic structure, structure of special parts, shield structure, and of support structure, and example of design of structure, ITER, are stated. (S.Y.)

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)

Single pressure vessel (SPV) nickel-hydrogen battery design

Energy Technology Data Exchange (ETDEWEB)

Coates, D.; Grindstaff, B.; Fox, C. [Eagle-Picher Industries, Inc., Joplin, MO (United States)

1995-07-01

Single pressure vessel (SPV) technology combines an entire multi-cell nickel-hydrogen (NiH{sub 2}) space battery within a single pressure vessel. SPV technology has been developed to improve the performance (volume/mass) of the NiH{sub 2} system at the battery level and ultimately to reduce overall battery cost and increase system reliability. Three distinct SPV technologies are currently under development and in production. Eagle-Picher has license to the COMSAT Laboratories technology, as well as internally developed independent SPV technology. A third technology resulted from the acquisition of Johnson Controls NiH{sub 2} battery assets in June, 1994. SPV batteries are currently being produced in 25 ampere-hour (Ah), 35 Ah and 50 Ah configurations. The battery designs have an overall outside diameter of 10 inches (25.4 centimeters).

Thermohydraulics in a high-temperature gas-cooled reactor prestressed-concrete reactor vessel during unrestricted core-heatup accidents

International Nuclear Information System (INIS)

Kroeger, P.G.; Colman, J.; Araj, K.

1983-01-01

The hypothetical accident considered for siting considerations in High Temperature Gas-Cooled Reactors (HTGR) is the so called Unrestricted Core Heatup Accident (UCHA), in which all forced circulation is lost at initiation, and none of the auxillary cooling loops can be started. The result is a gradual slow core heatup, extending over days. Whether the liner cooling system (LCS) operates during this time is of crucial importance. If it does not, the resulting concrete decomposition of the prestressed concrete reactor vessel (PCRV) will ultimately cause containment building (CB) failure after about 6 to 10 days. The primary objective of the work described here was to establish for such accident conditions the core temperatures and approximate fuel failure rates, to check for potential thermal barrier failures, and to follow the PCRV concrete temperatures, as well as PCRV gas releases from concrete decomposition. The work was done for the General Atomic Corporation Base Line Zero reactor of 2240 MW(t). Most results apply at least qualitatively also to other large HTGR steam cycle designs

High methane natural gas/air explosion characteristics in confined vessel.

Science.gov (United States)

Tang, Chenglong; Zhang, Shuang; Si, Zhanbo; Huang, Zuohua; Zhang, Kongming; Jin, Zebing

2014-08-15

The explosion characteristics of high methane fraction natural gas were investigated in a constant volume combustion vessel at different initial conditions. Results show that with the increase of initial pressure, the peak explosion pressure, the maximum rate of pressure rise increase due to a higher amount (mass) of flammable mixture, which delivers an increased amount of heat. The increased total flame duration and flame development time result as a consequence of the higher amount of flammable mixture. With the increase of the initial temperature, the peak explosion pressures decrease, but the pressure increase during combustion is accelerated, which indicates a faster flame speed and heat release rate. The maximum value of the explosion pressure, the maximum rate of pressure rise, the minimum total combustion duration and the minimum flame development time is observed when the equivalence ratio of the mixture is 1.1. Additionally, for higher methane fraction natural gas, the explosion pressure and the maximum rate of pressure rise are slightly decreased, while the combustion duration is postponed. The combustion phasing is empirically correlated with the experimental parameters with good fitting performance. Furthermore, the addition of dilute gas significantly reduces the explosion pressure, the maximum rate of pressure rise and postpones the flame development and this flame retarding effect of carbon dioxide is stronger than that of nitrogen. Copyright © 2014 Elsevier B.V. All rights reserved.

Olivine, dolomite and ceramic filters in one vessel to produce clean gas from biomass.

Science.gov (United States)

Rapagnà, Sergio; Gallucci, Katia; Foscolo, Pier Ugo

2018-01-01

Heavy organic compounds produced during almond shells gasification in a steam and/or air atmosphere, usually called tar, are drastically reduced in the product gas by using simultaneously in one vessel a ceramic filter placed in the freeboard and a mixture of olivine and dolomite particles in the fluidized bed of the gasifier. The content of tar in the product gas during a reference gasification test with air, in presence of fresh olivine particles only, was 8600mg/Nm 3 of dry gas. By gasifying biomass with steam at the same temperature level of 820°C in a bed of olivine and dolomite (20% by weight), and in the presence of a catalytic ceramic filter inserted in the freeboard of the fluidized bed gasifier, the level of tar was brought down to 57mg/Nm 3 of dry producct gas, with a decrease of more than two orders of magnitude. Copyright © 2017 Elsevier Ltd. All rights reserved.

Fatigue of non-welded pressure vessels made of high strength steel

International Nuclear Information System (INIS)

Rauscher, F.

2003-01-01

When using high strength steels for pressure vessels, cyclic fatigue requirements may become decisive for the design. Within a European research project, two typical non-welded types of vessels--gas cylinders as used for gas transportation and hydraulic accumulators with screwed in ends--were investigated. The results of the fatigue analyses and of the testing of these vessels are described here. Special attention is drawn to the evaluation of the stresses in the threads used for threaded in flat ends and rings, because the usual formulae for bolted connections cannot be used. In the case of sharp notches and of threads, the experiments showed that the fatigue calculation gave conservative results. The unexpected failure of the gas cylinders in the cylindrical part and at the onset of the end showed that the fatigue analyses according to prEN13445-3 clause 18 is non-conservative for these surfaces without mechanical preparation, and need special consideration. Based on the investigations, a stress concentration factor for small fabrication notches and a new surface finish factor is proposed

Fatigue of non-welded pressure vessels made of high strength steel

Energy Technology Data Exchange (ETDEWEB)

Rauscher, F

2003-03-01

When using high strength steels for pressure vessels, cyclic fatigue requirements may become decisive for the design. Within a European research project, two typical non-welded types of vessels--gas cylinders as used for gas transportation and hydraulic accumulators with screwed in ends--were investigated. The results of the fatigue analyses and of the testing of these vessels are described here. Special attention is drawn to the evaluation of the stresses in the threads used for threaded in flat ends and rings, because the usual formulae for bolted connections cannot be used. In the case of sharp notches and of threads, the experiments showed that the fatigue calculation gave conservative results. The unexpected failure of the gas cylinders in the cylindrical part and at the onset of the end showed that the fatigue analyses according to prEN13445-3 clause 18 is non-conservative for these surfaces without mechanical preparation, and need special consideration. Based on the investigations, a stress concentration factor for small fabrication notches and a new surface finish factor is proposed.

Design of the ITER vacuum vessel

International Nuclear Information System (INIS)

Ioki, K.; Johnson, G.; Shimizu, K.; Williamson, D.

1995-01-01

The ITER vacuum vessel is a major safety barrier and must support electromagnetic loads during plasma disruptions and vertical displacement events (VDE) and withstand plausible accidents without losing confinement.The vacuum vessel has a double wall structure to provide structural and electrical continuity in the toroidal direction. The inner and outer shells and poloidal stiffening ribs between them are joined by welding, which gives the vessel the required mechanical strength. The space between the shells will be filled with steel balls and plate inserts to provide additional nuclear shielding. Water flowing in this space is required to remove nuclear heat deposition, which is 0.2-2.5% of the total fusion power. The minor and major radii of the tokamak are 3.9 m and 13 m respectively, and the overall height is 15 m. The total thickness of the vessel wall structure is 0.4-0.7 m.The inboard and outboard blanket segments are supported from the vacuum vessel. The support structure is required to withstand a large total vertical force of 200-300 MN due to VDE and to allow for differential thermal expansion.The first candidate for the vacuum vessel material is Inconel 625, due to its higher electric resistivity and higher yield strength, even at high temperatures. Type 316 stainless steel is also considered a vacuum vessel material candidate, owing to its large database and because it is supported by more conventional fabrication technology. (orig.)

Baking results of KSTAR vacuum vessel

International Nuclear Information System (INIS)

Kim, S. T.; Kim, Y. J.; Kim, K. M.; Im, D. S.; Joung, N. Y.; Yang, H. L.; Kim, Y. S.; Kwon, M.

2009-01-01

The Korea Superconducting Tokamak Advanced Research (KSTAR) is an advanced superconducting tokamak designed to establish a scientific and technological basis for an attractive fusion reactor. The fusion energy in the tokamak device is released through fusion reactions of light atoms such as deuterium or helium in hot plasma state, of which temperature reaches several hundreds of millions Celsius. The high temperature plasma is created in the vacuum vessel that provides ultra high vacuum status. Accordingly, it is most important for the vacuum condition to keep clean not only inner space but also surface of the vacuum vessel to make high quality plasma. There are two methods planned to clean the wall surface of the KSTAR vacuum vessel. One is surface baking and the other is glow discharge cleaning (GDC). To bake the vacuum vessel, De-Ionized (DI) water is heated to 130 .deg. C and circulated in the passage between double walls of the vacuum vessel (VV) in order to bake the surface. The GDC operation uses hydrogen and inert gas discharges. In this paper, general configuration and brief introduction of the baking result will be reported

Baking results of KSTAR vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, S. T.; Kim, Y. J.; Kim, K. M.; Im, D. S.; Joung, N. Y.; Yang, H. L.; Kim, Y. S.; Kwon, M. [National Fusion Research Institute, Daejeon (Korea, Republic of)

2009-05-15

The Korea Superconducting Tokamak Advanced Research (KSTAR) is an advanced superconducting tokamak designed to establish a scientific and technological basis for an attractive fusion reactor. The fusion energy in the tokamak device is released through fusion reactions of light atoms such as deuterium or helium in hot plasma state, of which temperature reaches several hundreds of millions Celsius. The high temperature plasma is created in the vacuum vessel that provides ultra high vacuum status. Accordingly, it is most important for the vacuum condition to keep clean not only inner space but also surface of the vacuum vessel to make high quality plasma. There are two methods planned to clean the wall surface of the KSTAR vacuum vessel. One is surface baking and the other is glow discharge cleaning (GDC). To bake the vacuum vessel, De-Ionized (DI) water is heated to 130 .deg. C and circulated in the passage between double walls of the vacuum vessel (VV) in order to bake the surface. The GDC operation uses hydrogen and inert gas discharges. In this paper, general configuration and brief introduction of the baking result will be reported.

An interior vessel viewing system for DIII-D

International Nuclear Information System (INIS)

Senior, R.

1989-11-01

It was anticipated that there could be damage to the interior walls of the vacuum vessel during operations of the DIII-D tokamak. A method of viewing the inside of the vessel from the outside was required, that would allow the interior walls to be inspected visually for damage and to locate any debris resulting from operations. A miniature closed circuit television color camera system was developed which could be inserted into one of several ports of the vessel during a 'clean' vent, i.e., vented to inert gas. The system has pan, tilt and zoom capability and carries its own lighting. The use of this system allows a quick assessment of the condition of the vessel to be made under 'clean' vent conditions. This precludes the need for the permit process and manned entry into the vessel which would allow air inside the vessel. A permanent record of the inspection can then be made on video tape. The design and configuration of this camera system is presented and its use as a diagnostic tool discussed. 2 refs., 5 figs

An introduction to the analysis of multi-cavity prestressed concrete pressure vessels

International Nuclear Information System (INIS)

Silva, M.C.A.T. da.

1986-01-01

The present work is a study of multi-cavity prestressed concrete pressure vessels (PCRV) for nuclear reactors. A review is made of the designs, analises and models of multi-cavity concrete pressure vessels. A preliminary evaluation of the NONSAP program for applications in complex three-dimensional structures such as a multi-cavity pressure vessel is also made. A model of a PCRV of a 1000 MW(e) high-temperature gas cooled reactor was selected for a three-dimensional analysis with the NONSAP program. The results obtained are compared with experimental data. (Author) [pt

An experimental study on coolability through the external reactor vessel cooling according to RPV insulation design

Energy Technology Data Exchange (ETDEWEB)

Kang, Kyoung Ho; Koo, Kil Mo; Park, Rae Joon; Cho, Young Ro; Kim, Sang Baik

2004-01-01

LAVA-ERVC experiments have been performed to investigate the effect of insulation design features on the water accessibility and coolability in case of the external reactor vessel cooling. Alumina iron thermite melt was used as corium stimulant. And the hemispherical test vessel is linearly scaled-down of RPV lower plenum. 4 tests have been performed varying the melt composition and the configuration of the insulation system. Due to the limited steam venting capacity through the insulation, steam binding occurred inside the annulus in the LAVA- ERVC-1, 2 tests which were performed for simulating the KSNP insulation design. This steam binding brought about incident heat up of the vessel outer surface at the upper part in the LAVA-ERVC-1, 2 tests. On the contrary, in the LAVA-ERVC-3, 4 tests which were performed for simulating the APR1400 insulation design, the temperatures of the vessel outer surface maintained near saturation temperature. Sufficient water ingression and steam venting through the insulation lead to effective cooldown of the vessel characterized by nucleate boiling in the LAVA-ERVC-3, 4 tests. From the LAVA-ERVC experimental results, it could be preliminarily concluded that if pertinent modification of the insulation design focused on the improvement of water ingression and steam venting should be preceded the possibility of in-vessel corium retention through the external vessel cooling could be considerably increased.

Review of the Conceptual Design for In-Vessel Fuel Handling Machines in SFR

International Nuclear Information System (INIS)

Kim, S. H.; Koo, G. H.

2012-01-01

The main in-vessel fuel handling machines in sodium cooled fast reactor(SFR) are composed of the in-vessel transfer machine(IVTM) and the rotating plug. These machines perform the function to handle fuel assemblies inside the reactor core during the refueling time. The IVTM should be able to access all areas above the reactor core and the fuel transfer port which can discharge the fuel assembly by the rotation of the rotating plug. In the 600 MWe demonstration reactor, the conceptual design of the in-vessel fuel handling machines was carried out. As shown in Fig. 1, the invessel fuel handling machines of the demonstration reactor are the double rotating plug type. With reference to the given core configuration of the demonstration reactor, the arrangement design of the rotating plug was carried out by using the developed simulation program. At present, the conceptual design of SFR prototype reactor which has small capacity of about 100 MWe is being started. Thus, it is necessary the economical efficiency and the reliability of the in-vessel fuel handling machines are reviewed according to the reduction of the power capacity. In this study, the preliminary design concepts of the main invessel fuel handling machines according to the fuel handling type are compared. Also, the design characteristics for the driving mechanism of the IVTM in the demonstration reactor and the recovery concept from the malfunction are reviewed

Conceptual design of helium gas turbine for MHTGR-GT

International Nuclear Information System (INIS)

Matsuo, E.; Tsutsumi, M.; Ogata, K.; Nomura, S.

1996-01-01

Conceptual designs of the direct-cycle helium gas turbine for a practical unit (450 MWt) and an experimental unit (1200kWt) of MHTGR were conducted and the results as shown below were obtained. The power conversion vessel for this practical unit can further be downsized to an outside diameter of 7.4m and a height of 22m as compared with the conventional design examples. Comparison of the conceptual designs of helium gas turbines using single-shaft type employing the axial-flow compressor and twin-shaft type employing the centrifugal compressor shows that the former provides advantages in terms of structure and control designs whereas the latter offers a higher efficiency. In order to determine which of them should be selected, a further study to investigate various aspects of safety features and startup characteristics will be needed. Either of the two types can provide a cycle efficiency of 46 to 48%. The third mode natural frequencies of the twin-shart type's low-pressure rotational shaft and the single shaft type are below the designed rotational speed, but their vibrational controls are made available using the magnetic bearing system. Elevation of the natural frequency for the twin-shaft type would be possible by altering the arrangements of its shafting configuration. As compared with the earlier conceptual designs, the overall systems configuration can be made simpler and more compact; five stages of turbines for the single-shaft type and seven stages of turbines for the twin-shaft type employing one shaft for the low-pressure compressor and the power turbine and; 26 stages of compressors for the axial-flow type with the single shaft system and five stages of compressors for the centrifugal type with the twin-shaft system. 9 refs, 12 figs, 4 tabs

Design study on steam generator integration into the VVER reactor pressure vessel

International Nuclear Information System (INIS)

Hort, J.; Matal, O.

2004-01-01

The primary circuit of VVER (PWR) units is arranged into loops where the heat generated by the reactor is removed by means of main circulating pumps, loop pipelines and steam generators, all located outside the reactor pressure vessel. If the primary circuit and reactor core were integrated into one pressure vessel, as proposed, e.g., within the IRIS project (WEC), a LOCA situation would be limited by the reactor pressure vessel integrity only. The aim of this design study regarding the integration of the steam generator into the reactor pressure vessel was to identify the feasibility limits and some issues. Fuel elements and the reactor pressure vessel as used in the Temelin NPP were considered for the analysis. From among the variants analyzed, the variant with steam generators located above the core and vertically oriented circulating pumps at the RPV lower bottom seems to be very promising for future applications

Application of the ASME code in designing containment vessels for packages used to transport radioactive materials

International Nuclear Information System (INIS)

Raske, D.T.; Wang, Z.

1992-01-01

The primary concern governing the design of shipping packages containing radioactive materials is public safety during transport. When these shipments are within the regulatory jurisdiction of the US Department of Energy, the recommended design criterion for the primary containment vessel is either Section III or Section VIII, Division 1, of the ASME Boiler and Pressure Vessel Code, depending on the activity of the contents. The objective of this paper is to discuss the design of a prototypic containment vessel representative of a packaging for the transport of high-level radioactive material

Conceptual design finalisation of the ITER In-Vessel Viewing and Metrology System (IVVS)

Energy Technology Data Exchange (ETDEWEB)

Dubus, Gregory, E-mail: gregory.dubus@f4e.europa.eu [Fusion for Energy, c/ Josep Pla, n°2 - Torres Diagonal Litoral - Edificio B3, 08019 Barcelona (Spain); Puiu, Adrian; Damiani, Carlo; Van Uffelen, Marco; Lo Bue, Alessandro; Izquierdo, Jesus; Semeraro, Luigi [Fusion for Energy, c/ Josep Pla, n°2 - Torres Diagonal Litoral - Edificio B3, 08019 Barcelona (Spain); Martins, Jean-Pierre; Palmer, Jim [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

2013-10-15

The In-Vessel Viewing and Metrology System (IVVS) is a fundamental tool for the ITER machine operations, aiming at performing inspections as well as providing information related to the erosion of in-vessel components. Periodically or on request, the IVVS probes will be deployed into the Vacuum Vessel from their storage positions (still within the ITER primary confinement) in order to perform both viewing and metrology on plasma facing components (blanket, divertor, heating/diagnostic plugs, test blanket modules) and, more generically, to provide information on the status of the in-vessel components. In 2011, the IO proposed to simplify and strengthen the six IVVS port extensions situated at the divertor level. Among other important consequences, such as the relocation of the Glow Discharge Cleaning (GDC) electrodes at other levels of the machine, this major design change implied the need for a substantial redesign of the IVVS plug, which took part to an on-going effort to bring the integrated IVVS concept – including the scanning probe and its deployment system – to the level of maturity suitable for the Conceptual Design Review. This paper gives an overview of the various design and R and D activities in progress: plug design integration, probe concept validation under environmental conditions, development of a metrology strategy, the whole supported by a nuclear analysis.

Conceptual design of the handling and storage system of the spent target vessel for neutron scattering facility 2

International Nuclear Information System (INIS)

Adachi, Junichi; Kaminaga, Masanori; Sasaki, Shinobu; Haga, Katsuhiro; Aso, Tomokazu; Kinoshita, Hidetaka; Hino, Ryutaro

2002-01-01

In designing the neutron scattering facility, a spent target vessel should be replaced with remote handling devices in order to protect radioactive exposure, since it would be highly activated through the high energy neutron irradiation caused by the spallation reaction between mercury of the target material and the MW-class proton beam. In the storage of the spent target vessel, it is necessary to consider decay heat of the target vessel and mercury contamination caused by vaporization of the residual mercury in the vessel. A conceptual design has been carried out to establish basic concept and to clarify its specification of main equipments on handling and storage systems for the spent target vessel. This report presents the basic concept and a system plot plan based on latest design works of remote handling devices such as a spent target vessel storage cask and a target vessel exchange trolley, which aim at reasonability and simplification. In addition, storage systems for the spent moderator vessel, the spent proton beam window and the spent reflector vessel are also investigated based on the plot plan. (author)

Stress analysis of blanket vessel for JAERI experimental fusion reactor

International Nuclear Information System (INIS)

Sako, K.; Minato, A.

1979-01-01

A blanket structure of JAERI Experimental Fusion Reactor (JXFR) consists of about 2,300 blanket cells with round cornered rectangular cross sections (twelve slightly different shapes) and is placed in a vacuum vessel. Each blanket vessel is a double-walled thin-shell structure made of Type 316 stainless steel with a spherical domed surface at the plasma side. Ribs for coolant channel are provided between inner and outer walls. The blanket cell contains Li 2 O pebbles and blocks for tritium breeding and stainless steel blocks for neutron reflection. A coolant is helium gas at 10 kgf/cm 2 (0.98 MPa) and its inlet and outlet temperatures are 300 0 C and 500 0 C. The maxima of heat flux and nuclear heating rate at the first wall are 12 W/cm 2 and 2 W/cc. A design philosophy of the blanket structure is based on high tritium breeding ratio and more effective shielding performance. The thin-shell vessel with a rectangular cross section satisfies the design philosophy. We have designed the blanket structure so that the adjacent vessels are mutually supporting in order to decrease the large deformation and stress due to internal pressure in case of the thin-shell vessel. (orig.)

Application of gas shielded arc welding and submerged arc welding for fabrication of nuclear reactor vessels

International Nuclear Information System (INIS)

Gehani, M.L.; Rodrigues, W.D.

1976-01-01

The remarkable progress made in the development of knowhow and expertise in the manufacture of equipment for nuclear power plants in India is outlined. Some of the specific advances made in the application of higher efficiency weld processes for fabrication of nuclear reactor vessels and the higher level of quality attained are discussed in detail. Modifications and developments in submerged arc, gas tungsten arc and gas metal arc processes for welding of Calandria which have been a highly challenging and rewarding experience are discussed. Future scope for making the gas metal arc process more economical by using various gas-mixes like Agron + Oxygen, Argon + Carbon Dioxide, Argon + Nitrogen (for Copper Alloys) etc., in various proportions are outlined. Quality and dimensional control exercised in these jobs of high precision are highlighted. (K.B.)

State-of-the-art and prospets for designing and constraction of prestressed concrete pressure vessels

International Nuclear Information System (INIS)

Anon.

1985-01-01

Short review of reports submitted to the symposium on pressure vessels, which was conducted in Calgary (Canada), has been presented. New tendencies of designing of prestressed concrete pressure vessels (PCPV) for nuclear for nuclear reactors are noted. Construction of hot vessel liner is studied. A conclusion is drawn on prospects of PCPV creation

The design of lifting attachments for the erection of large diameter and heavy wall pressure vessels

International Nuclear Information System (INIS)

Antalffy, Leslie P.; Miller, George A.; Kirkpatrick, Kenneth D.; Rajguru, Anil; Zhu, Yong

2016-01-01

Lifting attachments for the erection of large diameter and heavy wall pressure vessels require special consideration to ensure that their attachment to their vessel shells or heads do not overstress the vessel during the erection process when lifting these from grade onto their respective foundations. Today, in refinery and petrochemical services, large diameter vessels with diameters ranging up to 15 m and reactors with lifting weights in the range of 700–1400 tons are not uncommon. In today's fabrication market, these vessels may be purchased and fabricated in shops dispersed globally and will require unique equipment for their safe handling, transportation and subsequent erection. The challenge is to design the lifting attachments in such a manner that the attachments provide a safe, cost effective and effective solution based upon the limitations of the job site lift equipment available for erection. Such equipment for the transportation and subsequent lifting of large diameter and heavy wall pressure equipment is usually scarce and quite expensive. Planning ahead, well in advance of the lift date is almost a mandatory requirement. Usually, the specific parameters of the vessel to be lifted and the lifting equipment available at the site will dictate the type of lifting attachments to be designed for the vessel. Once the type of vessel attachment has been chosen, careful consideration must be given to the design of attachments to the pressure vessel in consideration to ensure that the vessel and lifting components are not overstressed during the lifting process. The paper also discusses different types of lifting attachments that may be attached to each end of the vessel either by bolting or welding and discusses the pros and cons of each. The paper also provides an example of a finite element analysis (FEA) of a top nozzle, a FEA of a pair of lifting trunnions and a FEA of welded on lifting lugs for buried pipe. The purpose of the paper is to outline the

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

International Nuclear Information System (INIS)

1996-01-01

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

Requirements for thermal insulation on prestressed concrete reactor vessels

International Nuclear Information System (INIS)

Neylan, A.J.; Wistrom, J.D.

1979-01-01

During the past decade, extensive design, construction, and operating experience on concrete pressure vessels for gas-cooled reactor applications has accumulated. Excellent experience has been obtained to date on the structural components (concrete, prestressing systems, liners, penetrations, and closures) and the thermal insulation. Three fundamentally different types of insulation systems have been employed to ensure the satisfactory performance of this component, which is critical to the overall success of the prestressed concrete reactor vessel (PCRV). Although general design criteria have been published, the requirements for design, materials, and construction are not rigorously addressed in any national or international code. With the more onerous design conditions being imposed by advanced reactor systems, much greater attention has been directed to advance the state of the art of insulation systems for PCRVs. This paper addresses some of the more recent developments in this field being performed by General Atomic Company and others. (author)

Design, fabrication and test of double-wall vacuum vessel for JT-60U

International Nuclear Information System (INIS)

Uchikawa, Takashi; Ioki, Kimihiro; Ninomiya, Hiromasa.

1994-01-01

A double-wall vacuum vessel was designed and fabricated for JT-60U (an upgraded machine of JT-60), which has a plasma current up to 6 MA and a large plasma volume (100 m 3 ). A new concept of Inconel 625 all-welded structure was adopted to the vessel, that comprises an inner plate, square tubes and an outer plate. The vacuum vessel with a multi-arc D-shaped cross section was fabricated by using hot-sizing press. The electromagnetic and structural analysis has been performed for plasma disruption loads. Dynamic responses of the vessel were measured during plasma disruptions, and the observed displacement had a good agreement with the result of FEM analysis. (author)

Structural design considerations in the Mirror Fusion Test Facility (MFTF-B) vacuum vessel

International Nuclear Information System (INIS)

Vepa, K.; Sterbentz, W.H.

1981-01-01

In view of favorable results from the Tandem Mirror Experiment (TMX) also at LLNL, the MFTF project is now being rescoped into a large tandem mirror configuration (MFTF-B), which is the mainline approach to a mirror fusion reactor. This paper concerns itself with the structural aspects of the design of the vessel. The vessel and its intended functions are described. The major structural design issues, especially those influenced by the analysis, are described. The objectives of the finite element analysis and their realization are discussed at length

Study on Off-Design Steady State Performances of Helium Gas Turbo-compressor for HTGR-GT

International Nuclear Information System (INIS)

Qisen Ren; Xiaoyong Yang; Zhiyong Huang; Jie Wang

2006-01-01

The high temperature gas-cooled reactor (HTGR) coupled with direct gas turbine cycle is a promising concept in the future of nuclear power development. Both helium gas turbine and compressor are key components in the cycle. Under normal conditions, the mode of power adjustment is to control total helium mass in the primary loop using gas storage vessels. Meanwhile, thermal power of reactor core is regulated. This article analyzes off-design performances of helium gas turbine and compressors for high temperature gas-cooled reactor with gas turbine cycle (HTGR-GT) at steady state level of electric power adjustment. Moreover, performances of the cycle were simply discussed. Results show that the expansion ratio of turbine decreases as electric power reduces but the compression ratios of compressors increase, efficiencies of both turbine and compressors decrease to some extent. Thermal power does not vary consistently with electric power, the difference between these two powers increases as electric power reduces. As a result of much thermal energy dissipated in the temperature modulator set at core inlet, thermal efficiency of the cycle has a widely reduction under partial load conditions. (authors)

Design of the Intersector Welding Robot for vacuum vessel assembly and maintenance

International Nuclear Information System (INIS)

Jones, L.; Dagenais, J.-F.; Daenner, W.; Maisonnier, D.

2000-01-01

Next Step Fusion Devices require on-site (field weld) joining of sectors of the thick-walled vacuum vessel for structural and vacuum integrity. EFDA (European Fusion Development Agreement) is supporting an R and D programme to investigate processes for assembly of the vacuum vessel and to carry out cutting, re-welding and inspection for remote sector replacement, forming part of the overall VV/blanket research effort. In order to direct the process end-effectors along the field joint zone, a track-mounted Intersector Welding Robot (IWR) on a mock-up of a region of the vacuum vessel has been designed and is described in this paper. A rail-mounted hexapod type robot offers six axes of motion over a limited work envelope with high payload to robot weight ratio. A solution to the production of reduced pressure local vacuum is the installation of short, lightweight segments bolted to each other and the vessel wall. The various process heads can be mounted using end-effectors of special design. To minimise the supply and interface problems for the IWR prototype, its motion control and electronic systems will be embedded locally. A laser scan with camera forms the on-line seam tracking capability to compensate for rail and seam deviations

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.

Ex-vessel remote maintenance design for the Compact Ignition Tokamak

International Nuclear Information System (INIS)

Spampinato, P.T.; Macdonald, D.

1987-01-01

The use of deuterium-tritium (D-T) fuel for operation of the Compact Ignition Tokamak (CIT) imposes a requirement for remote handling technology for ex-vessel maintenance operations on auxiliary machine components. These operations consist of repairing and replacing components such as diagnostic, radio-frequency (rf) heating, and fueling systems using remotely operated maintenance equipment in the test cell. In addition, ex-vessel maintenance design also includes developing hot cell facilities for equipment decontamination, repair, and solid radioactive waste handling. The test cell maintenance philosophy is markedly influenced by the neutron/gamma shield surrounding the machine that allows personal access into the test cell one day after shutdown. Hence, maintenance operations can be performed hands-on in the test cell with the shield intact and must be remotely performed when the shield is disassembled for machine access. The constricted access to the auxiliary components of the machine affect the design requirements for the maintenance equipment and impose major spatial constraints. Several major areas of the maintenance system design are being addressed in fiscal year 1987. These include conceptual design of the manipulator system, preliminary remote equipment research and development, and definition of the hot cell, decontamination, and equipment repair facility requirements. The manipulator work includes investigating transporters and viewing/lighting subsystems. 2 figs

Development of ITER in-vessel viewing and metrology systems

Energy Technology Data Exchange (ETDEWEB)

Obara, Kenjiro; Kakudate, Satoshi; Nakahira, Masataka; Ito, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-04-01

The ITER in-vessel viewing system is vital for detecting and locating damage to in-vessel components such as the blankets and divertors and in monitoring and assisting in-vessel maintenance. This system must be able to operate at high temperature (200degC) under intense gamma radiation ({approx}30 kGy/h) in a high vacuum or 1 bar inert gas. A periscope viewing system was chosen as a reference due to its clear, wide view and a fiberscope viewing system chosen as a backup for viewing in narrow confines. According to the ITER R and D program, both systems and a metrology system are being developed through the joint efforts of Japan, the U.S., and RF Home Teams. This paper outlines design and technology development mainly on periscope in-vessel viewing and laser metrology contributed by the Japan Home Team. (author)

Development of ITER in-vessel viewing and metrology systems

International Nuclear Information System (INIS)

Obara, Kenjiro; Kakudate, Satoshi; Nakahira, Masataka; Ito, Akira

1998-01-01

The ITER in-vessel viewing system is vital for detecting and locating damage to in-vessel components such as the blankets and divertors and in monitoring and assisting in-vessel maintenance. This system must be able to operate at high temperature (200degC) under intense gamma radiation (∼30 kGy/h) in a high vacuum or 1 bar inert gas. A periscope viewing system was chosen as a reference due to its clear, wide view and a fiberscope viewing system chosen as a backup for viewing in narrow confines. According to the ITER R and D program, both systems and a metrology system are being developed through the joint efforts of Japan, the U.S., and RF Home Teams. This paper outlines design and technology development mainly on periscope in-vessel viewing and laser metrology contributed by the Japan Home Team. (author)

Design criteria and pressure vessel codes - an American view

International Nuclear Information System (INIS)

Tuppeny, W.H.

1975-01-01

To the pressure vessel designer, codes and criteria represent the common ground where the stress analyst and the metallurgist must interact and evolve rules and procedures which will ensure safety and open-ended responsiveness to technological, economic, and environmental change. The paper briefly discusses the evolution and rationale behind the current ASME code sections -emphasizing those portions applicable to designs operating in the creep range. The author then proposes a plan of action so that the analysts and materials people can make optimum use of time and resources, and evolve data and design criteria which will be responsive to changing technology and the economic and safety requirements of the future. (author)

Cooling system for the connecting rings of a fast neutron reactor vessel

International Nuclear Information System (INIS)

Martin, J.-P.; Malaval, Claude

1974-01-01

A description is given of a cooling system for the vessel connecting rings of a fast neutron nuclear reactor, particularly of a main vessel containing the core of the reactor and a volume of liquid metal coolant at high temperature and a safety vessel around the main vessel, both vessels being suspended to a rigid upper slab kept at a lower temperature. It is mounted in the annular space between the two vessels and includes a neutral gas circuit set up between the wall of the main vessel to be cooled and that of the safety vessel itself cooled from outer. The neutral gas system comprises a plurality of ventilators fitted in holes made through the thickness of the upper slab and opening on to the space between the two vessels. It also includes two envelopes lining the walls of these vessels, establishing with them small section channels for the circulation of the neutral gas cooled against the safety vessel and heated against the main vessel [fr

Materials considerations for UF6 gas-core reactor. Interim report for preliminary design study

International Nuclear Information System (INIS)

Wagner, P.

1977-04-01

The limiting materials problem in a high-temperature UF 6 core reactor is the corrosion of the core containment vessel. The UF 6 , the lower fluorides of uranium, and the fluorine that exist at the anticipated reactor operating conditions (1000 K and about one atmosphere UF 6 ) are all corrosive. Because of this, the materials evaluation effort for this reactor design study has concentrated on the identification of a viable system for the containment vessel that meets both the materials and neutronic requirements. A study of the literature has revealed that the most promising corrosion-resistant candidates are Ni or Ni-Al alloys. One of the conclusions of this work is that the containment vessel use a nickel liner or clad since the use of Ni as a structural member is precluded by its relative blackness to thermal neutrons. Estimates of corrosion rates of Ni and Ni-Al alloys, the effects of the pressure and temperature of F 2 on the corrosion rates, calculated equilibrium gas compositions at reactor core operating conditions, suggested methods of fabrication, and recommendations for future research and development are included

Design and development of the ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Koizumi, K.; Nakahira, M.; Itou, Y.; Tada, E. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan); Johnson, G.; Ioki, K.; Elio, F.; Iizuka, T.; Sannazzaro, G.; Takahashi, K.; Utin, Y.; Onozuka, M. [ITER Joint Central Team (JCT), Garching (Germany); Nelson, B. [US Home Team, Oak Ridge National Laboratory (United States); Vallone, C. [EU Home Team, NET Team, Garching (Germany); Kuzmin, E. [RF Home Team, Efremov Institute, City (Russian Federation)

1998-09-01

In ITER, the vacuum vessel (VV) is designed to be a water cooled, double-walled toroidal structure made of 316LN stainless steel with a D-shaped cross section approximately 9 m wide and 15 m high. The design work which began at the beginning of the ITER-EDA is nearing completion by resolving the technical issues. In parallel with the design activities, the R and D program, full-scale VV sector model project, was initiated in 1995 to resolve the design and fabrication issues. The full-scale sector model corresponds to an 18 sector (9 sub-sector x 2) and is being fabricated on schedule. To date, 60% of the fabrication had been completed. The fabrication of full-scale model including sector-to-sector connection will be completed by the end of 1997 and performance tests are scheduled until the end of ITER-EDA. This paper describes the latest status of the ITER VV design and the full-scale sector model project. (orig.) 3 refs.

Streamlined vessels for speedboats: Macro modifications of shark skin design applications

Science.gov (United States)

Ibrahim, M. D.; Amran, S. N. A.; Zulkharnain, A.; Sunami, Y.

2018-01-01

Functional properties of shark denticles have caught the attention of engineers and scientist today due to the hydrodynamic effects of its skin surface roughness. The skin of a fast swimming shark reveals riblet structures that help to reduce skin friction drag, shear stresses, making its movement to be more efficient and faster. Inspired by the structure of the shark skin denticles, our team has conducted a study on alternative on improving the hydrodynamic design of marine vessels by applying the simplified version of shark skin skin denticles on the surface hull of the vessels. Models used for this study are constructed and computational fluid dynamic (CFD) simulations are then carried out to predict the effectiveness of the hydrodynamic effects of the biomimetic shark skins on those models. Interestingly, the numerical calculated results obtained shows that the presence of biomimetic shark skin implemented on the vessels give improvements in the maximum speed as well as reducing the drag force experience by the vessels. The pattern of the wave generated post cruising area behind the vessels can also be observed to reduce the wakes and eddies. Theoretically, reduction of drag force provides a more efficient vessel with a better cruising speed. To further improve on this study, the authors are now actively arranging an experimental procedure in order to verify the numerical results obtained by CFD. The experimental test will be carried out using an 8 metre flow channel provided by University Malaysia Sarawak, Malaysia.

Reactor pressure vessel embrittlement of NPP borssele: Design lifetime and lifetime extension

International Nuclear Information System (INIS)

Blom, F.J.

2007-01-01

Embrittlement of the reactor pressure vessel of the Borssele nuclear power plant has been investigated taking account of the design lifetime of 40 years and considering 20 years subsequent lifetime extension. The paper presents the current licensing status based on considerations of material test data and of US nuclear regulatory standards. Embrittlement status is also evaluated against German and French nuclear safety standards. Results from previous fracture toughness and Charpy tests are investigated by means of the Master curve toughness transition approach. Finally, state of the art insights are investigated by means of literature research. Regarding the embrittlement status of the reactor pressure vessel of Borssele nuclear power plant it is concluded that there is a profound basis for the current license up to the original end of the design life in 2013. The embrittlement temperature changes only slightly with respect to the acceptance criterion adopted postulating further operation up to 2033. Continued safe operation and further lifetime extension are therefore not restricted by reactor pressure vessel embrittlement

Standard High Solids Vessel Design De-inventory Simulant Qualification

Energy Technology Data Exchange (ETDEWEB)

Fiskum, Sandra K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Burns, Carolyn A.M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gauglitz, Phillip A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Linn, Diana T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Peterson, Reid A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smoot, Margaret R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2017-09-12

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is working to develop a Standard High Solids Vessel Design (SHSVD) process vessel. To support testing of this new design, WTP engineering staff requested that a Newtonian simulant be developed that would represent the de-inventory (residual high-density tank solids cleanout) process. Its basis and target characteristics are defined in 24590-WTP-ES-ENG-16-021 and implemented through PNNL Test Plan TP-WTPSP-132 Rev. 1.0. This document describes the de-inventory Newtonian carrier fluid (DNCF) simulant composition that will satisfy the basis requirement to mimic the density (1.18 g/mL ± 0.1 g/mL) and viscosity (2.8 cP ± 0.5 cP) of 5 M NaOH at 25 °C.1 The simulant viscosity changes significantly with temperature. Therefore, various solution compositions may be required, dependent on the test stand process temperature range, to meet these requirements. Table ES.1 provides DNCF compositions at selected temperatures that will meet the density and viscosity specifications as well as the temperature range at which the solution will meet the acceptable viscosity tolerance.

Failure probability analysis on mercury target vessel

International Nuclear Information System (INIS)

Ishikura, Syuichi; Futakawa, Masatoshi; Kogawa, Hiroyuki; Sato, Hiroshi; Haga, Katsuhiro; Ikeda, Yujiro

2005-03-01

Failure probability analysis was carried out to estimate the lifetime of the mercury target which will be installed into the JSNS (Japan spallation neutron source) in J-PARC (Japan Proton Accelerator Research Complex). The lifetime was estimated as taking loading condition and materials degradation into account. Considered loads imposed on the target vessel were the static stresses due to thermal expansion and static pre-pressure on He-gas and mercury and the dynamic stresses due to the thermally shocked pressure waves generated repeatedly at 25 Hz. Materials used in target vessel will be degraded by the fatigue, neutron and proton irradiation, mercury immersion and pitting damages, etc. The imposed stresses were evaluated through static and dynamic structural analyses. The material-degradations were deduced based on published experimental data. As a result, it was quantitatively confirmed that the failure probability for the lifetime expected in the design is very much lower, 10 -11 in the safety hull, meaning that it will be hardly failed during the design lifetime. On the other hand, the beam window of mercury vessel suffered with high-pressure waves exhibits the failure probability of 12%. It was concluded, therefore, that the leaked mercury from the failed area at the beam window is adequately kept in the space between the safety hull and the mercury vessel by using mercury-leakage sensors. (author)

The effect of diffusivity on gas-liquid mass transfer in stirred vessels. Experiments at atmospheric and elevated pressures

NARCIS (Netherlands)

Versteeg, G.F.; Blauwhoff, P.M.M.; Swaaij, W.P.M. van

1987-01-01

Mass transfer has been studied in gas-liquid stirred vessels with horizontal interfaces which appeared to the eye to be completely smooth. Special attention has been paid to the influence of the coefficient of molecular diffusion. The results are compared with those published before. The simplifying

Leak detection device for nuclear reactor pressure vessel

International Nuclear Information System (INIS)

Ikeda, Jun.

1988-01-01

Purpose: To test the leakage of a nuclear reactor pressure vessel during stopping for a short period of time with no change to the pressure vessel itself. Constitution: The device of the present invention comprises two O-rings disposed on the flange surface that connects a pressure vessel main body and an upper cover, a leak-off pipeway derived from the gap of the O-rings at the flange surface to the outside of the pressure vessel, a pressure detection means connected to the end of the pipeway, a humidity detection means disposed to the lead-off pipeway, a humidity detection means disposed to the lead-off pipeway, and gas supply means and gas suction means disposed each by way of a check valve to a side pipe branched from the pipeway. After stopping the operation of the nuclear reactor and pressurizing the pressure vessel by filling water, gases supplied to the gap between the O-rings at the flange surface by opening the check valve. In a case where water in the pressure vessel should leak to the flange surface, when gas suction is applied by properly opening the check valve, increase in the humidity due to the steams of leaked water diffused into the gas is detected to recognize the occurrence of leakage. (Kamimura, M.)

Impact of chemistry on Standard High Solids Vessel Design mixing

Energy Technology Data Exchange (ETDEWEB)

Poirier, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-03-02

The plan for resolving technical issues regarding mixing performance within vessels of the Hanford Waste Treatment Plant Pretreatment Facility directs a chemical impact study to be performed. The vessels involved are those that will process higher (e.g., 5 wt % or more) concentrations of solids. The mixing equipment design for these vessels includes both pulse jet mixers (PJM) and air spargers. This study assesses the impact of feed chemistry on the effectiveness of PJM mixing in the Standard High Solids Vessel Design (SHSVD). The overall purpose of this study is to complement the Properties that Matter document in helping to establish an acceptable physical simulant for full-scale testing. The specific objectives for this study are (1) to identify the relevant properties and behavior of the in-process tank waste that control the performance of the system being tested, (2) to assess the solubility limits of key components that are likely to precipitate or crystallize due to PJM and sparger interaction with the waste feeds, (3) to evaluate the impact of waste chemistry on rheology and agglomeration, (4) to assess the impact of temperature on rheology and agglomeration, (5) to assess the impact of organic compounds on PJM mixing, and (6) to provide the technical basis for using a physical-rheological simulant rather than a physical-rheological-chemical simulant for full-scale vessel testing. Among the conclusions reached are the following: The primary impact of precipitation or crystallization of salts due to interactions between PJMs or spargers and waste feeds is to increase the insoluble solids concentration in the slurries, which will increase the slurry yield stress. Slurry yield stress is a function of pH, ionic strength, insoluble solids concentration, and particle size. Ionic strength and chemical composition can affect particle size. Changes in temperature can affect SHSVD mixing through its effect on properties such as viscosity, yield stress, solubility

Standard practice for examination of seamless, Gas-Filled, pressure vessels using acoustic emission

CERN Document Server

American Society for Testing and Materials. Philadelphia

2009-01-01

1.1 This practice provides guidelines for acoustic emission (AE) examinations of seamless pressure vessels (tubes) of the type used for distribution or storage of industrial gases. 1.2 This practice requires pressurization to a level greater than normal use. Pressurization medium may be gas or liquid. 1.3 This practice does not apply to vessels in cryogenic service. 1.4 The AE measurements are used to detect and locate emission sources. Other nondestructive test (NDT) methods must be used to evaluate the significance of AE sources. Procedures for other NDT techniques are beyond the scope of this practice. See Note 1. Note 1—Shear wave, angle beam ultrasonic examination is commonly used to establish circumferential position and dimensions of flaws that produce AE. Time of Flight Diffraction (TOFD), ultrasonic examination is also commonly used for flaw sizing. 1.5 The values stated in inch-pound units are to be regarded as the standard. The values given in parentheses are for information only. 1.6 This standa...

75 FR 54527 - Defense Federal Acquisition Regulation Supplement; Government Rights in the Design of DoD Vessels...

Science.gov (United States)

2010-09-08

...-AG50 Defense Federal Acquisition Regulation Supplement; Government Rights in the Design of DoD Vessels.... Section 825 clarifies the Government's rights in technical data in the designs of a DoD vessel, boat... cite DFARS Case 2008-D039. SUPPLEMENTARY INFORMATION: A. Background This final rule implements section...

Design and material selection for ITER first wall/blanket, divertor and vacuum vessel

Science.gov (United States)

Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Gohar, Y.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Lousteau, D.; Onozuka, M.; Parker, R.; Sannazzaro, G.; Tivey, R.

1998-10-01

Design and R&D have progressed on the ITER vacuum vessel, shielding and breeding blankets, and the divertor. The principal materials have been selected and the fabrication methods selected for most of the components based on design and R&D results. The resulting design changes are discussed for each system.

Landfill gas management facilities design guidelines

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-03-15

In British Columbia, municipal solid waste landfills generate over 1000 tonnes of methane per year; landfill gas management facilities are required to improve the environmental performance of solid waste landfills. The aim of this document, developed by the British Columbia Ministry of the Environment, is to provide guidance for the design, installation, and operation of landfill gas management facilities to address odor and pollutant emissions issues and also address health and safety issues. A review of technical experience and best practices in landfill gas management facilities was carried out, as was as a review of existing regulations related to landfill gas management all over the world. This paper provides useful information to landfill owners, operators, and other professionals for the design of landfill gas management facilities which meet the requirements of landfill gas management regulations.

Design and material selection for ITER first wall/blanket, divertor and vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Gohar, Y.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Lousteau, D.; Onozuka, M.; Parker, R.; Sannazzaro, G.; Tivey, R. [ITER JCT, Garching (Germany)

1998-10-01

Design and R and D have progressed on the ITER vacuum vessel, shielding and breeding blankets, and the divertor. The principal materials have been selected and the fabrication methods selected for most of the components based on design and R and D results. The resulting design changes are discussed for each system. (orig.) 11 refs.

Evaluation of the integrity of reactor vessels designed to ASME Code, Sections I and/or VIII

International Nuclear Information System (INIS)

Hoge, K.G.

1976-01-01

A documented review of nuclear reactor pressure vessels designed to ASME Code, Sections I and/or VIII is made. The review is primarily concerned with the design specifications and quality assurance programs utilized for the reactor vessel construction and the status of power plant material surveillance programs, pressure-temperature operating limits, and inservice inspection programs. The following ten reactor vessels for light-water power reactors are covered in the report: Indian Point Unit No. 1, Dresden Unit No. 1, Yankee Rowe, Humboldt Bay Unit No. 3, Big Rock Point, San Onofre Unit No. 1, Connecticut Yankee, Oyster Creek, Nine Mile Point Unit No. 1, and La Crosse

Exhaust gas processing facility

International Nuclear Information System (INIS)

Terada, Shin-ichi.

1995-01-01

The facility of the present invention comprises a radioactive liquid storage vessel, an exhaust gas dehumidifying device for dehumidifying gases exhausted from the vessel and an exhaust gas processing device for reducing radioactive materials in the exhaust gases. A purified gas line is disposed to the radioactive liquid storage vessel for purging exhaust gases generated from the radioactive liquid, then dehumidified and condensed liquid is recovered, and exhaust gases are discharged through an exhaust gas pipe disposed downstream of the exhaust gas processing device. With such procedures, the scale of the exhaust gas processing facility can be reduced and exhaust gases can be processed efficiently. (T.M.)

Multipurpose service vessels. Versatile toolkits for well intervention

Energy Technology Data Exchange (ETDEWEB)

Adamson, S [Dowell, Las Morochas (Venezuela); Cupello, F; Hicks, J; Keenleyside, M [Sedco Forex, Las Morochas (Venezuela); Formas, D; Gabillard, C [Sedco Forex, Montrouge (France); Gamarra, F; Sanchez, A [Lagoven SA, Tia Juana (Venezuela)

1997-12-31

The industry has entered a new area in offshore support operations. Today, novel concepts and designs offer an expanded range of capabilities from a single vessel rather than the multiple boats and barges that have been used in the past. This continuing evolution in marine services is rapidly transforming well workover and intervention activities, and solving logistics and performance problems that have challenged oil and gas operators for decades. 13 figs., 5 refs.

Multipurpose service vessels. Versatile toolkits for well intervention

Energy Technology Data Exchange (ETDEWEB)

Adamson, S. [Dowell, Las Morochas (Venezuela); Cupello, F.; Hicks, J.; Keenleyside, M. [Sedco Forex, Las Morochas (Venezuela); Formas, D.; Gabillard, C. [Sedco Forex, Montrouge (France); Gamarra, F.; Sanchez, A. [Lagoven SA, Tia Juana (Venezuela)

1996-12-31

The industry has entered a new area in offshore support operations. Today, novel concepts and designs offer an expanded range of capabilities from a single vessel rather than the multiple boats and barges that have been used in the past. This continuing evolution in marine services is rapidly transforming well workover and intervention activities, and solving logistics and performance problems that have challenged oil and gas operators for decades. 13 figs., 5 refs.

Design and thermal/hydraulic characteristics of the ITER-FEAT vacuum vessel

International Nuclear Information System (INIS)

Onozuka, M.; Ioki, K.; Sannazzaro, G.; Utin, Y.; Yoshimura, H.

2001-01-01

Recent progress in structural design and thermal and hydraulic assessment of the vacuum vessel (VV) for ITER-FEAT is presented. Because of the direct attachment of the blanket modules to the VV, the module support structures are recessed into the double-wall VV, partially replacing the stiffening ribs between the VV shells to simplify the VV structure. Structural integrity of the VV is provided by the ribs and the module support structures with local reinforcement ribs. The detailed structural design of the VV taking account of the fabricability and code/standard acceptance is presented. Cost reduction of the VV fabrication using casting or forging is proposed. A high heat removal capability is required for the VV cooling to keep the thermal stress below the allowable. It is expected that natural thermo-gravitational convection due to the heat flux from the vessel wall to the water will enhance heat transfer characteristics even in the low flow velocity region

Design and thermal/hydraulic characteristics of the ITER-FEAT vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Onozuka, M. E-mail: onozukm@itereu.de; Ioki, K.; Sannazzaro, G.; Utin, Y.; Yoshimura, H

2001-11-01

Recent progress in structural design and thermal and hydraulic assessment of the vacuum vessel (VV) for ITER-FEAT is presented. Because of the direct attachment of the blanket modules to the VV, the module support structures are recessed into the double-wall VV, partially replacing the stiffening ribs between the VV shells to simplify the VV structure. Structural integrity of the VV is provided by the ribs and the module support structures with local reinforcement ribs. The detailed structural design of the VV taking account of the fabricability and code/standard acceptance is presented. Cost reduction of the VV fabrication using casting or forging is proposed. A high heat removal capability is required for the VV cooling to keep the thermal stress below the allowable. It is expected that natural thermo-gravitational convection due to the heat flux from the vessel wall to the water will enhance heat transfer characteristics even in the low flow velocity region.

Process, background and design criteria of the gas cleaning at Puertollano IGCC

Energy Technology Data Exchange (ETDEWEB)

Pisa, J. [Elcogas, Madrid (Spain)

1998-11-01

The Puertollano IGCC plant selected cooling by a water-tube boiler with upstream quenching at high velocities that requires a dust-free cooling gas at not less than 250{degree}C in order not to penalise the heat recovery efficiency. A filtration system for gas dedusting in the 250{degree}C temperature range has been installed and will be commissioned at the end of 1997. The gas cleaning concept is completed with a Venturi Scrubber, a COS hydrolysis reactor and a MDEA column to strip the sulphuric acid to yield clean gas. The gasification island is based upon the PRENFLO system which is an entrained-flow system with dry feeding. The selection of the filter system arrangement considered the limited operational experience in comparable operating conditions and acknowledged the flexibility of the filter system versus the cyclone-scrubber as far as easier load variation operation, the reduction of residues needing deposition and increased slag flow, as well as easier maintenance. Additionally to the ceramic test filters in Furstenhausen (PRENFLO) and Deer Park near Houston (SHELL), ceramic candle-type filter were selected in Buggenum and at Wabash River, and for the KoBra plant. The main criteria for the selection of the filter system and the type of candle were: separation efficiency to match clean gas limits; uniform distribution of the dust-laden gas to the filters; wear-resistant routing of the dust-laden gas flow; need for a supporting structure which must cope with sudden pressure fluctuations; optimised pulse gas system; and maintenance and repair. Based upon the above criteria, the PRENFLO concept requirements and the gas turbine specification, an arrangement with two pressure filter vessels with LLB design and filter elements manufactured by Schumacher has been installed in Puertollano. 2 figs., 3 tabs.

Background gas density and beam losses in NIO1 beam source

Energy Technology Data Exchange (ETDEWEB)

Sartori, E., E-mail: emanuele.sartori@igi.cnr.it; Veltri, P.; Serianni, G. [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), C.so Stati Uniti 4, 35127 Padova (Italy); Cavenago, M. [INFN-LNL, v.le dell’Università 2, I-35020 Legnaro (PD) (Italy)

2016-02-15

NIO1 (Negative Ion Optimization 1) is a versatile ion source designed to study the physics of production and acceleration of H- beams up to 60 keV. In ion sources, the gas is steadily injected in the plasma source to sustain the discharge, while high vacuum is maintained by a dedicated pumping system located in the vessel. In this paper, the three dimensional gas flow in NIO1 is studied in the molecular flow regime by the Avocado code. The analysis of the gas density profile along the accelerator considers the influence of effective gas temperature in the source, of the gas temperature accommodation by collisions at walls, and of the gas particle mass. The calculated source and vessel pressures are compared with experimental measurements in NIO1 during steady gas injection.

Design criteria for the structural analysis of shipping cask containment vessels

International Nuclear Information System (INIS)

Anon.

1979-01-01

10 CFR Part 71, Sections 71.35 and 71.36, require that packages used to transport radioactive materials meet specified normal and hypothetical accident conditions. Acceptable design criteria are presented for use in the structural analysis of the containment vessels of Type B packages used to transport irradiated nuclear fuel. Alternative design criteria meeting the structural requirements of 10 CFR Part 71, Section 71.35 and 71.36, may also be used

Guide to the periodic inspection of nuclear reactor steel pressure vessels

International Nuclear Information System (INIS)

1969-01-01

This Guide is intended to provide general information and guidance to reactor owners or operators, inspection authorities, certifying authorities or regulatory bodies who are responsible for establishing inspection procedures for specific reactors or reactor types, and for the preparation of national codes or standards. The recommendations of the Guide apply primarily to water-cooled steel reactor vessels which are at a sufficiently early stage of design so that recommendations to provide accessibility for inspection can be incorporated into the early stages of design and inspection planning. However, much of the contents of the Guide are also applicable in part to vessels for other reactor types, such as gas-cooled, pressure-tube, or liquid-metal-cooled reactors, and also to some existing water-cooled reactors and reactors which are in advanced stage of design or construction. 46 refs, figs, 1 tab

Integration of ITER in-vessel diagnostic components in the vacuum vessel

International Nuclear Information System (INIS)

Encheva, A.; Bertalot, L.; Macklin, B.; Vayakis, G.; Walker, C.

2009-01-01

The integration of ITER in-vessel diagnostic components is an important engineering activity. The positioning of the diagnostic components must correlate not only with their functional specifications but also with the design of the major parts of ITER torus, in particular the vacuum vessel, blanket modules, blanket manifolds, divertor, and port plugs, some of which are not yet finally designed. Moreover, the recently introduced Edge Localised Mode (ELM)/Vertical Stability (VS) coils mounted on the vacuum vessel inner wall call for not only more than a simple review of the engineering design settled down for several years now, but also for a change in the in-vessel distribution of the diagnostic components and their full impact has yet to be determined. Meanwhile, the procurement arrangement (a document defining roles and responsibilities of ITER Organization and Domestic Agency(s) (DAs) for each in-kind procurement including technical scope of work, quality assurance requirements, schedule, administrative matters) for the vacuum vessel must be finalized. These make the interface process even more challenging in terms of meeting the vacuum vessel (VV) procurement arrangement's deadline. The process of planning the installation of all the ITER diagnostics and integrating their installation into the ITER Integrated Project Schedule (IPS) is now underway. This paper covers the progress made recently on updating and issuing the interfaces of the in-vessel diagnostic components with the vacuum vessel, outlines the requirements for their attachment and summarises the installation sequence.

Pressure vessel rupture within a chamber: the pressure history on the chamber wall

International Nuclear Information System (INIS)

Baum, M.R.

1989-04-01

Generally there is a large number of pressure vessels containing high pressure gas on power stations and chemical plant. In many instances, particularly on power plant, these vessels are within the main building. If a pressure vessel were to fail, the surrounding structures would be exposed to blast loads and the forces resulting from jets of fluid issuing from the breached vessel. In the case where the vessel is in a relatively closed chamber there would also be a general overpressurisation of the chamber. At the design stage it is therefore essential to demonstrate that the plant could be safely shut down in the event of a pressure vessel failure, that is, it must be shown that the chamber will not collapse thus putting the building at risk or hazarding equipment essential for a safe shut down. Such an assessment requires the loads applied to the chamber walls, roof, etc. to be known. (author)

Design and R and D for the ITER vacuum vessel

International Nuclear Information System (INIS)

Ioki, K.; Johnson, G.; Onozuka, M.; Sannazzaro, G.; Utin, Y.; Iizuka, T.; Parker, R.; Koizumi, K.; Kuzmin, E.; Maisonnier, D.; Nelson, B.

1998-01-01

The current design and key R and D results for the Vacuum Vessel (VV) for the International Thermonuclear Experimental Reactor (ITER) are presented. During the past two years the basic VV design has remained unchanged. Additional details have been defined in key areas and recent R and D results have indicated where further improvements can be made. R and D results have also confirmed the feasibility of important aspects of the design such as limiting weld distortions to acceptable levels and achieving required tolerances with a large welded structure. Recent design progress includes the development of a structural design strategy for the VV, modification of the inboard structure, employment of ferromagnetic material between the VV shells, and confirmation of the cooling characteristics for the VV. This report presents the current design and how it has been affected by R and D results. (authors)

Design and R and D for the ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Ioki, K.; Johnson, G.; Onozuka, M.; Sannazzaro, G.; Utin, Y.; Iizuka, T.; Parker, R. [ITER Joint Work Site, Garching (Germany); Koizumi, K. [Japan Atomic Energy Research Inst., Naka (Japan); Kuzmin, E. [Efremov Insitute, Saint Petersburg (Russian Federation); Maisonnier, D. [NET Team, Garching (Germany); Nelson, B. [Oak Ridge National Lab., TN (United States)

1998-07-01

The current design and key R and D results for the Vacuum Vessel (VV) for the International Thermonuclear Experimental Reactor (ITER) are presented. During the past two years the basic VV design has remained unchanged. Additional details have been defined in key areas and recent R and D results have indicated where further improvements can be made. R and D results have also confirmed the feasibility of important aspects of the design such as limiting weld distortions to acceptable levels and achieving required tolerances with a large welded structure. Recent design progress includes the development of a structural design strategy for the VV, modification of the inboard structure, employment of ferromagnetic material between the VV shells, and confirmation of the cooling characteristics for the VV. This report presents the current design and how it has been affected by R and D results. (authors)

Containment Performance Evaluation of a Sodium Fire Event Due to Air Ingress into the Cover Gas Region of the Reactor Vessel in the PGSFR

International Nuclear Information System (INIS)

Ahn, Sang June; Chang, Won-Pyo; Kang, Seok Hun; Choi, Chi-Woong; Yoo, Jin; Lee, Kwi Lim; Jeong, Jae-Ho; Lee, Seung Won; Jeong, Taekyeong; Ha, Kwi-Seok

2015-01-01

Comparing with the light water reactor, sodium as a reactor coolant violently reacts with oxygen in the containment atmosphere. Due to this chemical reaction, heat generated from the combustion heat increases the temperature and pressure in the containment atmosphere. The structural integrity of the containment building which is a final radiological defense barrier is threaten. A sodium fire event in the containment due to air ingress into the cover gas region in the reactor vessel is classified as one of the design basis events in the PGSFR. This event comes from a leak or crack on the reactor upper closure header surface. It accompanys an event of the radiological fission products release to the inside the containment. In this paper, evaluation for the sodium fire and radiological influence due to air ingress into the cover gas region of the reactor vessel is described. To evaluate this event, the CONTAIN-LMR, MACCS-II and OR-IGEN-II codes are used. For the sodium pool fire event in the containment, the performance evaluation and radiological influence are carried out. In the thermal hydraulic aspects, the 1 cell containment yields the most conservative result. In this event, the maximum temperature and pressure in the containment are calculated 0.185 MPa, 280.0 .deg. C, respectively. The radiological dose at the EAB and LPZ are below the acceptance criteria specified in the 10CFR100

Containment Performance Evaluation of a Sodium Fire Event Due to Air Ingress into the Cover Gas Region of the Reactor Vessel in the PGSFR

Energy Technology Data Exchange (ETDEWEB)

Ahn, Sang June; Chang, Won-Pyo; Kang, Seok Hun; Choi, Chi-Woong; Yoo, Jin; Lee, Kwi Lim; Jeong, Jae-Ho; Lee, Seung Won; Jeong, Taekyeong; Ha, Kwi-Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

Comparing with the light water reactor, sodium as a reactor coolant violently reacts with oxygen in the containment atmosphere. Due to this chemical reaction, heat generated from the combustion heat increases the temperature and pressure in the containment atmosphere. The structural integrity of the containment building which is a final radiological defense barrier is threaten. A sodium fire event in the containment due to air ingress into the cover gas region in the reactor vessel is classified as one of the design basis events in the PGSFR. This event comes from a leak or crack on the reactor upper closure header surface. It accompanys an event of the radiological fission products release to the inside the containment. In this paper, evaluation for the sodium fire and radiological influence due to air ingress into the cover gas region of the reactor vessel is described. To evaluate this event, the CONTAIN-LMR, MACCS-II and OR-IGEN-II codes are used. For the sodium pool fire event in the containment, the performance evaluation and radiological influence are carried out. In the thermal hydraulic aspects, the 1 cell containment yields the most conservative result. In this event, the maximum temperature and pressure in the containment are calculated 0.185 MPa, 280.0 .deg. C, respectively. The radiological dose at the EAB and LPZ are below the acceptance criteria specified in the 10CFR100.

Gas tagging system development in Japan

International Nuclear Information System (INIS)

Sekiguchi, N.; Rindo, H.; Akiyama, T.; Miyazawa, T.; Heki, H.

1981-05-01

The Gas tagging method has been considered to be most desirable for a failed fuel location system for the fast breeder reactor, regarding the component reduction in the reactor vessel and rapid location during reactor operation. The gas tagging system has been designed by referring to R and D results obtained in Japan and other countries. The designed system is comprised of tag gas filling pins, cover gas sampling system, tag gas recovery and enrichment system, tag gas analyzer and system control and data handling computers. The main specifications for this system have been decided as follows; 1) Main function is location of failed fuels in core and a part of blanket region, 2) Identification capability is each subassembly, 3) Time for identification is within a few days, 4) Continuous operation with automatic start at fuel failure, 5) Detection sensitivity must cover both gas leak and pin burst. In designing the gas tagging system, the following R and D items were selected; 1) System design study, 2) Tag gas capsule development, 3) Modeling the tag gas behavior in reactor primary cooling system, 4) Tag gas recovery and enrichment system, 5) Computer code development for tag gas isotope ratio change estimation. Details of the Japanese gas tagging system development appear in this paper. (author)

Comparison Between Conventional Design and Cathode Gas Recirculation Design of a Direct-Syngas Solid Oxide Fuel Cell–Gas Turbine Hybrid Systems Part I: Design Performance

Directory of Open Access Journals (Sweden)

Vahid Azami

2017-06-01

Keywords: Solid oxide fuel cell, Gas turbine, Cathode gas recirculation, Exergy. Article History: Received Feb 23rd 2017; Received in revised form May 26th 2017; Accepted June 1st 2017; Available online How to Cite This Article: Azami, V, and Yari, M. (2017 Comparison between conventional design and cathode gas recirculation design of a direct-syngas solid oxide fuel cell–gas turbine hybrid systems part I: Design performance. International Journal of Renewable Energy Develeopment, 6(2, 127-136. https://doi.org/10.14710/ijred.6.2.127-136

Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 3. Numerical investigation for thermal stratification phenomena in the upper plenum

International Nuclear Information System (INIS)

Muramatsu, Toshiharu; Yamaguchi, Akira

2002-06-01

A large-scale sodium-cooled fast breeder reactor in the feasibility studies on commercialized fast reactors has a feature of consideration of thorough simplified and compacted systems and components design to realize drastic economical improvements. Therefore, special attentions should be paid to thermohydraulic designs for gas entrainment behavior from free surface, flow-induced vibration of in-vessel components, thermal stratification in the plenum, thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. In-vessel thermohydraulic analyses were carried out using a multi-dimensional code AQUA to understand the thermal stratification characteristics in the upper plenum, and to investigate trade-off relations between gas entrainment and thermal stratification phenomena on in-vessel structures for the elimination of gas entrainment possibility. From the analysis, the following results were obtained. (1) Dummy plug insertion to a slit of the upper core structure is one of the effective measures to stabilize the in-vessel flow patterns and to mitigate in-vessel thermal shocks. (2) Though flow guide device such as a baffle ring attached to reactor vessel wall is an effective measure to eliminate impinging jet to dipped plate, rising characteristics of the thermal stratification interface are affected by the baffle ring devise. (3) Thermal stratification characteristics are not influenced very much by the installation of a partial inner barrel to the dipped plate, which is an effective measure to reduce the horizontal flow velocity components at free surface. (4) Labyrinth structures to the gap between the reactor vessel wall and the outer dipped plate have direct effects upon in-vessel thermal shock characteristics including thermal stratification phenomena due to the closing of flow path between the upper plenum and the free surface plenum. (author)

Mobile nuclear reactor containment vessel

International Nuclear Information System (INIS)

Thompson, R.E.; Spurrier, F.R.; Jones, A.R.

1978-01-01

A containment vessel for use in mobile nuclear reactor installations is described. The containment vessel completely surrounds the entire primary system, and is located as close to the reactor primary system components as is possible in order to minimize weight. In addition to being designed to withstand a specified internal pressure, the containment vessel is also designed to maintain integrity as a containment vessel in case of a possible collision accident

Conformable pressure vessel for high pressure gas storage

Science.gov (United States)

Simmons, Kevin L.; Johnson, Kenneth I.; Lavender, Curt A.; Newhouse, Norman L.; Yeggy, Brian C.

2016-01-12

A non-cylindrical pressure vessel storage tank is disclosed. The storage tank includes an internal structure. The internal structure is coupled to at least one wall of the storage tank. The internal structure shapes and internally supports the storage tank. The pressure vessel storage tank has a conformability of about 0.8 to about 1.0. The internal structure can be, but is not limited to, a Schwarz-P structure, an egg-crate shaped structure, or carbon fiber ligament structure.

Progress in the design and R and D of the ITER In-Vessel Viewing and Metrology System (IVVS)

Energy Technology Data Exchange (ETDEWEB)

Dubus, Gregory, E-mail: gregory.dubus@f4e.europa.eu [Fusion for Energy, c/ Josep Pla, n°2 – Torres Diagonal Litoral – Edificio B3, 08019 Barcelona (Spain); Puiu, Adrian; Bates, Philip; Damiani, Carlo [Fusion for Energy, c/ Josep Pla, n°2 – Torres Diagonal Litoral – Edificio B3, 08019 Barcelona (Spain); Reichle, Roger; Palmer, Jim [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

2014-10-15

The In-Vessel Viewing and Metrology System (IVVS) is a fundamental tool for the ITER machine operations, aiming at performing inspections as well as providing information related to the erosion of in-vessel components, which in turn is related to the amount of mobilised dust present in the Vacuum Vessel. Periodically or on request, the IVVS scanning probes will be deployed into the Vacuum Vessel in order to acquire both visual and metrological data on plasma facing components (blanket, divertor, heating/diagnostic plugs, and test blanket modules). Recent design changes made to the six IVVS port extensions implied the need for a substantial redesign of the IVVS integrated concept – including the scanning probe and its deployment system – in order to bring it to the level of maturity suitable for the Conceptual Design Review. This paper gives an overview of the concept design for IVVS as well as of the various engineering analyses and R and D activities carried out in support to this design: neutronic, seismic and electromagnetic analyses, probe actuation validation under environmental conditions.

Safety assessment of a multicavity prestressed concrete reactor vessel with hot liner

Energy Technology Data Exchange (ETDEWEB)

Lafitte, R.; Marchand, J. D. [Bonnard et Gardel, Ingenieurs-Conseil, Lausanne (Switzerland)

1981-01-15

The prestressed concrete reactor vessel of the high temperature reactor with helium turbine project differs from those realized up to this day by the important number of cavities, by the different cavity pressures and by a liner in contact with hot gas. For the cases of operating conditions, the computations can be based on an identical pressure in all the cavities. The overdimensioning of the vessel which results is not a determining factor at this stage of the project. The possible loss of leaktightness of the liner can introduce gas pressure into the walls of the vessel. The great thickness of the walls makes it impossible to withstand the resulting forces with prestressing in offering sufficient safety factor against collapse. It is thus important to design a drainage network largely dimensioned. The warm liner appears at this stage of the project too highly stressed by fatigue at the singularity points (ducts between cavities, angles). A solution is proposed which limits the variations of thermal stresses by using a steel with low coefficient of thermal expansion. The cavity closures, which are numerous and some with large dimensions are an important aspect of the vessel safety. A solution of reinforced concrete shell with independent liner is proposed.

Safety assessment of a multicavity prestressed concrete reactor vessel with hot liner

International Nuclear Information System (INIS)

Lafitte, R.; Marchand, J.D.

1981-01-01

The prestressed concrete reactor vessel of the high temperature reactor with helium turbine project differs from those realized up to this day by the important number of cavities, by the different cavity pressures and by a liner in contact with hot gas. For the cases of operating conditions, the computations can be based on an identical pressure in all the cavities. The overdimensioning of the vessel which results is not a determining factor at this stage of the project. The possible loss of leaktightness of the liner can introduce gas pressure into the walls of the vessel. The great thickness of the walls makes it impossible to withstand the resulting forces with prestressing in offering sufficient safety factor against collapse. It is thus important to design a drainage network largely dimensioned. The warm liner appears at this stage of the project too highly stressed by fatigue at the singularity points (ducts between cavities, angles). A solution is proposed which limits the variations of thermal stresses by using a steel with low coefficient of thermal expansion. The cavity closures, which are numerous and some with large dimensions are an important aspect of the vessel safety. A solution of reinforced concrete shell with independent liner is proposed

Aluminium vacuum vessel/first surface conceptual design for a commercial tokamak hybrid reactor

International Nuclear Information System (INIS)

Culbert, M.

1981-01-01

The purpose of this investigation was to develop a design concept for a commercial tokamak hybrid reactor (CTHR) vacuum vessel/first surface system which satisfies the engineering requirements for a commercial environment. An important distinction between the design constraints associated with 'pure' fusion and fusion-fission hybrid power reactors is that energy extraction from the first wall is not critical from the point of view of hybrid system economics. This allows the consideration of low temperature structural material for first wall application. The mechanical arrangement consists of a series of internally finned aluminium tube banks running poloidally around the torus. The coolant manifolds are at the top and bottom of the torus. The vessel is divided into sectors, the length of which depends on the spacing between TF coils. The tubes in each sector are welded to tube sheets which are in turn welded to semi-cylindrical manifolds which distribute the coolant uniformly to the tubes. The tubes, which are approx. equal to 2.5 cm in diameter at the manifold location, traverse the torus poloidal periphery and change from a circular cross section to a 2:1 elliptical cross section at the horizontal midplane. The arched tube is designed to be self-supporting between the manifold locations. The vacuum vessel's first surface will be plasma flamed sprayed aluminum applied to the tubular wall. (orig./GG)

Blanket and vacuum vessel design of the next tokamak. (Swimming pool type)

International Nuclear Information System (INIS)

Iida, H.; Minato, A.; Kitamura, K.

1983-01-01

The structural design study of a reactor module for a swimming pool type reactor (SPTR) was conducted. Since pool water plays the role of radiation shielding in the SPTR, the module does not have a solid shield. It consists of tritium breeding blankets, divertor collector plates and a vacuum vessel. The object of this study is to show the reactor module design which has a simple structure and a sufficient tritium breeding ratio. A large coverage of the plasma chamber surface with tritium breeding blanket is essential in order to obtain a high tritium breeding ratio. A breeding blanket is also placed behind the divertor collector plate, i.e. in the upper and lower region, as well as in the outboard and inboard regions of the module. A concept in which the first wall is an integral part of the blanket is employed to minimize the thickness of structural and cooling material brazed in front of the breeding material (Li 2 O) and to enhance the tritium breeding capability. In order to simplify the module structure the vacuum vessel and breeding blanket is also integrated in the inboard region. One of the features inherent in the swimming pool type reactor is an additional external force on the vacuum vessel, namely hydraulic pressure. A detailed structural analysis of the vacuum vessel is performed. Divertor collector plates are assemblies of co-axial tubes. They minimize the electromagnetic force on the plate induced by the plasma disruption. A thermal and structural analysis and life time estimation of the first wall and divertor collector plates are performed. (author)

Designs for remote inspection of the ALMR Reactor Vessel Auxiliary Cooling System (RVACS)

International Nuclear Information System (INIS)

Sweeney, F.J.; Carroll, D.G.; Chen, C.; Crane, C.; Dalton, R.; Taylor, J.R.; Tosunoglu, S.; Weymouth, T.

1993-01-01

One of the most important safety systems in General Electric's (GI) Advanced Liquid Metal Reactor (ALMR) is the Reactor Vessel Auxiliary Cooling System (RVACS). Because of high temperature, radiation, and restricted space conditions, GI desired methods to remotely inspect the RVACS, emissive coatings, and reactor vessel welds during normal refueling operations. The DOE/NE Robotics for Advanced Reactors program formed a team to evaluate the ALMR design for remote inspection of the RVACS. Conceptual designs for robots to perform the required inspection tasks were developed by the team. Design criteria for these remote systems included robot deployment, power supply, navigation, environmental hardening of components, tether management, communication with an operator, sensing, and failure recovery. The operation of the remote inspection concepts were tested using 3-D simulation models of the ALMR. In addition, the team performed an extensive technology review of robot components that could survive the environmental conditions in the RVACS

Prestressed concrete reactor vessels: review of design and failure criteria

International Nuclear Information System (INIS)

Endebrock, E.G.

1975-03-01

The design and failure criteria of prestressed concrete reactor vessels (PCRVs) are reviewed along with the analysis methods. The mechanical properties of concrete under multiaxial stresses are not adequately quantified or described to permit an accurate analysis of a PCRV. Structural analysis of PCRVs almost universally utilizes a finite element which encounters difficulties in numerical solution of the governing equations and in treatment of fractured elements. (U.S.)

Probabilistic Assessment of the Design and Safety of HSLA-100 Steel Confinement Vessels

Energy Technology Data Exchange (ETDEWEB)

R.M. Dolin

2003-03-03

This probabilistic approach for assessing the design and safety of the HSLA-100 steel confinement vessel used for a DynEx test involved the probability of failure for several scenarios, in which a fragment may penetrate the vessel. The samples involve vessel thicknesses of 1 inch, 2 inches, and 5.25 inches--the combined thicknesses of the 2 inch containment vessel and the 3.25 inch safety vessel. Two simulation approaches were used for each scenario to assess the probability of failure. The Likelihood of Occurrence method simultaneously models all likely fragment events of a test, for which the net probability of failure is the sum of all the fragment events. The Stochastic Sampling method determines the probability of a fragment perforation on the basis of a logical model and takes the overall probability that an experiment results in failure as the maximum probability for any fragment event. With margin and safety assessments taken into account, it was concluded that the one and two inch thicknesses by themselves are inadequate for containing a DynEx test. The 5.25 inch thickness was determined to be safe by the Likelihood of Occurrence method and nearly adequate by the Stochastic Sampling simulation.

Helium gas turbine conceptual design by genetic/gradient optimization

International Nuclear Information System (INIS)

Yang, Long; Yu, Suyuan

2003-01-01

Helium gas turbine is the key component of the power conversion system for direct cycle High Temperature Gas-cooled Reactors (HTGR), of which an optimal design is essential for high efficiency. Gas turbine design currently is a multidisciplinary process in which the relationships between constraints, objective functions and variables are very noisy. Due to the ever-increasing complexity of the process, it has becomes very hard for the engineering designer to foresee the consequences of changing certain parts. With classic design procedures which depend on adaptation to baseline design, this problem is usually averted by choosing a large number of design variables based on the engineer's judgment or experience in advance, then reaching a solution through iterative computation and modification. This, in fact, leads to a reduction of the degree of freedom of the design problem, and therefore to a suboptimal design. Furthermore, helium is very different in thermal properties from normal gases; it is uncertain whether the operation experiences of a normal gas turbine could be used in the conceptual design of a helium gas turbine. Therefore, it is difficult to produce an optimal design with the general method of adaptation to baseline. Since their appearance in the 1970s, Genetic algorithms (GAs) have been broadly used in many research fields due to their robustness. GAs have also been used recently in the design and optimization of turbo-machines. Researchers at the General Electronic Company (GE) developed an optimization software called Engineous, and used GAs in the basic design and optimization of turbines. The ITOP study group from Xi'an Transportation University also did some work on optimization of transonic turbine blades. However, since GAs do not have a rigorous theory base, many problems in utilities have arisen, such as premature convergence and uncertainty; the GA doesn't know how to locate the optimal design, and doesn't even know if the optimal solution

Vessel size effect on the characteristic frequency of the free surface fluctuations

International Nuclear Information System (INIS)

Nam, Ho Yun; Kim, Min Joon; Kim, Jong Man; Choi, Byoung Hae

2004-01-01

Studies of the free surface fluctuations is one of the important topics in a liquid metal nuclear reactor using sodium as the coolant that has a free surface in the upper plenum of the reactor vessel. The main reasons for the study on the free surface fluctuations can be summarized as: 1. to secure the structural integrity of a reactor vessel by considering the thermal stress on the vessel wall induced by the fluctuations of the free surface between the hot sodium and cold cover gas, 2. to prevent the cover gas entrainment at the free surface of the sodium because the entrained gas causes a change in the reactivity and also reduces the heat removal capability in the core. Some experimental studies on the free surface fluctuations have been reported. However, most of them focus on the gas entrainment phenomena and only a few works concern the basic characteristics of the free surface fluctuations. Since the thermal stress on the wall is strongly dependent on the amplitude and frequency of the free surface fluctuations, studies on the amplitudes and frequencies should receive more attention. In Nam, empirical formulae on the amplitudes and frequencies with respect to the geometric and hydraulic parameters were introduced. It is an interesting result, but the experiment was performed within the parameter range near the onset point of the fluctuations. In the real reactor condition, larger sized fluctuations may exist and the formula needs to be modified. In this study, we performed experiments on the free surface fluctuations, especially on larger sized fluctuations and made an analysis of the amplitudes and frequencies. The main focus of this paper is the effect of the vessel size on the characteristic frequencies. It is thought to be helpful for finding the scaling laws, for example, designing a scale-down experiment

Leukemic Cells "Gas Up" Leaky Bone Marrow Blood Vessels.

Science.gov (United States)

Itkin, Tomer; Rafii, Shahin

2017-09-11

In this issue of Cancer Cell, Passaro et al. demonstrate how leukemia through aberrant induction of reactive oxygen species and nitric oxide production trigger marrow vessel leakiness, instigating pro-leukemic function. Disrupted tumor blood vessels promote exhaustion of non-malignant stem and progenitor cells and may facilitate leukemia relapse following chemotherapeutic treatment. Copyright © 2017. Published by Elsevier Inc.

Assessment of alternative vessel and blanket design on ITER operation

Energy Technology Data Exchange (ETDEWEB)

Cavinato, M., E-mail: mario.cavinato@f4e.europa.e [FUSION FOR ENERGY Joint Undertaking, 08019 Barcelona (Spain); Portone, A.; Saibene, G.; Sartori, R. [FUSION FOR ENERGY Joint Undertaking, 08019 Barcelona (Spain); Albanese, R.; Ambrosino, G.; Ariola, M. [Associazione Euratom-ENEA-CREATE, DIMET, Universita degli Studi di Napoli (Italy); Artaserse, G. [Associazione Euratom-ENEA-CREATE, DIMET, Universita degli Studi di Reggio Calabria (Italy); Mattei, M. [Associazione Euratom-ENEA-CREATE, DIAM, Seconda Universita di Napoli, Via Roma 29, Aversa, CE 81031 Italy (Italy); Pironti, A. [Associazione Euratom-ENEA-CREATE, DIMET, Universita degli Studi di Napoli (Italy); Villone, F. [Associazione Euratom-ENEA-CREATE, DIMET, Universita degli Studi di Cassino (Italy)

2010-12-15

In the framework of the ITER project, an investigation has been conducted on an alternative vessel and blanket design, aimed at reducing cost and production risk. The modifications proposed have a strong impact on plasma control since they affect the main conducting structures surrounding the plasma column, providing passive stabilization but at the same time shielding the field generated by the active coils to control the plasma motion and shape. An extensive analysis was performed to assess the plasma vertical controllability and the modified requirements to the in-vessel vertical stability coils system as well as to the external Poloidal Field coils system. A similar analysis was aimed at assessing the performance of the shape control system in presence of the modified structures. The effect on plasma breakdown was also evaluated in terms of maximum initial loop voltage, quality of magnetic null and the flux loss related to the breakdown delay that was quantified under the same hypothesis employed by ITER for the baseline design. Furthermore, the modified design presents issues for the magnetic diagnostic system, related to the shielding of the probes by the eddy currents, which were analysed with a 3D model. The results of the analyses performed have some general interest in particular regarding the influence on plasma stability of 3D structures with close proximity to the plasma. The present paper aims at giving an overview of the analyses that have been carried out and a summary of the results in terms of impact of the modified design on plasma control and scenario, and in general an evaluation of the role of passive structure in plasma vertical stability and shape control.

Welding distortion control in double walled KSTAR vacuum vessel fabrication

International Nuclear Information System (INIS)

Oh, D. W.; Lee, G. T.; Kim, H. K.; Yang, H. L.; Bak, J. S.

2004-01-01

The KSTAR(Korea Superconducting Tokamak Advanced Research) vacuum vessel is designed to be a double walled structure made of 12mm thick 316LN stainless steel with a D shaped cross-section about 4 m height. Vacuum vessel was pre-fabricated in two parts, 180 degree and 157.5 degree sectors in toroidal direction to meet the transportation purpose. These two parts have to be welded on site with ±2mm allowable fabrication tolerances. 1/3 scaled mock-up model was used to estimate the welding distortion and to ensure the weld quality of vacuum vessel. Gas Tungsten Arc Welding(GTAW), which has been approved by procedure qualification test, was used during mock-up test and vacuum vessel site fabrication. Welding distortion could be managed by allowing for distortion in opposite direction, by applying high restraint using lots of strong backs, by controlling the welding heat input with symmetrical welding sequence. The integrity of the site welding joint was assured by radiographic test, ultrasonic test and leak test with helium detecting method

Design for an MHD power plant as a prime mover for a Naval Vessel

International Nuclear Information System (INIS)

Paluszek, M.A.

1981-01-01

A Magnetohydrodynamic Power Plant, designed to be the prime mover for a Naval Vessel, is presented. The system is an open cycle, fossil fueled, subsonic MHD Faraday generator with directly fired air preheaters. A superconducting electric transmission drives the propellers and a standard naval steam plant is used as a bottoming cycle. The increased overall efficiency achievable with this plant allows a lighter, smaller volume ship to accommodate the same payload and reduces the overall fuel cost of the vessel

The design study of the JT-60SU device. No. 4. The vacuum vessel and cryostat of JT-60SU

International Nuclear Information System (INIS)

Neyatani, Yuzuru; Ushigusa, Kenkichi; Tobita, Kenji

1997-03-01

The vacuum vessel and the cryostat for the JT-60 Super Upgrade (JT-60SU) have been designed. Two types of the complex materials for the vacuum vessel were chosen on the basis of the avoidance of tritium occlusion and the low irradiation, i.e. (1) SUS316 covered by tungsten plate (30mm thickness) as a γ-ray shielding, (2) Ti-6Al-4V alloy covered by SUS430 plate (1mm thickness) as a tritium protector. Selecting the double skin type of vacuum vessel with toroidally continued structure gave the basic design of the vacuum vessel satisfying the design criteria of the vessel strength for the electromagnetic force, heat load and the property of radiation shielding. The characteristics of the SUS316 covered by tungsten plate type is that as the tungsten can shield the γ-ray, the dose rate inside the vacuum vessel during the maintenance can reduce effectively. The advantage of the Ti-6Al-4V alloy covered by SUS430 plate type vacuum vessel is the quick reduction of the radioactive isotope because of no production of the isotopes with long half-life periods. Channel type and vertical type of the divertor were designed. The sector type of toroidally separated structure was selected for the remote handling. The material of the armor plate was not determined because no material endure the high heat load on the divertor. The cryostat composing the dome and the tank was designed. The electromagnetic force by the eddy current, generated at the plasma start up phase and at the quench of CS super-conducting coil, were small compared to the force produced by the stress limit. (author)

Structural design of shield-integrated thin-wall vacuum vessel and manufacturing qualification tests for International Thermonuclear Experimental Reactor (ITER)

International Nuclear Information System (INIS)

Shimizu, Katsusuke; Shibui, Masanao; Koizumi, Koichi; Kanamori, Naokazu; Nishio, Satoshi; Sasaki, Takashi; Tada, Eisuke

1992-09-01

Conceptual design of shield-integrated thin-wall vacuum vessel has been done for ITER (International Thermonuclear Experimental Reactor). The vacuum vessel concept is based on a thin-double-wall structure, which consists of inner and outer plates and rib stiffeners. Internal shielding structures, which provide neutron irradiation shielding to protect TF coils, are set up between the inner plate and the outer plate of the vessel to avoid complexity of machine systems such as supporting systems of blanket modules. The vacuum vessel is assembled/disassembled by remote handling, so that welding joints are chosen as on-site joint method from reliability of mechanical strength. From a view point of assembling TF coils, the vacuum vessel is separated at the side of port, and is divided into 32 segments similar to the ITER-CDA reference design. Separatrix sweeping coils are located in the vacuum vessel to reduce heat fluxes onto divertor plates. Here, the coil structure and attachment to the vacuum vessel have been investigated. A sectorized saddle-loop coil is available for assembling and disassembling the coil. To support electromagnetic loads on the coils, they are attached to the groove in the vacuum vessel by welding. Flexible multi-plate supporting structure (compression-type gravity support), which was designed during CDA, is optimized by investigating buckling and frequency response properties, and concept on manufacturing and fabrication of the gravity support are proposed. Partial model of the vacuum vessel is manufactured for trial, so that fundamental data on welding and fabrication are obtained. From mechanical property tests of weldment and partial models, mechanical intensity and behaviors of the weldment are obtained. Informations on FEM-modeling are obtained by comparing analysis results with experimental results. (author)

Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

International Nuclear Information System (INIS)

Yun, Sunghwan; Kim, Sang Ji

2015-01-01

In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH 2 and B 4 C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor

Performance evaluation and solar radiation capture of optimally inclined box type solar cooker with parallelepiped cooking vessel design

International Nuclear Information System (INIS)

Sethi, V.P.; Pal, D.S.; Sumathy, K.

2014-01-01

Highlights: • Optimally inclined solar cooker is presented for efficient cooking. • A new parallelepiped shaped cooking vessel for higher solar radiation capture is presented. • Optimum tilt angles of the boosted mirror are computed for maximization of reflected components. • Solar radiation capture ratios show the better cooking performance of inclined cooker. • Standard performance parameters establish the better cooking performance of inclined cooker. - Abstract: An optimally inclined box type solar cooker with single booster mirror is presented along with design and development of a novel parallelepiped shaped cooking vessel design for efficient cooking especially in winter conditions. The main feature of new parallelepiped shaped design is its longer inclined south wall (facing the sun) and a trapezoidal cavity on the vessel lid for greater heat transfer to the food material. The ends of the vessel towards east and west direction are minimized. The cooking performance parameters of proposed inclined cooker coupled with new vessel design were compared with horizontally placed identical cooker of same material and dimensions coupled with conventional cylindrical vessel design during winter month (January) of the year 2010 at Ludhiana climate (30°N 77°E), India. Results showed that the first and the second figures of merit (F 1 and F 2 ) for inclined cooker were 0.16 and 0.54 as compared to 0.14 and 0.43 for horizontally placed cooker. Time taken to boil the water τ boil and standard cooking power P n was 37% less and 40% more respectively in parallelepiped shaped cooking vessel of inclined cooker as compared to conventional cylindrical vessel of horizontally placed cooker. A mathematical model is developed to compute the total solar radiation availability on the absorber plate of inclined as well as horizontal cooker which establishes the better cooking performance of the inclined cooker due to greater width of sun rays intercepting the absorber

Conceptual design of small-sized HTGR system (4). Plant design and technical feasibility

International Nuclear Information System (INIS)

Ohashi, Hirofumi; Sato, Hiroyuki; Yan, Xing L.; Sumita, Junya; Nomoto, Yasunobu; Tazawa, Yujiro; Noguchi, Hiroki; Imai, Yoshiyuki; Tachibana, Yukio

2013-09-01

Japan Atomic Energy Agency (JAEA) has started a conceptual design of a 50MWt small-sized high temperature gas cooled reactor (HTGR) for steam supply and electricity generation (HTR50S), which is a first-of-kind of the commercial plant or a demonstration plant of a small-sized HTGR system for steam supply to the industries and district heating and electricity generation by a steam turbine, to deploy in developing countries in the 2020s. HTR50S was designed for steam supply and electricity generation by the steam turbine with the reactor outlet temperature of 750degC as a reference plant configuration. On the other hand, the intermediate heat exchanger (IHX) will be installed in the primary loop to demonstrate the electricity generation by the helium gas turbine and hydrogen production by thermochemical water splitting by utilizing the secondary helium loop with the reactor outlet temperature of 900degC as a future plant configuration. The plant design of HTR50S for the steam supply and electricity generation was performed based on the plant specification and the requirements for each system taking into account for the increase of the reactor outlet coolant temperature from 750degC to 900degC and the installation of IHX. The technical feasibility of HTR50S was confirmed because the designed systems (i.e., reactor internal components, reactor pressure vessel, vessel cooling system, shutdown cooling system, steam generator (SG), gas circulator, SG isolation and drainage system, reactor containment vessel, steam turbine and heat supply system) satisfies the design requirements. The conceptual plant layout was also determined. This paper provides the summary of the plan design and technical feasibility of HTR50S. (author)

Fluid-solid contact vessel having fluid distributors therein

Science.gov (United States)

Jones, Jr., John B.

1980-09-09

Rectangularly-shaped fluid distributors for large diameter, vertical vessels include reinforcers for high heat operation, vertical sides with gas distributing orifices and overhanging, sloped roofs. Devices are provided for cleaning the orifices from a buildup of solid deposits resulting from the reactions in the vessel.

2XIIB vacuum vessel: a unique design

International Nuclear Information System (INIS)

Hibbs, S.M.; Calderon, M.O.

1975-01-01

The 2XIIB mirror confinement experiment makes unique demands on its vacuum system. The confinement coil set encloses a cavity whose surface is comprised of both simple and compound curves. Within this cavity and at the core of the machine is the operating vacuum which is on the order of 10 -9 Torr. The vacuum container fits inside the cavity, presenting an inside surface suitable for titanium getter pumping and a means of removing the heat load imposed by incandescent sublimator wires. In addition, the cavity is constructed of nonmagnetic and nonconducting materials (nonmetals) to avoid distortion of the pulsed confinement field. It is also isolated from mechanical shocks induced in the machine's main structure when the coils are pulsed. This paper describes the design, construction, and operation of the 2XIIB high-vacuum vessel that has been performing successfully since early 1974

Method of burying vessel containing radioactive waste

International Nuclear Information System (INIS)

Koga, Yoshihito.

1989-01-01

A float having an inert gas sealed therein is attached to a tightly closed vessel containing radioactive wastes. The vessel is inserted and kept in a small hole for burying the tightly closed vessel in an excavated shaft in rocks such as of granite or rock salts, while filling bentonite as shielding material therearound. In this case, the float is so adjusted that the apparent specific gravity is made equal or nearer between the tightly closed vessel and the bentonite, so that the rightly closed vessel does not sink and cause direct contact with the rocks even if bentonite flows due to earthquakes, etc. This can prevent radioactivity contamination through water in the rocks. (S.K.)

Effect of fuel assembly mechanical design changes on dynamic response of reactor pressure vessel system under extreme loadings

International Nuclear Information System (INIS)

Bhandari, D.R.; Hankinson, M.F.

1993-01-01

This paper presents the results of a study to assess the effect of fuel assembly mechanical design changes on the dynamic response of a pressurized water reactor vessel and reactor internals under Loss-Of-Coolant Accident (LOCA) conditions. The results of this study show that the dynamic response of the reactor vessel internals and the core under extreme loadings, such as LOCA, is very sensitive to fuel assembly mechanical design changes. (author)

Minimum weight designs for reinforcement of spherical pressure vessels with flush radial nozzles

International Nuclear Information System (INIS)

Yeo, K.T.; Robinson, M.

1978-01-01

A cylinder-sphere pressure vessel, reinforced in the sphere by a section of constant thickness, has been analysed from the point of view of minimum weight. The reinforcement is allowed to be offset from the main sphere and the design has to be such that the test pressure of the vessel equals the limit pressure. It is shown that in most circumstances an economy of weight may be obtained by making the reinforcement thicker, but less extensive, than suggested in a previous proposal. Further benefit can be obtained by offsetting the reinforcement radially outwards so that the inside surfaces of main sphere and reinforcement are flush. (author)

Gas turbine designer computer program - a study of using a computer for preliminary design of gas turbines

Energy Technology Data Exchange (ETDEWEB)

Petersson, Rickard

1995-11-01

This thesis presents calculation schemes and theories for preliminary design of the fan, high pressure compressor and turbine of a gas turbine. The calculations are presented step by step, making it easier to implement in other applications. The calculation schemes have been implemented as a subroutine in a thermodynamic program. The combination of the thermodynamic cycle calculation and the design calculation turned out to give quite relevant results, when predicting the geometry and performance of an existing aero engine. The program developed is able to handle several different gas turbines, including those in which the flow is split (i.e. turbofan engines). The design process is limited to the fan, compressor and turbine of the gas turbine, the rest of the components have not been considered. Output from the program are main geometry, presented both numerically and as a scale plot, component efficiencies, stresses in critical points and a simple prediction of turbine blade temperatures. 11 refs, 21 figs, 1 tab

Reactor-vessel-sectioning demonstration

International Nuclear Information System (INIS)

Lundgren, R.A.

1981-07-01

A successful technical demonstration of simulated reactor vessel sectioning was completed using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel: an air arc gouger was chosen to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. After the simulated vessel was successfully cut from the SS side, another cut was made, starting from the carbon steel side. This cut was also successful. Cutting from the carbon steel side has the advantages of cost reduction since the air arc gouging step is eliminated and contamination controlled because the molten metal is blown inward

Design progress of the ITER vacuum vessel sectors and port structures

International Nuclear Information System (INIS)

Utin, Yu.; Ioki, K.; Alekseev, A.; Bachmann, Ch.; Cho, S.; Chuyanov, V.; Jones, L.; Kuzmin, E.; Morimoto, M.; Nakahira, M.; Sannazzaro, G.

2007-01-01

Recent progress of the ITER vacuum vessel (VV) design is presented. As the ITER construction phase approaches, the VV design has been improved and developed in more detail with the focus on better performance, improved manufacture and reduced cost. Based on achievements of manufacturing studies, design improvement of the typical VV Sector (no. 1) has been nearly finalized. Design improvement of other sectors is in progress-in particular, of the VV Sectors no. 2 and no. 3 which interface with tangential ports for the neutral beam (NB) injection. For all sectors, the concept for the in-wall shielding has progressed and developed in more detail. The design progress of the VV sectors has been accompanied by the progress of the port structures. In particular, design of the NB ports was advanced with the focus on the beam-facing components to handle the heat input of the neutral beams. Structural analyses have been performed to validate all design improvements

Gamma dose rate estimation and operation management suggestions for decommissioning the reactor pressure vessel of HTR-PM

Energy Technology Data Exchange (ETDEWEB)

Sheng Fang; Hong Li; Jianzhu Cao; Wenqian Li; Feng Xie; Jiejuan Tong [Institute of Nuclear and New Energy Technology, Tsinghua, University, Beijing (China)

2013-07-01

China is now designing and constructing a high temperature gas cooled reactor-pebble bed module (HTR-PM). In order to investigate the future decommissioning approach and evaluate possible radiation dose, gamma dose rate near the reactor pressure vessel was calculated for different cooling durations using QAD-CGA program. The source term of this calculation was provided by KORIGEN program. Based on the calculated results, the spatial distribution and temporal changes of gamma dose rate near reactor pressure vessel was systematically analyzed. A suggestion on planning decommissioning operation of reactor pressure vessel of HTRPM was given based on calculated dose rate and the Chinese Standard GB18871-2002. (authors)

Zone separator for multiple zone vessels

Science.gov (United States)

Jones, John B.

1983-02-01

A solids-gas contact vessel, having two vertically disposed distinct reaction zones, includes a dynamic seal passing solids from an upper to a lower zone and maintaining a gas seal against the transfer of the separate treating gases from one zone to the other, and including a stream of sealing fluid at the seal.

Study of new structures adapted to gas-graphite and gas-heavy water reactors

International Nuclear Information System (INIS)

Martin, R.; Roche, R.

1964-01-01

The experience acquired as a result of the operation of the Marcoule reactors and of the construction and start-up of the E.D.F. reactors on the one hand, and the conclusions of research and tests carried out out-of-pile on the other hand, lead to a considerable change in the general design of reactors of the gas-graphite type. The main modifications envisaged are analysed in the paper. The adoption of an annular fuel element and of a down-current cooling will make it possible to increase considerably the specific power and the power output of each channel; as a result there will be a considerable reduction in the number of the channels and a corresponding increase in the size of the unit cell. The graphite stack will have to be adapted to there new conditions. For security reasons, the use of prestressed concrete for the construction of the reactor vessel is becoming more widespread; they could lead to the exchangers and the fuel-handling apparatus becoming integrated inside the vessel (the so-called 'attic' device). A full-size mode) of this attic has been built at Saclay with the participation of EURATOM; the operational results obtained are presented as well as a new original design for the control rods. As for as the gas-heavy-water system is concerned, the research is carried out on two points of design; the first, which retains the use of horizontal pressure tubes, takes into account the experience acquired during the construction of the EL 4 reactor of which it will constitute an extrapolation; the second, arising from the research carried out on the gas-graphite system, will use a pre-stressed concrete vessel for holding the pressure, the moderator being almost at the same pressure as the cooling fluid and the fuel being placed in vertical channels. The relative merits of these two variants are analysed in the present paper. (authors) [fr

Design concept of conducting shell and in-vessel components suitable for plasma vertical stability and remote maintenance scheme in DEMO reactor

Energy Technology Data Exchange (ETDEWEB)

Utoh, Hiroyasu, E-mail: uto.hiroyasu@jaea.go.jp [Japan Atomic Energy Agency, Obuchi, Rokkasho-mura, Aomori-ken 039-3212 (Japan); International Fusion Energy Research Centre, 2-166, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Takase, Haruhiko [Japan Atomic Energy Agency, Obuchi, Rokkasho-mura, Aomori-ken 039-3212 (Japan); International Fusion Energy Research Centre, 2-166, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Sakamoto, Yoshiteru; Tobita, Kenji [Japan Atomic Energy Agency, Obuchi, Rokkasho-mura, Aomori-ken 039-3212 (Japan); Mori, Kazuo; Kudo, Tatsuya [Japan Atomic Energy Agency, Obuchi, Rokkasho-mura, Aomori-ken 039-3212 (Japan); International Fusion Energy Research Centre, 2-166, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Someya, Youji; Asakura, Nobuyuki; Hoshino, Kazuo; Nakamura, Makoto; Tokunaga, Shinsuke [Japan Atomic Energy Agency, Obuchi, Rokkasho-mura, Aomori-ken 039-3212 (Japan)

2016-02-15

Highlights: • Conceptual design of in-vessel component including conducting shell has been investigated. • The conducting shell design for plasma vertical stability was clarified from the plasma vertical stability analysis. • The calculation results showed that the double-loop shell has the most effect on plasma vertical stability. - Abstract: In order to realize a feasible DEMO, we designed an in-vessel component including the conducting shell. The project is affiliated with the broader approach DEMO design activities and is conceptualized from a plasma vertical stability and engineering viewpoint. The dependence of the plasma vertical stability on the conducing shell parameters and the electromagnetic force at plasma disruption were investigated in numerical simulations (programmed in the 3D eddy current analysis code and a plasma position control code). The simulations assumed the actual shape and position of the vacuum vessel and in-vessel components. The plasma vertical stability was most effectively maintained by the double-loop shell.

Pendulum support of the W7-X plasma vessel: Design, tests, manufacturing, assembly, critical aspects, status

Energy Technology Data Exchange (ETDEWEB)

Missal, B., E-mail: bernd.missal@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Teilinstitut Greifswald, Wendelsteinstraße 1, D-17491 Greifswald (Germany); Leher, F.; Schiller, T. [MAN Diesel and Turbo SE, Werftstraße 17, 94469 Deggendorf (Germany); Friedrich, P. [Universität Rostock, FB Maschinenbau und Schiffstechnik, Albert-Einsteins-Straße 2, 18051 Rostock (Germany); Capriccioli, A. [ENEA Frascati, Fusion Technology Unit, Frascati (Italy)

2014-10-15

Highlights: • Plasma vessel support has to allow vertical adjustment and horizontal passive movement. • Planar sliding tables with PTFE do not fulfill all requirements. • Pendulums can fulfill all requirements. • Geometry and material of spherical bearings had to be optimized in calculations and tests. • Optimized pendulums were manufactured and assembled. - Abstract: The superconducting helical advanced stellarator Wendelstein 7-X (W7-X) is under construction at the Max-Planck-Institut für Plasmaphysik (IPP) in Greifswald, Germany. The three dimensional shape of plasma will be generated by 50 non-planar magnetic coils. The plasma vessel geometry follows exactly this three dimensional shape of plasma. To ensure the superconductivity of coils a cryo vacuum has to be generated. Therefore the coils and their support structure are enclosed within the outer vessel. Plasma vessel, coil structures and outer vessel have to be supported separately. This paper will describe the vertical supports of plasma vessel which have to fulfill two special requirements, vertical adjustability and horizontal mobility. These two tasks will be carried out by plasma vessel supports (PVS) with hydraulic cylinders, special sliding tables during assembly and pendulum supports during operating phase. The paper will give an overview of design, calculation, tests, fabrication, assembly, critical aspects and status of PVS.

Heat treatment device for extending the life of a pressure vessel, particularly a reactor pressure vessel

International Nuclear Information System (INIS)

Krauss, P.; Mueller, E.; Poerner, H.; Weber, R.

1979-01-01

A support body in the form of an insulating cylinder is tightly sealed by connected surfaces at its outer circumference to the inner wall of the pressure vessel. It forms an annular heating space. The heat treatment or tempering of the pressure vessel takes place with the reactor space empty and screened from the outside by ceiling bolts. Heating gas or an induction winding can be used as the means of heating. (DG) [de

Design study of a new vacuum vessel for Doublet III

International Nuclear Information System (INIS)

Rawls, J.M.; Davis, L.G.; Anderson, P.M.

1980-10-01

The principal thrust of the project was to examine a single design in enough depth to gain confidence in the feasibility and desirability of specific design features. However, a valuable spin-off of the project was to develop information of a more generic character to aid in future studies of possibilities for Doublet III. For example, we now feel that Doublet III can be reconfigured with any of a variety of new vacuum vessels, poloidal coil sets, and auxiliary heating systems within three years of project initiation, a period that is short compared to the time scale for developing a completely new facility. In addition, this can be accomplished at a fraction of the cost required to develop a comparable facility

Design of Air Ventilation System for Cargo Hold Vessels Using Solar Desiccant

Directory of Open Access Journals (Sweden)

Alam Baheramsyah

2017-09-01

Full Text Available One of the facilities and infrastructure of the vessel is the ventilation system in the cargo hold to maintain the quality. One attempt to avoid high moisture ratios is to provide a dry air supply by using desiccants. The purpose of this thesis is to design the system of air ventilation with solar desiccant by analysis the calculation with decrease air humidity ratio after passing desiccant rotor as well as fulfillment needs of heater and cooling system using heat of exhaust gas and seawater as well as fulfillment of electricity need using solar energy. From the result of analysis obtain to provide air supply in the cargo hold of 437.5 m3 / hour, the specification of rotor desiccant has a diameter of 550 mm with thickness 200 mm to decrease ratio of outside air humidity equal to 83.1% become 46.5%. Dehumidification air temperature of 47.7oC will be lowered to 35oC by using the sea water cooling media. As for the reactivation air heater requirement of 24.292 kW would be to fulfilled by utilizing the exhaust power of 498.12 kW. And for the electric power needs of the syetm is 34,488 wp will be supplied from the total solar module is 33 units with 345 wp per-capacity.

Design and preliminary analysis of in-vessel core catcher made of high-temperature ceramics material in PWR

International Nuclear Information System (INIS)

Xu Hong; Ma Li; Wang Junrong; Zhou Zhiwei

2011-01-01

In order to protect the interior wall of pressure vessel from melting, as an additional way to external reactor vessel cooling (ERVC), a kind of in-vessel core catcher (IVCC) made of high-temperature ceramics material was designed. Through the high-temperature and thermal-resistance characteristic of IVCC, the distributing of heat flux was optimized. The results show that the downward average heat flux from melt in ceramic layer reduces obviously and the interior wall of pressure vessel doesn't melt, keeping its integrity perfectly. Increasing of upward heat flux from metallic layer makes the upper plenum structure's temperature ascend, but the temperature doesn't exceed its melting point. In conclusion, the results indicate the potential feasibility of IVCC made of high-temperature ceramics material. (authors)

Assessing the feasibility of a high-temperature, helium-cooled vacuum vessel and first wall for the Vulcan tokamak conceptual design

International Nuclear Information System (INIS)

Barnard, H.S.; Hartwig, Z.S.; Olynyk, G.M.; Payne, J.E.

2012-01-01

The Vulcan conceptual design (R = 1.2 m, a = 0.3 m, B 0 = 7 T), a compact, steady-state tokamak for plasma–material interaction (PMI) science, must incorporate a vacuum vessel capable of operating at 1000 K in order to replicate the temperature-dependent physical chemistry that will govern PMI in a reactor. In addition, the Vulcan divertor must be capable of handling steady-state heat fluxes up to 10 MW m −2 so that integrated materials testing can be performed under reactor-relevant conditions. A conceptual design scoping study has been performed to assess the challenges involved in achieving such a configuration. The Vulcan vacuum system comprises an inner, primary vacuum vessel that is thermally and mechanically isolated from the outer, secondary vacuum vessel by a 10 cm vacuum gap. The thermal isolation minimizes heat conduction between the high-temperature helium-cooled primary vessel and the water-cooled secondary vessel. The mechanical isolation allows for thermal expansion and enables vertical removal of the primary vessel for maintenance or replacement. Access to the primary vessel for diagnostics, lower hybrid waveguides, and helium coolant is achieved through ∼1 m long intra-vessel pipes to minimize temperature gradients and is shown to be commensurate with the available port space in Vulcan. The isolated primary vacuum vessel is shown to be mechanically feasible and robust to plasma disruptions with analytic calculations and finite element analyses. Heat removal in the first wall and divertor, coupled with the ability to perform in situ maintenance and replacement of divertor components for scientific purposes, is achieved by combining existing helium-cooled techniques with innovative mechanical attachments of plasma facing components, either in plate-type helium-cooled modules or independently bolted, helium-jet impingement-cooled tiles. The vacuum vessel and first wall design enables a wide range of potential PFC materials and configurations to

The TPX vacuum vessel and in-vessel components

International Nuclear Information System (INIS)

Heitzenroeder, P.; Bialek, J.; Ellis, R.; Kessel, C.; Liew, S.

1994-01-01

The Tokamak Physics Experiment (TPX) is a superconducting tokamak with double-null diverters. TPX is designed for 1,000-second discharges with the capability of being upgraded to steady state operation. High neutron yields resulting from the long duration discharges require that special consideration be given to materials and maintainability. A unique feature of the TPX is the use of a low activation, titanium alloy vacuum vessel. Double-wall vessel construction is used since it offers an efficient solution for shielding, bakeout and cooling. Contained within the vacuum vessel are the passive coil system, Plasma Facing Components (PFCs), magnetic diagnostics, and the internal control coils. All PFCs utilize carbon-carbon composites for exposed surfaces

40 CFR 65.144 - Fuel gas systems and processes to which storage vessel, transfer rack, or equipment leak...

Science.gov (United States)

2010-07-01

...; (ii) Transformed by chemical reaction into materials that are not regulated materials; (iii... section for a storage vessel, the owner or operator shall prepare a design evaluation (or engineering...

Gas emission during laparoscopic colorectal surgery using a bipolar vessel sealing device: A pilot study on four patients

Directory of Open Access Journals (Sweden)

Gianella Michele

2008-09-01

Full Text Available Abstract Background Dissection during laparoscopic surgery produces smoke containing potentially toxic substances. The aim of the present study was to analyze smoke samples produced during laparoscopic colon surgery using a bipolar vessel sealing device (LigaSure™. Methods Four consecutive patients undergoing left-sided colectomy were enrolled in this pilot study. Smoke was produced by the use of LigaSure™. Samples (5,5l were evacuated from the pneumoperitoneum in a closed system into a reservoir. Analysis was performed with CO2-laser-based photoacoustic spectroscopy and confirmed by a Fourier-transform infrared spectrum. The detected spectra were compared to the available spectra of known toxins. Results Samples from four laparoscopic sigmoid resections were analyzed. No relevant differences were noted regarding patient and operation characteristics. The gas samples were stable over time proven by congruent control measurements as late as 24 h after sampling. The absorption spectra differed considerably between the patients. One broad absorption line at 100 ppm indicating H2O and several unknown molecules were detected. With a sensitivity of alpha min ca 10-5 cm-1 no known toxic substances like phenol or indole were identified. Conclusion The use of a vessel sealing device during laparoscopic surgery does not produce known toxic substances in relevant quantity. Further studies are needed to identify unknown molecules and to analyze gas emission under various conditions.

Behavior of a corium jet in high pressure melt ejection from a reactor pressure vessel

International Nuclear Information System (INIS)

Frid, W.

1988-04-01

Discharge of the molten core debris from a pressurized reactor vessel has been recognized as an important accident scenario for pressurized water reactors. Recent high-pressure melt streaming experiments conducted at Sandia National Laboratories, designed to study cavity and containment events related to melt ejection, have resulted in two important observations: (1) Expansion and breakup of the ejected molten jet. (2) Significant aerosol generation during the ejection process. The expansion and breakup of the jet in the experiments are attributed to rapid evolution of the pressurizing gas (nitrogen or hydrogen) dissolved in the melt. It has been concluded that aerosol particles may be formed by condensation of melt vapor and mechanical breakup of the melt and generation. It was also shown that the above stated phenomena are likely to occur in reactor accidents. This report provides results from analytical and experimental investigations on the behavior of a gas supersaturated molten jet expelled from a pressurized vessel. Aero-hydrodynamic stability of liquid jets in gas, stream degassing of molten metals, and gas bubble nucleation in molten metals are relevant problems that are addressed in this work

Natural Gas Container Transportation: the Alternative Way to Solve the World’s Energy Transportation Problems

Directory of Open Access Journals (Sweden)

A.M. Shendrik

2014-03-01

Full Text Available The container gas transportation for low and medium level consumers as an alternative to pipelines is considered. The options for gas supply schemes, based on road and rail transport are given. The advantages and disadvantages of both types of gas transporting are described, the areas of their effective using are separated in the article. Promising implementations of technology in environment of economic crisis and also considering world trends of energy development are presented. The most advanced organization of compressed gas condensate transportation of unprepared gas fields in large diameter universal cylindrical balloons (up to 1000 mm are reasoned. The problem of compressed gas sea transportation are well disclosed, but the alternative ways of gas transportation by land are not investigated enough. Compressed Natural Gas (CNG Technology - is new promising technology for natural gas transportation by specially designed vessels – CNG-vessels. The feature of this technology is that natural gas can be downloaded directly near gas deposits and unloaded - directly into the customer's network. This eliminates significant capital investments in underwater pipelining or gas liquefaction plants. The main objects of investment are CNG-vessels themselves. The most attractive places for implementation of CNG-technology are sea (offshore natural gas deposits. Numerous international experts estimate the natural gas transportation by CNG-vessels in 1.5-2.0 times more cost-beneficial in comparison with offshore pipelines transportation, or in comparison with LNG (Liquefied Natural Gas shipping with natural gas transportation volume between 0.5 and 4.0 billion cubic meters per year on the route from 250 to 2,500 sea miles. This technology makes possible to provide gas supplement to the mountain and abounding in water areas, remote and weakly gasified regions. Described technology deserves special attention in the case of depleted and low-power oil and

Design and fabrication methods of FW/blanket, divertor and vacuum vessel for ITER

Science.gov (United States)

Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Ibbott, C.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Tivey, R.; Utin, Y.; Yamada, M.

2000-12-01

Design has progressed on the vacuum vessel, FW/blanket and Divertor for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design [K. Ioki et al., J. Nucl. Mater. 258-263 (1998) 74]. Design and fabrication methods of the components have been improved to achieve ˜50% reduction of the construction cost. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R&D performed so far during the Engineering Design Activities (EDAs) are still applicable. Further cost reduction methods are also being investigated and additional R&D is being performed.

Clay Corner: Recreating Chinese Bronze Vessels.

Science.gov (United States)

Gamble, Harriet

1998-01-01

Presents a lesson where students make faux Chinese bronze vessels through slab or coil clay construction after they learn about the history, function, and design of these vessels. Utilizes a variety of glaze finishes in order to give the vessels an aged look. Gives detailed guidelines for creating the vessels. (CMK)

Reductions in cost and greenhouse gas emissions with new bulk ship designs enabled by the Panama Canal expansion

International Nuclear Information System (INIS)

Lindstad, Haakon; Jullumstrø, Egil; Sandaas, Inge

2013-01-01

Historically, fuel costs have been small compared with the fixed costs of a bulk vessel, its crewing and management. Today, however, fuel accounts for more than 50% of the total costs. In combination with an introduction of stricter energy efficiency requirements for new vessels, this might make design improvement a necessity for all new bulk vessels. This is in contradiction to traditional bulk vessel designs, where the focus has been on maximizing the cargo-carrying capacity at the lowest possible building cost and not on minimizing the energy consumption. Moreover, the Panama Canal has historically been an important design criterion, while the new canal locks from 2014 will significantly increase the maximum size of vessels that can pass. The present paper provides an assessment of cost and emissions as a function of alternative bulk vessel designs with focus on a vessel's beam, length and hull slenderness, expressed by the length displacement ratio for three fuel price scenarios. The result shows that with slenderer hull forms the emissions drop. With today's fuel price of 600 USD per ton of fuel, emissions can thus be reduced by up to 15–25% at a negative abatement cost. - Highlights: • We have assessed cost and emissions as a function of alternative bulk vessel designs. • The design focus has been on vessel beam, length, hull slenderness and bow section length. • The assessment has taken into account three different fuel price scenarios. • When the block coefficient is reduced and the hull becomes more slender the emissions drop. • With a fuel price of 600 USD/t, emissions can be reduced by up to 15–25% at a negative abatement cost

Software for natural gas pipeline design and simulation (gaspisim ...

African Journals Online (AJOL)

Software for natural gas pipeline design and simulation (gaspisim) ... This paper focuses on the development of software for optimum design and simulation of natural gas pipeline. General ... EMAIL FREE FULL TEXT EMAIL FREE FULL TEXT

Design of Eco Friendly Shallow Draft Fishing Vessel

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Sunardi Sunardi

2016-04-01

Full Text Available One of the main problem of inland waterways fisheries is the transportation of fish from ponds to fish market during low tide trough inland waterways with 0.6m water depth.The boat is experiences grounding due to water depth of the river is not sufficient for the fishing boat to carry fish at it’s maximum2 tones capacity or experience dead freight . This condition forces fisherman to wait until the high tide from the sea, this delay causes the quality of the fish is decreasing.Besides the problem dead freight  problem the existing vessel is causes environmental problem such as erosion of the river bank due to wake wash. The other important issue is the increases of fuel price and it’s scarcity.  This paper presents the results of comparison of existing monohull fishing boat and two other alternativecatamaran designs. The catamaran design alternatives are is ordinary catamaran and flat side catamaran.  Both of the catamaran fishing boat design shows that the catamaran boat with 0.5m draft is able to carry more than 2 tonnes payload during low tide water depth.  The CFD simulation results shows that flat side catamaran resistance is more than 17.7% lower compared to ordinary catamaran and 44% lower compared to monohull. It means that the consumption of flat side catamaran is lowest compared to two other type of hull design. The flat side catamaran also produces lowest wake wash compared to o two other design. The low wake wash means more friendly to environment.

High-performance fiber/epoxy composite pressure vessels

Science.gov (United States)

Chiao, T. T.; Hamstad, M. A.; Jessop, E. S.; Toland, R. H.

1978-01-01

Activities described include: (1) determining the applicability of an ultrahigh-strength graphite fiber to composite pressure vessels; (2) defining the fatigue performance of thin-titanium-lined, high-strength graphite/epoxy pressure vessel; (3) selecting epoxy resin systems suitable for filament winding; (4) studying the fatigue life potential of Kevlar 49/epoxy pressure vessels; and (5) developing polymer liners for composite pressure vessels. Kevlar 49/epoxy and graphite fiber/epoxy pressure vessels, 10.2 cm in diameter, some with aluminum liners and some with alternation layers of rubber and polymer were fabricated. To determine liner performance, vessels were subjected to gas permeation tests, fatigue cycling, and burst tests, measuring composite performance, fatigue life, and leak rates. Both the metal and the rubber/polymer liner performed well. Proportionately larger pressure vessels (20.3 and 38 cm in diameter) were made and subjected to the same tests. In these larger vessels, line leakage problems with both liners developed the causes of the leaks were identified and some solutions to such liner problems are recommended.

Mechanical Behavior of A Metal Composite Vessels Under Pressure At Cryogenic Temperatures

Science.gov (United States)

Tsaplin, A. I.; Bochkarev, S. V.

2016-01-01

Results of an experimental investigation into the deformation and destruction of a metal composite vessel with a cryogenic gas are presented. Its structure is based on basalt, carbon, and organic fibers. The vessel proved to be serviceable at cryogenic temperatures up to a burst pressure of 45 MPa, and its destruction was without fragmentation. A mathematical model adequately describing the rise of pressure in the cryogenic vessel due to the formation of a gaseous phase upon boiling of the liquefied natural gas during its storage without drainage at the initial stage is proposed.

Studies on structural analysis related to the design of the JT-60 vacuum vessel

International Nuclear Information System (INIS)

Takatsu, Hideyuki

1987-06-01

Studies on structural analysis of a vacuum vessel of tokamak-type fusion devices are presented. The present studies are proposals for the structural analysis procedures of the tokamak-type fusion devices and are composed of five parts, each of which covers the fundamental area required for the structural analysis and design; stress analysis, dynamic response analysis, fatigue evaluation, buckling analysis and seismic analysis. Special attention is paid to the critical component, bellows and the critical load, electromagnetic forces. A new finite element method modeling technique is proposed for the stress analysis of U-shaped bellows, where the bellows is replaced by an orthotropic plate having the same stiffness as the bellows. The applicability of the present modeling technique is confirmed by verification tests. Dynamic response and fatigue of the vacuum vessel are critical issues of the structural analysis and design of the tokamak-type fusion devices. Detailed dynamic response analyses of the JT-60 vacuum vessel are presented paying special attention to the dynamic behavior of the U-shaped bellows, where the above-mentioned modeling technique of the U-shaped bellows is applied. A fatigue evaluation method of the vacuum vessel under the dynamic electromagnetic forces is proposed, which utilizes the results of the detailed dynamic response analysis. In the present method, fatigue evaluation method for random loads is applied. Torsional fatigue strength of the welded bellows is experimentally evaluated aiming the application to the port of the fusion device and it is shown that the welded bellows reveals elastic buckling and spiral distortion under a small angle of tortion. Two formulae are proposed to evaluate the stress of the welded bellows under the forced angle of tortion. (author)

Design optimization of anisotropic pressure vessels with manufacturing uncertainties accounted for

International Nuclear Information System (INIS)

Walker, M.; Tabakov, P.Y.

2013-01-01

Accurate optimal design solutions for most engineering structures present considerable difficulties due to the complexity and multi-modality of the functional design space. The situation is made even more complex when potential manufacturing tolerances must be accounted for in the optimizing process. The present study provides an original in-depth analysis of the problem and then a new technique for determining the optimal design of engineering structures, with manufacturing tolerances accounted for, is proposed and demonstrated. The numerical examples used to demonstrate the technique involve the design optimization of anisotropic fibre-reinforced laminated pressure vessels. It is assumed that the probability of any tolerance value occurring within the tolerance band, compared with any other, is equal, and thus it is a worst-case scenario approach. A genetic algorithm with fitness sharing, including a micro-genetic algorithm, has been found to be very suitable to use, and implemented in the technique

Metallurgical failure investigation of a pipe connector fracture of an expansion vessel

International Nuclear Information System (INIS)

Neidel, Andreas

2016-01-01

A pipe connector of an expansion vessel of a safety heat exchanger was torn off in a test facility's natural gas compressor. From a material point of view, the cause of the damage is a fatigue fracture induced by pulsating bending stress. The fatigue fracture originated from both, the pipe's outer surface as well as from its inner surface, which is consistent with the given stress situation (pulsating bending stress). Material defects or welding-induced flaws were not observed. Corrosion, wear, or thermal overload which may have promoted the damage, were not observed either. The primary cause was a major design error. Cases of dynamic load were obviously not duly taken into account during designing, so that the free-swinging mass of the expansion vessel which was mounted to a pipe of a diameter of only half an inch and, furthermore, installed in an angle of 45 (additional static preload.), could cause the fatigue failure induced by pulsating bending stress in the zone of highest stresses at the transition of the expansion vessel and the the pipe connector due to dynamic operating loads which always occur in plants like these.

An Approach for Selection of Flow Regime and Models for Conservative Evaluation of a Vessel Integrity Monitoring System for Water-Cooled Vacuum Vessels

International Nuclear Information System (INIS)

Pointer, W. David; Ruggles, Arthur E.

2003-01-01

Thin-walled vacuum containment vessels cooled by circulating water jackets are often utilized in research and industrial applications where isolation of equipment or experiments from the influences of the surrounding environment is desirable. The development of leaks in these vessels can result in costly downtime for the facility. A Vessel Integrity Monitoring System (VIMS) is developed to detect leak formation and estimate the size of the leak to allow evaluation of the risk associated with continued operation. A wide range of leak configurations and fluid flow phenomena are considered in the evaluation of the rate at which a tracer gas dissolved in the cooling jacket water is transported into the vacuum vessel. A methodology is presented that uses basic fluid flow models and careful evaluation of their ranges of applicability to provide a conservative estimate of the transport rates for the tracer gas and hence the time required for the VIMS to detect a leak of a given size

GAS DRYER DESIGN FOR VERMICELLI USING MICROCONTROLLER ATMEGA 8535

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Akhmad Mufasil

2013-04-01

Full Text Available The purpose of this research is designing a gas dryer for vermicelli using a microcontroller ATmega 8535 that can overcome the obstacles in the drying department. The microcontroller ATmega 8535 used to regulate the temperature of the drying chamber. This tool design is made to compare between conventional dryer and a theoritical gas dryer in the factory. The method used in this paper is field study in the factory. Based on the experience then theoritical gas dryer design from the factory’s datas. A microcontroller Atmega 8535 is added to the dryer to easier the dryer operation. The result of the theoritical gas dryer design for vermicelli using a microcontroller ATmega 8535, showed a faster time drying for 131,04 kg of vermicelli need 47,72 minutes at a temperature of 40- 55° C.

AFB/open cycle gas turbine conceptual design study

Science.gov (United States)

Dickinson, T. W.; Tashjian, R.

1983-09-01

Applications of coal fired atmospheric fluidized bed gas turbine systems in industrial cogeneration are identified. Based on site-specific conceptual designs, the potential benefits of the AFB/gas turbine system were compared with an atmospheric fluidized design steam boiler/steam turbine system. The application of these cogeneration systems at four industrial plant sites is reviewed. A performance and benefit analysis was made along with a study of the representativeness of the sites both in regard to their own industry and compared to industry as a whole. A site was selected for the conceptual design, which included detailed site definition, AFB/gas turbine and AFB/steam turbine cogeneration system designs, detailed cost estimates, and comparative performance and benefit analysis. Market and benefit analyses identified the potential market penetration for the cogeneration technologies and quantified the potential benefits.

In service inspection of SUPERPHENIX 1 vessels: MIR

International Nuclear Information System (INIS)

Asty, M.; Viard, J.; Lerat, B.; Saglio, R.

1985-01-01

Although no in-service inspection constraints were imposed on the Phenix vessels, the Safety Authorities asked that the design of SUPERPHENIX 1 makes it possible to monitor throughout the lifetime of the reactor, surface and internal defects on the main vessel. A pool design and the presence of heat baffles inside the main vessel make access from the inside of the vessel impossible. Thus, an inspection can only be performed from the outside of the main vessel: the distance between the walls of the main and safety vessels is such that an inspection device can be introduced into the corresponding space. As the design of the reactor precludes radiographic inspection, the method which was selected for monitoring internal defects in the main vessel is ultrasonics. However, the anisotropic structure of austenitic stainless steel welds limits the performance of this technique. The authors present the in-service inspection device, MIR, which has been specially developed for the visual and ultrasonic examination of SUPERPHENIX 1 vessels

General design and main problems of a gas-heavy-water power reactor contained in a pressure vessel; Conception generale et principaux problemes d'un reacteur de puissance eau lourde-gaz contenu dans un caisson resistant

Energy Technology Data Exchange (ETDEWEB)

Roche, R; Gaudez, J C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

In the framework of research carried out on a CO{sub 2}-cooled power reactor moderated by heavy water, the so-called 'pressure vessel' solution involves the total integration of the core, of the primary circuit (exchanges and blowers) and of the fuel handling machine inside a single, strong, sealed vessel made of pre-stressed concrete. A vertical design has been chosen: the handling 'attic' is placed above the core, the exchanges being underneath. This solution makes it possible to standardize the type of reactor which is moderated by heavy-water or graphite and cooled by a downward stream of carbon dioxide gas; it has certain advantages and disadvantages with respect to the pressure tube solution and these are considered in detail in this report. Extrapolation presents in particular.problems due specifically to the heavy water (for example its cooling,its purification, the balancing of the pressures of the heavy water and of the gas, the assembling of the internal structures, the height of the attic, etc. (authors) [French] Dans le cadre des etudes d'un reacteur de puissance modere a l'eau lourde et refroidi-au gaz carbonique, la solution dite 'en caisson' consiste en une integration totale du coeur, du circuit primaire (echangeurs et soufflantes) et du dispositif de manutention du combustible a l'interieur d'un meme caisson etanche et resistant en beton precontraint. La disposition envisagee est verticale; le grenier de manutention est dispose au-dessus du coeur, les echangeurs en dessous. Cette solution, qui permet d'uniformiser les types de reacteurs moderes a l'eau lourde ou au graphite et refroidis par une circulation descendante de gaz carbonique presente, par rapport a la solution a tube de force, des avantages et des inconvenients qui sont analyses dans cette etude. L'extrapolation pose, en particulier, des problemes specifiques a l'eau lourde (tels que son refroidissement, son epuration, l'equilibrage des pression entre l'eau lourde et le gaz, le montage

Design and fabrication methods of FW/blanket, divertor and vacuum vessel for ITER

International Nuclear Information System (INIS)

Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Ibbott, C.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Tivey, R.; Utin, Y.; Yamada, M.

2000-01-01

Design has progressed on the vacuum vessel, FW/blanket and Divertor for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design [K. Ioki et al., J. Nucl. Mater. 258-263 (1998) 74]. Design and fabrication methods of the components have been improved to achieve ∼50% reduction of the construction cost. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R and D performed so far during the Engineering Design Activities (EDAs) are still applicable. Further cost reduction methods are also being investigated and additional R and D is being performed

Design and fabrication methods of FW/blanket, divertor and vacuum vessel for ITER

Energy Technology Data Exchange (ETDEWEB)

Ioki, K. E-mail: iokik@itereu.deiokik@ipp.mpg.de; Barabash, V.; Cardella, A.; Elio, F.; Ibbott, C.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Tivey, R.; Utin, Y.; Yamada, M

2000-12-01

Design has progressed on the vacuum vessel, FW/blanket and Divertor for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design [K. Ioki et al., J. Nucl. Mater. 258-263 (1998) 74]. Design and fabrication methods of the components have been improved to achieve {approx}50% reduction of the construction cost. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R and D performed so far during the Engineering Design Activities (EDAs) are still applicable. Further cost reduction methods are also being investigated and additional R and D is being performed.

Pressure vessel and method therefor

Science.gov (United States)

Saunders, Timothy

2017-09-05

A pressure vessel includes a pump having a passage that extends between an inlet and an outlet. A duct at the pump outlet includes at least one dimension that is adjustable to facilitate forming a dynamic seal that limits backflow of gas through the passage.

Structural analysis and manufacture for the vacuum vessel of experimental advanced superconducting tokamak (EAST) device

International Nuclear Information System (INIS)

Song Yuntao; Yao Damao; Wu Songata; Weng Peide

2006-01-01

The experimental advanced superconducting tokamak (EAST) is an advanced steady-state plasma physics experimental device, which has been approved by the Chinese government and is being constructed as the Chinese national nuclear fusion research project. The vacuum vessel, that is one of the key components, will have to withstand not only the electromagnetic force due to the plasma disruption and the Halo current, but also the pressure of boride water and the thermal stress due to the 250 deg. C baking out by the hot pressure nitrogen gas, or the 100 deg. C hot wall during plasma operation. This paper is a report of the mechanical analyses of the vacuum vessel. According to the allowable stress criteria of American Society of Mechanical Engineers, Boiler and Pressure Vessel Committee (ASME), the maximum integrated stress intensity on the vacuum vessel is 396 MPa, less than the allowable design stress intensity 3S m (441 MPa). At the same time, some key R and D issues are presented, which include supporting system, bellows and the assembly of the whole vacuum vessel

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)

The design, safety and project development status of the modular high temperature gas-cooled reactor in the United States

International Nuclear Information System (INIS)

Mears, L.D.; Dean, R.A.

1987-01-01

The cooperative government and industry Modular High Temperature Gas-Cooled Reactor (MHTGR) Program in the United States has advanced a 350 MW(t) plant design through the conceptual development stage. The system incorporates an annular core of prismatic fuel elements within a steel pressure vessel connected, in a side-by-side arrangement, by a concentric duct to a second steel vessel containing a steam generator and helium coolant circulator. The reference plant design consists of four reactor modules installed in separate below-grade silos, providing steam to two conventional turbine generators. The nominal net plant output is 540 MW(e). The small reactor system takes unique advantage of the high temperature capability of the refractory coated fuel and the large thermal inertia of the graphite moderator to provide a design capable of withstanding a complete loss of active core cooling without causing excessive core heatup and significant release of fission products from the fuel. Present program activities are concentrated on interactions with the Nuclear Regulatory Commission aimed at obtaining a Licensability Statement. A project initiative to build a prototype plant which would demonstrate the MHTGR-unique licensing process, plant performance, costs and schedule plus establish an industrial infrastructure to proceed with follow-on commercial MHTGR plants by the turn of the century, is being undertaken by the utility/vendor participants (author)

High Performance Marine Vessels

CERN Document Server

Yun, Liang

2012-01-01

High Performance Marine Vessels (HPMVs) range from the Fast Ferries to the latest high speed Navy Craft, including competition power boats and hydroplanes, hydrofoils, hovercraft, catamarans and other multi-hull craft. High Performance Marine Vessels covers the main concepts of HPMVs and discusses historical background, design features, services that have been successful and not so successful, and some sample data of the range of HPMVs to date. Included is a comparison of all HPMVs craft and the differences between them and descriptions of performance (hydrodynamics and aerodynamics). Readers will find a comprehensive overview of the design, development and building of HPMVs. In summary, this book: Focuses on technology at the aero-marine interface Covers the full range of high performance marine vessel concepts Explains the historical development of various HPMVs Discusses ferries, racing and pleasure craft, as well as utility and military missions High Performance Marine Vessels is an ideal book for student...

Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 1. Numerical investigation for the rationalization of hydrodynamics in the upper plenum

International Nuclear Information System (INIS)

Muramatsu, Toshiharu; Yamaguchi, Akira

2002-02-01

A large-scale sodium-cooled fast breeder reactor in feasibility studies on commercialized fast reactors has a tendency of consideration of thorough simplified and compacted system designs to realize drastic economical improvements. Therefore, special attention should be paid to thermohydraulic designs for a gas entrainment behavior from free surfaces, a flow-induced vibration of in-vessel components, a thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. In-vessel thermohydraulic analyses were carried out using a multi-dimensional code AQUA to understand the thermohydraulic characteristics in the upper plenum, and to investigate suitable in-vessel structure for the elimination of gas entrainment possibility. From the analysis, the following results were obtained. (1) It is difficult to rationale in-vessel flow patterns through adjustments of porous ratio and pressure loss for a hold down plate and baffle plates installed in an upper core structure. (2) Dummy plug insertion to a slit of the upper core structure is one of effective measures to stabilize in-vessel flow patterns. (3) Flow guide devices such as a baffle ring and a partial inner barrel are also effective measures to eliminate impinging jet to a dipped plate (D/P) and to reduce horizontal flow velocity components at free surface. (4) Installations of labyrinth structures to a R/V - D/P gap is successful for decreasing of free surface horizontal flows. (5) Gap closing of an in-vessel fuel pot and two cold trap components has the effects of reductions for free surface horizontal flows and for the difference of free surface levels. Following future investigations are important preventive measures against the gas entrainment from the free surface. (1) Flattening of spatial axial velocity distributions at the R/V - D/P gap. (2) Alleviation measures of vortex concentration at free surface. (3) Separation measures of 3-dimensional vortex

Gas storage and processing device

International Nuclear Information System (INIS)

Kobayashi, Yoshihiro.

1988-01-01

Purpose: To improve the gas solidification processing performance in a gas storing and processing device for solidifying treatment of radioactive gaseous wastes (krypton 85) by ion injection method. Constitution: The device according to the present invention is constituted by disposing a coil connected with a magnetic field power source to the outer circumference of an outer cathode vessel, so that axial magnetic fields are formed to the inside of the outer cathode vessel. With such a device, thermoelectrons released from the thermocathode downwardly collide against gaseous radioactive wastes at high probability while moving spirally by the magnetic fields. The thus formed gas ions are solidified by sputtering in the cathode in the vessel. (Horiuchi, T.)

Design prediction for long term stress rupture service of composite pressure vessels

Science.gov (United States)

Robinson, Ernest Y.

1992-01-01

Extensive stress rupture studies on glass composites and Kevlar composites were conducted by the Lawrence Radiation Laboratory beginning in the late 1960's and extending to about 8 years in some cases. Some of the data from these studies published over the years were incomplete or were tainted by spurious failures, such as grip slippage. Updated data sets were defined for both fiberglass and Kevlar composite stand test specimens. These updated data are analyzed in this report by a convenient form of the bivariate Weibull distribution, to establish a consistent set of design prediction charts that may be used as a conservative basis for predicting the stress rupture life of composite pressure vessels. The updated glass composite data exhibit an invariant Weibull modulus with lifetime. The data are analyzed in terms of homologous service load (referenced to the observed median strength). The equations relating life, homologous load, and probability are given, and corresponding design prediction charts are presented. A similar approach is taken for Kevlar composites, where the updated stand data do show a turndown tendency at long life accompanied by a corresponding change (increase) of the Weibull modulus. The turndown characteristic is not present in stress rupture test data of Kevlar pressure vessels. A modification of the stress rupture equations is presented to incorporate a latent, but limited, strength drop, and design prediction charts are presented that incorporate such behavior. The methods presented utilize Cartesian plots of the probability distributions (which are a more natural display for the design engineer), based on median normalized data that are independent of statistical parameters and are readily defined for any set of test data.

Gas Turbine/Solar Parabolic Trough Hybrid Designs: Preprint

Energy Technology Data Exchange (ETDEWEB)

Turchi, C. S.; Ma, Z.; Erbes, M.

2011-03-01

A strength of parabolic trough concentrating solar power (CSP) plants is the ability to provide reliable power by incorporating either thermal energy storage or backup heat from fossil fuels. Yet these benefits have not been fully realized because thermal energy storage remains expensive at trough operating temperatures and gas usage in CSP plants is less efficient than in dedicated combined cycle plants. For example, while a modern combined cycle plant can achieve an overall efficiency in excess of 55%; auxiliary heaters in a parabolic trough plant convert gas to electricity at below 40%. Thus, one can argue the more effective use of natural gas is in a combined cycle plant, not as backup to a CSP plant. Integrated solar combined cycle (ISCC) systems avoid this pitfall by injecting solar steam into the fossil power cycle; however, these designs are limited to about 10% total solar enhancement. Without reliable, cost-effective energy storage or backup power, renewable sources will struggle to achieve a high penetration in the electric grid. This paper describes a novel gas turbine / parabolic trough hybrid design that combines solar contribution of 57% and higher with gas heat rates that rival that for combined cycle natural gas plants. The design integrates proven solar and fossil technologies, thereby offering high reliability and low financial risk while promoting deployment of solar thermal power.

Aluminum and Other Coatings for the Passivation of Tritium Storage Vessels

Energy Technology Data Exchange (ETDEWEB)

Spencer, W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Korinko, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-11-16

Using a highly sensitive residual gas analyzer, the off-gassing of hydrogen, water, and hydrocarbons from surface-treated storage vessels containing deuterium was measured. The experimental storage vessels were compared to a low-off-gassing, electro-polished 304L canister. Alternative vessels were made out of aluminum, or were coatings on 304L steel. Coatings included powder pack aluminide, electro-plated aluminum, powder pack chromide, dense electro-plated chromium, copper plated, and copper plated with 25 and 50 percent nano-diamond. Vessels were loaded with low pressure deuterium to observe exchange with protium or hydrogen as observed with formation of HD and HDO. Off gas of D₂O or possible CD₄ was observed at mass 20. The main off-gas in all of the studies was H₂. The studies indicated that coatings required significant post-coating treatment to reduce off-gas and enhance the permeation barrier from gases likely added during the coating process. Dense packed aluminum coatings needed heating to drive off water. Electro-plated aluminum, chromium and copper coatings appeared to trap hydrogen from the plating process. Nano-diamond appeared to enhance the exchange rate with hydrogen off gas, and its coating process trapped significant amounts of hydrogen. Aluminum caused more protium exchange than chromium-treated surfaces. Aluminum coatings released more water, but pure aluminum vessels released small amounts of hydrogen, little water, and generally performed well. Chromium coating had residual hydrogen that was difficult to totally outgas but otherwise gave low residuals for water and hydrocarbons. Our studies indicated that simple coating of as received 304L metal will not adequately block hydrogen. The base vessel needs to be carefully out-gassed before applying a coating, and the coating process will likely add additional hydrogen that must be removed. Initial simple bake-out and leak checks up to 350° C for a few hours was

A Novel Boil-Off Gas Re-Liquefaction Using a Spray Recondenser for Liquefied Natural-Gas Bunkering Operations

Directory of Open Access Journals (Sweden)

Jiheon Ryu

2016-11-01

Full Text Available This study presents the design of a novel boil-off gas (BOG re-liquefaction technology using a BOG recondenser system. The BOG recondenser system targets the liquefied natural gas (LNG bunkering operation, in which the BOG phase transition occurs in a pressure vessel instead of a heat exchanger. The BOG that is generated during LNG bunkering operation is characterized as an intermittent flow with various peak loads. The system was designed to temporarily store the transient BOG inflow, condense it with subcooled LNG and store the condensed liquid. The superiority of the system was verified by comparing it with the most extensively employed conventional re-liquefaction system in terms of consumption energy and via an exergy analysis. Static simulations were conducted for three compositions; the results indicated that the proposed system provided 0 to 6.9% higher efficiencies. The exergy analysis indicates that the useful work of the conventional system is 24.9%, and the useful work of the proposed system is 26.0%. Process dynamic simulations of six cases were also performed to verify the behaviour of the BOG recondenser system. The results show that the pressure of the holdup in the recondenser vessel increased during the BOG inflow mode and decreased during the initialization mode. The maximum pressure of one of the bunkering cases was 3.45 bar. The system encountered a challenge during repetitive operations due to overpressurizing of the BOG recondenser vessel.

Modelling of Rotor-gas bearings for Feedback Controller Design

DEFF Research Database (Denmark)

Theisen, Lukas Roy Svane; Niemann, Hans Henrik

2014-01-01

Controllable rotor-gas bearings are popular oering adaptability, high speed operation, low friction and clean operation. Rotor-gas bearings are however highly sensitive to disturbances due to the low friction of the injected gas. These undesirable damping properties call for controllers, which ca...... and are shown to accurately describe the dynamical behaviour of the rotor-gas bearing. Design of a controller using the identied models is treated and experiments verify the improvement of the damping properties of the rotor-gas bearing.......Controllable rotor-gas bearings are popular oering adaptability, high speed operation, low friction and clean operation. Rotor-gas bearings are however highly sensitive to disturbances due to the low friction of the injected gas. These undesirable damping properties call for controllers, which can...... be designed from suitable models describing the relation from actuator input to measured shaft position. Current state of the art models of controllable gas bearings however do not provide such relation, which calls for alternative strategies. The present contribution discusses the challenges for feedback...

Aspects of the design and structural analysis of the prestressed cast iron nuclear reactor pressure vessel

International Nuclear Information System (INIS)

Thomas, R.G.

1978-09-01

The development of the prestressed cast iron nuclear reactor pressure vessel up to the present time is reviewed, and the current status is outlined of the techniques used for its structural analysis. Details of the manufacturing processes involved in the production of the castings, and problems of inspecting them to the standards required for a nuclear application are discussed. A method for the detailed modelling of the cast iron segments is proposed, using the finite element technique with plate bending elements, and criteria for obtaining accurate results are derived. The application of the technique to the analysis of a single cast segment situated in the wall of a PCIPV has enabled an accurate determination of the stress field to be made. Account is taken of the effect of the vessel displacements on the tendon stresses at normal vault pressure and at high overpressure. Studies by this method of several different casting designs have identified favourable features, which have been incorporated into an optimised design. The sensitivity of the structure to a machining error in a casting and to the failure or removal of circumferential and axial tendons is examined, making use of axisymmetric and three-dimensional global finite element solutions to provide boundary conditions for detailed local analyses. Some aspects of the economics of the cast iron reactor pressure vessel are discussed, and recommendations are made for further research in areas relevant to the assessment of the reliability of the vessel. (author)

Prestressed concrete reactor vessel for the HHT-670 MW(e) demonstration plant. Pt.1. Design of the multi-cavity prestressed concrete reactor vessel with warm liner

International Nuclear Information System (INIS)

Lafitte, R.; Marchand, J.D.

1979-01-01

The design studies and tests described in this paper were undertaken as part of ''PROJECT HHT'', a German-Swiss joint effort for the development of high-temperature helium cooled reactors with direct-cycle turbine. The prestressed concrete reactor pressure vessel encloses the core of the reactor itself, the heat exchangers (coolers and recuperators), the helium turbine, the main helium circuit, all nuclear and thermal equipment, and auxiliary reactor cooling equipment. In order to make the liner accessible for inspection, no thermal insulation is provided between the coolant and the liner. The temperature of the helium in contact with the liner is limited to 200 0 C, under all normal operation conditions of the reactor. In the HHT reactor pressure vessel, the resisting structure is protected thermally by a layer of warm concrete between the liner and the structural prestressed concrete. The main features of this pressure vessel are the marked pressure differences in the cavities during normal operation, and the use of warm liner. The objectives of the reference design were chiefly related to the sizing up of the main structure, taking into account the modifications to be expected in the material characteristics as a result of the high temperatures developed

Comparative study for the design of optimal composite pressure vessels

International Nuclear Information System (INIS)

Butt, A.M.; Haq, S.W.U.

2009-01-01

Composite pressure vessels require special design attention to the dome region because of the varying wind angles generated using the filament winding process. Geometric variations in the dome region cause the fiber to change angels and thickness and hence offer difficulty to acquire a constant stress profile (isotensoid). Therefore a dome contour which allows an isotensoid behavior is the required structure. Two design methods to generate dome profiles for similar dome openings were investigated namely Netting Analysis and Optimal Design method. Both methods assume that loads are carried by the fiber alone (monotropic) ignoring the complete composite behavior. Former method produced a lower dome internal volume and a higher fiber thickness as compared to the later optimal design method when studied against different normalized dome opening radiuses. The optimal dome contour was studied in ANSYS with a trial design. The dome was considered to have transversely isotropic property with a dome contour based on monotropic model. While investigating the dome with non linear large displacement finite element analysis, the dome still exhibited isotensoid behavior with transverse isotropic material assignment. Elliptic integrals were used to generate the optimal dome contours and hence elliptic dome contours were formed which were isotensoid in nature with complete composite representation. (author)

High temperature gas cooled nuclear reactor

International Nuclear Information System (INIS)

Hosegood, S.B.; Lockett, G.E.

1975-01-01

For high-temperature gas cooled reactors it is considered advantageous to design the core so that the moderator blocks can be removed and replaced by some means of standpipes normally situated in the top of the reactor vessel. An arrangement is here described to facilitate these operations. The blocks have end faces shaped as irregular hexagons with three long sides of equal length and three short sides also of equal length, one short side being located between each pair of adjacent long sides, and the long sides being inclined towards one another at 60 0 . The block defines a number of coolant channels located parallel to its sides. Application of the arrangement to a high temperature gas-cooled reactor with refuelling standpipes is described. The standpipes are located in the top of the reactor vessel above the tops of the columns and are disposed coaxially above the hexagonal channels, with diameters that allow the passage of the blocks. (U.K.)

Reactor-vessel-sectioning demonstration

International Nuclear Information System (INIS)

Lundgren, R.A.

1981-09-01

A technical demonstration was successfully completed of simulated reactor vessel sectioning using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in. layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel; air arc gouging was selected to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. Three sectioning operations were demonstrated. For all three, the operating parameters were the same; but the position of the sample was varied. For the first cut, the sample was placed in a horizontal position, and it was successfully severed from the SS side. For the second cut, the sample was turned over and cut from the carbon steel side. Cutting from the carbon steel side has the advantages of cost reduction

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

Development of user-friendly structural design system for pressure vessels

International Nuclear Information System (INIS)

Sato, Takuya; Nomoto, Taeko; Kado, Kenichiro; Yagawa, Genki; Yoshimura, Shinobu.

1996-01-01

In this paper we describe a new user-friendly structural design system for pressure vessels, which is based on finite element stress analyses. The basic concept of the developed system is to minimize input data required for the finite element analysis and to perform the analysis quickly. To realize this, the system is equipped with the finite element modeling module based on fuzzy knowledge processing, the input data generation module, the finite element analyzer, the graphic user-interface module for analysis results, and the stress evaluation module. Fundamental performance of the present system is clearly demonstrated through the analysis of a top nozzle. (author)

JT-60SA vacuum vessel manufacturing and assembly

Energy Technology Data Exchange (ETDEWEB)

Masaki, Kei, E-mail: masaki.kei@jaea.go.jp [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan); Shibama, Yusuke K.; Sakurai, Shinji; Shibanuma, Kiyoshi; Sakasai, Akira [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan)

2012-08-15

Highlights: Black-Right-Pointing-Pointer The design of the JT-60SA vacuum vessel body was completed with the demonstration of manufacturing procedure by the mock-up fabrication of the 20 Degree-Sign upper half of VV. Black-Right-Pointing-Pointer The actual VV manufacturing has started since November 2009. Black-Right-Pointing-Pointer The first product of the VV 40 Degree-Sign sector was completed in May 2011. Black-Right-Pointing-Pointer A basic VV assembly scenario and procedure were studied to complete the 360 Degree-Sign VV including positioning method and joint welding. - Abstract: The JT-60SA vacuum vessel (VV) has a D-shaped poloidal cross section and a toroidal configuration with 10 Degree-Sign segmented facets. A double wall structure is adopted to ensure high rigidity at operational load and high toroidal one-turn resistance. The material is 316L stainless steel with low cobalt content (<0.05%). The design temperatures of the VV at plasma operation and baking are 50 Degree-Sign C and 200 Degree-Sign C, respectively. In the double wall, boric-acid water is circulated at plasma operation to reduce the nuclear heating of the superconducting magnets. For baking, nitrogen gas is circulated in the double wall after draining of the boric-acid water. The manufacturing of the VV started in November 2009 after a fundamental welding R and D and a trial manufacturing of 20 Degree-Sign upper half mock-up. The manufacturing of the first VV 40 Degree-Sign sector was completed in May 2011. A basic concept and required jigs of the VV assembly were studied. This paper describes the design and manufacturing of the vacuum vessel. A plan of VV assembly in torus hall is also presented.

Neutron and Gamma Fluxes and dpa Rates for HFIR Vessel Beltline Region (Present and Upgrade Designs)

Energy Technology Data Exchange (ETDEWEB)

Blakeman, E.D.

2001-01-11

The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) is currently undergoing an upgrading program, a part of which is to increase the diameters of two of the four radiation beam tubes (HB-2 and HB-4). This change will cause increased neutron and gamma radiation dose rates at and near locations where the tubes penetrate the vessel wall. Consequently, the rate of radiation damage to the reactor vessel wall at those locations will also increase. This report summarizes calculations of the neutron and gamma flux (particles/cm{sup 2}/s) and the dpa rate (displacements/atom/s) in iron at critical locations in the vessel wall. The calculated dpa rate values have been recently incorporated into statistical damage evaluation codes used in the assessment of radiation induced embrittlement. Calculations were performed using models based on the discrete ordinates methodology and utilizing ORNL two-dimensional and three-dimensional discrete ordinates codes. Models for present and proposed beam tube designs are shown and their results are compared. Results show that for HB-2, the dpa rate in the vessel wall where the tube penetrates the vessel will be increased by {approximately}10 by the proposed enlargement. For HB-4, a smaller increase of {approximately}2.6 is calculated.

Nuclear reactor coolant and cover gas system

International Nuclear Information System (INIS)

George, J.A.; Redding, A.H.; Tower, S.N.

1976-01-01

A core cooling system is disclosed for a nuclear reactor of the type utilizing a liquid coolant with a cover gas above free surfaces of the coolant. The disclosed system provides for a large inventory of reactor coolant and a balanced low pressure cover gas arrangement. A flow restricting device disposed within a reactor vessel achieves a pressure of the cover gas in the reactor vessel lower than the pressure of the reactor coolant in the vessel. The low gas pressure is maintained over all free surfaces of the coolant in the cooling system including a coolant reservoir tank. Reactor coolant stored in the reservoir tank allows for the large reactor coolant inventory provided by the invention

Design and application of a surface vessel for autonomous inland water monitoring

OpenAIRE

Hitz Gregory; Pomerleau Francois; Garneau Marie-Eve; Pradalier Cedric; Posch Thomas; Pernthaler Jakob; Siegwart Roland

2012-01-01

This article presents a novel autonomous surface vessel (ASV) that was designed and manufactured specifically for the monitoring of water resources resources that are not only constantly drained but also face the growing threat of mass proliferation (bloom) of noxious cyanobacteria. On one hand the distribution of these blooms in a given water body requires a surveillance of biological data at high spatial resolution on both vertical and horizontal axes whereas on the other hand the understan...

Design and issues of the ITER in-vessel components: ITER Joint central team and home teams

International Nuclear Information System (INIS)

Parker, R.R.

1998-01-01

This paper surveys the status of the design of the in-vessel components for ITER, in particular the major components, namely the vacuum vessel, blanket and first wall, and divertor, and the interface of selected ancillary systems such as those used for RF heating and current drive, and for diagnostics. The vacuum vessel is a double-walled structure constructed from two toroidal shells joined by ribs. The space between the skins is filled with shield plates directly cooled by water. The structural material is 316 LN IG (ITER grade). Toroidal supports joining the vessel midplane ports with the TF structure limit possible differential toroidal displacements, as might occur due to seismic or vertical displacement events (VDEs). A variety of load conditions corresponding to normal and off-normal loads have been considered and in all cases peak vessel stresses are within allowables. The blanket system consists of approximately 700 modules, each weighing ∝4 t. The integrated first wall consists of a beryllium-tiled copper mat bonded to the water-cooled SS shield block. The copper mat functions as a heat sink and has imbedded in it an array of SS tubes providing water cooling. The modules are mechanically attached to a toroidal backplate. Loads due to centered disruptions are reacted via hoop stress in the backplate, whereas net vertical and horizontal loads such as those arising from VDEs are transferred through the backplate and divertor supports to the vessel. (orig.)

Prestressed concrete vessels suitable for helium high temperature reactors

International Nuclear Information System (INIS)

Lockett, G.E.; Kinkead, A.N.

1967-02-01

In considering prestressed concrete vessels for use with helium cooled high temperature reactors, a number of new problems arise and projected designs involve new approaches and new solutions. These reactors, having high coolant outlet temperature from the core and relatively high power densities, can be built into compact designs which permit usefully high working pressures. Consequently, steam generators and circulating units tend to be small. Although circuit activity can be kept quite low with coated particle fuels, designs which involve entry for subsequent repair are not favoured, and coupled with the preferred aim of using fully shop fabricated units within the designs with removable steam generators which involve no tube welding inside the vessel. A particular solution uses a number of slim cylindrical assemblies housed in the wall of the pressure vessel and this vessel design concept is presented. The use of helium requires very high sealing standards and one of the important requirements is a vessel design which permits leak testing during construction, so that a repair seal can be made to any faulty part in a liner seam. Very good demountable joint seals can be made without particular difficulty and Dragon experience is used to provide solutions which are suitable for prestressed concrete vessel penetrations. The concept layout is given of a vessel meeting these requirements; the basis of design is outlined and special features of importance discussed. (author)

Neutronics studies for the design of the European DEMO vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Flammini, Davide, E-mail: davide.flammini@enea.it [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Villari, Rosaria; Moro, Fabio; Pizzuto, Aldo [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Bachmann, Christian [EUROfusion Consortium, Boltzmannstr. 2, 85748 Garching (Germany)

2016-11-01

Highlights: • MCNP calculation of nuclear heating, damage, helium production and neutron flux in DEMO HCLL and HCPB vacuum vessel at the inboard equatorial plane. • Study of impact of the poloidal gap between blanket modules, for several gap width, on vacuum vessel nuclear quantities. • Effect of the gap on nuclear heating result to be moderate, however high values of nuclear heating are found, even far from the gap with HCLL blanket. • Radiation damage limit of 2.75 DPA is met with a 1 cm wide gap. Helium production results very sensitive to the gap width. • Comparison between HCLL and HCPB blankets is shown for nuclear heating and neutron flux in the vacuum vessel. - Abstract: The DEMO vacuum vessel, a massive water cooled double-walled steel vessel, is located behind breeding blankets and manifolds and it will be subjected to an intense neutron and photon irradiation. Therefore, a proper evaluation of the vessel nuclear heat loads is required to assure adequate cooling and, given the significant lifetime neutron fluence of DEMO, the radiation damage limit of the vessel needs to be carefully controlled. In the present work nuclear heating, radiation damage (DPA), helium production, neutron and photon fluxes have been calculated on the vacuum vessel at the inboard by means of MCNP5 using a 3D Helium Cooled Lithium Lead (HCLL) DEMO model with 1572 MW of fusion power. In particular, the effect of the poloidal gap between the breeding-blanket segments on vacuum vessel nuclear loads has been estimated varying the gap width from 0 to 5 cm. High values of the nuclear heating (≈1 W/cm{sup 3}), which might cause intense thermal stresses, were obtained in inboard equatorial zone. The effect of the poloidal gap on the nuclear heating resulted to be moderate (within 30%). The radiation damage limit of 2.75 DPA on the vessel is almost met with 1 cm of poloidal gap over DEMO lifetime. A comparison with Helium Cooled Pebble Bed blanket is also provided.

In-service inspection robot for PFBR main vessel- concept

Energy Technology Data Exchange (ETDEWEB)

Rajendran, S; Ramakumar, M S [Bhabha Atomic Research Centre, Mumbai (India). Div. of Remote Handling and Robotics

1994-12-31

In-service inspection (ISI) of critical components in a nuclear reactor is one of the foremost and important tasks which reveals the state of health of the system, thereby ensuring the safety of the plant, personnel and environment. Prototype Fast Breeder Reactor (PFBR) is designed as a pool type reactor. A safety vessel is provided in the design which envelopes the main reactor vessel. The ISI of the main vessel is mandatory and will be carried out by a robot which will operate on this annular gap. The design of the robot is such that it can crawl around the vessel and into the gap at the bottom of the vessel relying on friction grip. The mobile robot will carry a CCTV camera and the inspection technique packages into the interspace, position and orient these to carry out the ISI of the main vessel. The paper discusses about the design features of the robot including the gripping mechanism and the crawling sequence to perform ISI of the reactor vessel. 3 figs.

In-service inspection robot for PFBR main vessel- concept

International Nuclear Information System (INIS)

Rajendran, S.; Ramakumar, M.S.

1994-01-01

In-service inspection (ISI) of critical components in a nuclear reactor is one of the foremost and important tasks which reveals the state of health of the system, thereby ensuring the safety of the plant, personnel and environment. Prototype Fast Breeder Reactor (PFBR) is designed as a pool type reactor. A safety vessel is provided in the design which envelopes the main reactor vessel. The ISI of the main vessel is mandatory and will be carried out by a robot which will operate on this annular gap. The design of the robot is such that it can crawl around the vessel and into the gap at the bottom of the vessel relying on friction grip. The mobile robot will carry a CCTV camera and the inspection technique packages into the interspace, position and orient these to carry out the ISI of the main vessel. The paper discusses about the design features of the robot including the gripping mechanism and the crawling sequence to perform ISI of the reactor vessel. 3 figs

Conceptual design of an in-vessel inspection robotic system for Tokamak environment

International Nuclear Information System (INIS)

Kumar, Prabhat; Raju, Daniel; Ranjan, Vaibhav; Patel, Prateek; Dave, Jatinkumar; Naik, Mehul

2013-01-01

An in-vessel inspection robotic system has been conceptualized for operation inside a tokamak vessel. The robotic system is envisaged to comprise of a robotic arm, end-effector, microcontroller and wireless communication system. The end-effector is envisaged to be a special purpose camera for in-situ inspection between plasma shots. The three-link robotic arm, designed for ITER-like environment, has 4 revolute joints- 3 providing manipulation in poloidal plane and the fourth one providing limited movement in adjacent toroidal planes. This paper provides the conceptual design of the system along with kinematic analysis of robotic arm. Solutions have been derived for forward and inverse kinematic models and the Jacobian matrix for the robotic arm linkage. In forward kinematic model, given a set of joint-link parameters, the position and orientation of end-effector are determined with respect to a reference frame. In inverse kinematic model, given the specified position and orientation of end-effector with respect to a reference frame, a set of joint variables are derived that would bring the end-effector into the required posture. Using Jacobian matrix, the relation between the end-effector velocity and the joint velocity of a manipulator is obtained i.e. given the individual joint velocity; the end-effector velocity is obtained. A CAD model has been generated using CATIA to simulate the kinematic model and carry out computational stress analysis. (author)

Enabling lean design of biomethane gas distribution grids

NARCIS (Netherlands)

Weidenaar, Teade; Jauregui Becker, Juan Manuel; Hoekstra, Sipke

2015-01-01

The Dutch gas distribution infrastructure faces several significant changes in the near future. One of the major changes is the production and injection of biomethane into the gas distribution grid. This article introduces a Design Synthesis Tool (DST) that automatically generates biomethane supply

Hinkley Point A gas duct repairs

International Nuclear Information System (INIS)

Curtis, R.F.

1996-01-01

In 1990, routine visual inspection of the Hinckley Point A Reactor 1 pressure vessel gas outlet ducts showed failures in the welded stud bolts retaining the insulation edging strips. Since the ducts are accessible only from within the pressure vessel, a remote repair technique that could be deployed via the vessel stand pipe had to be found. A drawn arc stud welding and work package formerly used at the Oldbury Power Station was modified for the purpose. The only manipulators with sufficient reach and adequate carrying capacity to deploy the package were the Sizewell SNAKES manipulators. One of these was modified to fit the Hinckley reactor and repairs have been successfully carried out. Similar studs on the gas ducts in Reactor 2, are shielded from visual inspection by a Z-clip feature. A technique using pulsed thermography was developed. The studs were heated for a short time at their exposed ends using a prefocused lamp and the heat decay patterns monitored by an infrared camera enabling attached and detached studs to be distinguished. The inspection package was deployed using the SNAKES manipulator again. In both operations, I-Grip computer modelling was used in the design of the package envelope and the deployment routes. (UK)

Burst pressure investigation of filament wound type IV composite pressure vessel

Science.gov (United States)

Farhood, Naseer H.; Karuppanan, Saravanan; Ya, H. H.; Baharom, Mohamad Ariff

2017-12-01

Currently, composite pressure vessels (PVs) are employed in many industries such as aerospace, transportations, medical etc. Basically, the use of PVs in automotive application as a compressed natural gas (CNG) storage cylinder has been growing rapidly. Burst failure due to the laminate failure is the most critical failure mechanism for composite pressure vessels. It is predominantly caused by excessive internal pressure due to an overfilling or an overheating. In order to reduce fabrication difficulties and increase the structural efficiency, researches and studies are conducted continuously towards the proper selection of vessel design parameters. Hence, this paper is focused on the prediction of first ply failure pressure for such vessels utilizing finite element simulation based on Tsai-Wu and maximum stress failure criterions. The effects of laminate stacking sequence and orientation angle on the burst pressure were investigated in this work for a constant layered thickness PV. Two types of winding design, A [90°2/∓θ16/90°2] and B [90°2/∓θ]ns with different orientations of helical winding reinforcement were analyzed for carbon/epoxy composite material. It was found that laminate A sustained a maximum burst pressure of 55 MPa for a sequence of [90°2/∓15°16/90°2] while the laminate B returned a maximum burst pressure of 45 MPa corresponding to a stacking sequence of [90°2/±15°/90°2/±15°/90°2/±15° ....] up to 20 layers for a constant vessel thickness. For verification, a comparison was done with the literature under similar conditions of analysis and good agreement was achieved with a maximum difference of 4% and 10% for symmetrical and unsymmetrical layout, respectively.

Gas Generation from K East Basin Sludges and Irradiated Metallic Uranium Fuel Particles Series III Testing

International Nuclear Information System (INIS)

Schmidt, Andrew J.; Delegard, Calvin H.; Bryan, Samuel A.; Elmore, Monte R.; Sell, Rachel L.; Silvers, Kurt L.; Gano, Susan R.; Thornton, Brenda M.

2003-01-01

The path forward for managing of Hanford K Basin sludge calls for it to be packaged, shipped, and stored at T Plant until final processing at a future date. An important consideration for the design and cost of retrieval, transportation, and storage systems is the potential for heat and gas generation through oxidation reactions between uranium metal and water. This report, the third in a series (Series III), describes work performed at the Pacific Northwest National Laboratory (PNNL) to assess corrosion and gas generation from irradiated metallic uranium particles (fuel particles) with and without K Basin sludge addition. The testing described in this report consisted of 12 tests. In 10 of the tests, 4.3 to 26.4 g of fuel particles of selected size distribution were placed into 60- or 800-ml reaction vessels with 0 to 100 g settled sludge. In another test, a single 3.72-g fuel fragment (i.e., 7150-mm particle) was placed in a 60 ml reaction vessel with no added sludge. The twelfth test contained only sludge. The fuel particles were prepared by crushing archived coupons (samples) from an irradiated metallic uranium fuel element. After loading the sludge materials (whether fuel particles, mixtures of fuel particles and sludge, or sludge-only) into reaction vessels, the solids were covered with an excess of K Basin water, the vessels closed and connected to a gas measurement manifold, and the vessels back-flushed with inert neon cover gas. The vessels were then heated to a constant temperature. The gas pressures and temperatures were monitored continuously from the times the vessels were purged. Gas samples were collected at various times during the tests, and the samples analyzed by mass spectrometry. Data on the reaction rates of uranium metal fuel particles with water as a function of temperature and particle size were generated. The data were compared with published studies on metallic uranium corrosion kinetics. The effects of an intimate overlying sludge layer

QFD-ANP Approach for the Conceptual Design of Research Vessels: A Case Study

Science.gov (United States)

Venkata Subbaiah, Kambagowni; Yeshwanth Sai, Koneru; Suresh, Challa

2016-10-01

Conceptual design is a subset of concept art wherein a new idea of product is created instead of a visual representation which would directly be used in a final product. The purpose is to understand the needs of conceptual design which are being used in engineering designs and to clarify the current conceptual design practice. Quality function deployment (QFD) is a customer oriented design approach for developing new or improved products and services to enhance customer satisfaction. House of quality (HOQ) has been traditionally used as planning tool of QFD which translates customer requirements (CRs) into design requirements (DRs). Factor analysis is carried out in order to reduce the CR portions of HOQ. The analytical hierarchical process is employed to obtain the priority ratings of CR's which are used in constructing HOQ. This paper mainly discusses about the conceptual design of an oceanographic research vessel using analytical network process (ANP) technique. Finally the QFD-ANP integrated methodology helps to establish the importance ratings of DRs.

Injector design for liner-on-target gas-puff experiments

Science.gov (United States)

Valenzuela, J. C.; Krasheninnikov, I.; Conti, F.; Wessel, F.; Fadeev, V.; Narkis, J.; Ross, M. P.; Rahman, H. U.; Ruskov, E.; Beg, F. N.

2017-11-01

We present the design of a gas-puff injector for liner-on-target experiments. The injector is composed of an annular high atomic number (e.g., Ar and Kr) gas and an on-axis plasma gun that delivers an ionized deuterium target. The annular supersonic nozzle injector has been studied using Computational Fluid Dynamics (CFD) simulations to produce a highly collimated (M > 5), ˜1 cm radius gas profile that satisfies the theoretical requirement for best performance on ˜1-MA current generators. The CFD simulations allowed us to study output density profiles as a function of the nozzle shape, gas pressure, and gas composition. We have performed line-integrated density measurements using a continuous wave (CW) He-Ne laser to characterize the liner gas density. The measurements agree well with the CFD values. We have used a simple snowplow model to study the plasma sheath acceleration in a coaxial plasma gun to help us properly design the target injector.

Functional design criteria for the retained gas sampler system

International Nuclear Information System (INIS)

Wootan, D.W.

1995-01-01

A Retained Gas Sampler System (RGSS) is being developed to capture and analyze waste samples from Hanford Flammable Gas Watch List Tanks to determine both the quantity and composition of gases retained in the waste. The RGSS consists of three main components: the Sampler, Extractor, and Extruder. This report describes the functional criteria for the design of the RGSS components. The RGSS Sampler is based on the WHC Universal Sampler design with modifications to eliminate gas leakage. The primary function of the Sampler is to capture a representative waste sample from a tank and transport the sample with minimal loss of gas content from the tank to the laboratory. The function of the Extruder is to transfer the waste sample from the Sampler to the Extractor. The function of the Extractor is to separate the gases from the liquids and solids, measure the relative volume of gas to determine the void fraction, and remove and analyze the gas constituents

THE STIRLING GAS REFRIGERATING MACHINE MECHANICAL DESIGN IMPROVING

Directory of Open Access Journals (Sweden)

V. V. Trandafilov

2016-06-01

Full Text Available To improve the mechanical design of the piston Stirling gas refrigeration machine the structural optimization of rotary vane Stirling gas refrigeration machine is carried out. This paper presents the results of theoretical research. Analysis and prospects of rotary vane Stirling gas refrigeration machine for domestic and industrial refrigeration purpose are represented. The results of a patent search by mechanisms of transformation of rotary vane machines are discussed.

THE STIRLING GAS REFRIGERATING MACHINE MECHANICAL DESIGN IMPROVING

Directory of Open Access Journals (Sweden)

V. V. Trandafilov

2016-02-01

Full Text Available To improve the mechanical design of the piston Stirling gas refrigeration machine the structural optimization of rotary vane Stirling gas refrigeration machine is carried out. This paper presents the results of theoretical research. Analysis and prospects of rotary vane Stirling gas refrigeration machine for domestic and industrial refrigeration purpose are represented. The results of a patent search by mechanisms of transformation of rotary vane machines are discussed

Integrating Multiple Autonomous Underwater Vessels, Surface Vessels and Aircraft into Oceanographic Research Vessel Operations

Science.gov (United States)

McGillivary, P. A.; Borges de Sousa, J.; Martins, R.; Rajan, K.

2012-12-01

use of UAS on oceanographic research vessels is just beginning. We report on several initial field efforts which demonstrated that UAS improve spatial and temporal mapping of ocean features, as well as monitoring marine mammal populations, ocean color, sea ice and wave fields and air-sea gas exchange. These studies however also confirm the challenges for shipboard computer systems ingesting and archiving UAS high resolution video, SAR and lidar data. We describe the successful inclusion of DTN communications for: 1) passing video data between two UAS or a UAS and ship; 2) for inclusion of ASVs as communication nodes for AUVs; as well as, 3) enabling extension of adaptive sampling software from AUVs and ASVs to include UAS. In conclusion, we describe how autonomous sampling systems may be best integrated into shipboard oceanographic vessel research to provide new and more comprehensive time-space ocean and atmospheric data collection that is important not only for scientific study, but also for sustainable ocean management, including emergency response capabilities. The recent examples of such integrated studies highlighted confirm ocean and atmospheric studies can more cost-effectively pursued, and in some cases only accomplished, by combining underwater, surface and aircraft autonomous systems with research vessel operations.

Development of KALIMER auxiliary sodium and cover gas management system

International Nuclear Information System (INIS)

Kwon, Sang Woon; Hwang, Sung Tae

1996-11-01

The objectives of this report are to develop and to describe the auxiliary liquid metal and cover gas management systems of KALIMER. the system includes following system: (1) Auxiliary liquid metal system (2) Inert gas receiving and processing system (3) Impurity monitoring and analysis system. Auxiliary liquid metal and cover gas management system of KALIMER was developed. Functions of each systems and design basis were describes. The auxiliary liquid metal system receives, transfers, and purifies all sodium used in the plant. The system furnishes the required sodium quantity at the pressure, temperature, flow rate, and purity specified by the interfacing system. The intermediated sodium processing subsystem (ISPS) provides continuous purification of IHTS sodium, as well as performs the initial fill operation for both the IHTS and reactor vessel. The primary sodium processing subsystem provides purification (cold trapping) for sodium used in the reactor vessel. The inert gas receiving and processing (IGRP) system provides liquefied and ambient gas storage, delivers inert gases of specified composition and purity at regulated flow rates and pressures to points of usage throughout the KALIMER, and accepts the contaminated gases through its vacuum facilities for storage and transfer to the gas radwaste system. Three gases are used in the KALIMER: helium, argon, and nitrogen. 11 tabs., 12 figs. (Author)

Development of KALIMER auxiliary sodium and cover gas management system

Energy Technology Data Exchange (ETDEWEB)

Kwon, Sang Woon; Hwang, Sung Tae [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-11-01

The objectives of this report are to develop and to describe the auxiliary liquid metal and cover gas management systems of KALIMER. the system includes following system: (1) Auxiliary liquid metal system (2) Inert gas receiving and processing system (3) Impurity monitoring and analysis system. Auxiliary liquid metal and cover gas management system of KALIMER was developed. Functions of each systems and design basis were describes. The auxiliary liquid metal system receives, transfers, and purifies all sodium used in the plant. The system furnishes the required sodium quantity at the pressure, temperature, flow rate, and purity specified by the interfacing system. The intermediated sodium processing subsystem (ISPS) provides continuous purification of IHTS sodium, as well as performs the initial fill operation for both the IHTS and reactor vessel. The primary sodium processing subsystem provides purification (cold trapping) for sodium used in the reactor vessel. The inert gas receiving and processing (IGRP) system provides liquefied and ambient gas storage, delivers inert gases of specified composition and purity at regulated flow rates and pressures to points of usage throughout the KALIMER, and accepts the contaminated gases through its vacuum facilities for storage and transfer to the gas radwaste system. Three gases are used in the KALIMER: helium, argon, and nitrogen. 11 tabs., 12 figs. (Author).

Progress of ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Ioki, K., E-mail: Kimihiro.Ioki@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Bayon, A. [F4E, c/ Josep Pla, No. 2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Choi, C.H.; Daly, E.; Dani, S.; Davis, J.; Giraud, B.; Gribov, Y.; Hamlyn-Harris, C.; Jun, C.; Levesy, B. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Kim, B.C. [NFRI, 52 Yeoeundong Yuseonggu, Daejeon 305-333 (Korea, Republic of); Kuzmin, E. [NTC “Sintez”, Efremov Inst., 189631 Metallostroy, St. Petersburg (Russian Federation); Le Barbier, R.; Martinez, J.-M. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Pathak, H. [ITER-India, A-29, GIDC Electronic Estate, Sector 25, Gandhinagar 382025 (India); Preble, J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Sa, J.W. [NFRI, 52 Yeoeundong Yuseonggu, Daejeon 305-333 (Korea, Republic of); Terasawa, A.; Utin, Yu. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); and others

2013-10-15

Highlights: ► This covers the overall status and progress of the ITER vacuum vessel activities. ► It includes design, R and D, manufacturing and approval process of the regulators. ► The baseline design was completed and now manufacturing designs are on-going. ► R and D includes ISI, dynamic test of keys and lip-seal welding/cutting technology. ► The VV suppliers produced full-scale mock-ups and started VV manufacturing. -- Abstract: Design modifications were implemented in the vacuum vessel (VV) baseline design in 2011–2012 for finalization. The modifications are mostly due to interface components, such as support rails and feedthroughs for the in-vessel coils (IVC). Manufacturing designs are being developed at the domestic agencies (DAs) based on the baseline design. The VV support design was also finalized and tests on scale mock-ups are under preparation. Design of the in-wall shielding (IWS) has progressed, considering the assembly methods and the required tolerances. Further modifications are required to be consistent with the DAs’ manufacturing designs. Dynamic tests on the inter-modular and stub keys to support the blanket modules are being performed to measure the dynamic amplification factor (DAF). An in-service inspection (ISI) plan has been developed and R and D was launched for ISI. Conceptual design of the VV instrumentation has been developed. The VV baseline design was approved by the agreed notified body (ANB) in accordance with the French Nuclear Pressure Equipment Order procedure.

Progress of ITER vacuum vessel

International Nuclear Information System (INIS)

Ioki, K.; Bayon, A.; Choi, C.H.; Daly, E.; Dani, S.; Davis, J.; Giraud, B.; Gribov, Y.; Hamlyn-Harris, C.; Jun, C.; Levesy, B.; Kim, B.C.; Kuzmin, E.; Le Barbier, R.; Martinez, J.-M.; Pathak, H.; Preble, J.; Sa, J.W.; Terasawa, A.; Utin, Yu.

2013-01-01

Highlights: ► This covers the overall status and progress of the ITER vacuum vessel activities. ► It includes design, R and D, manufacturing and approval process of the regulators. ► The baseline design was completed and now manufacturing designs are on-going. ► R and D includes ISI, dynamic test of keys and lip-seal welding/cutting technology. ► The VV suppliers produced full-scale mock-ups and started VV manufacturing. -- Abstract: Design modifications were implemented in the vacuum vessel (VV) baseline design in 2011–2012 for finalization. The modifications are mostly due to interface components, such as support rails and feedthroughs for the in-vessel coils (IVC). Manufacturing designs are being developed at the domestic agencies (DAs) based on the baseline design. The VV support design was also finalized and tests on scale mock-ups are under preparation. Design of the in-wall shielding (IWS) has progressed, considering the assembly methods and the required tolerances. Further modifications are required to be consistent with the DAs’ manufacturing designs. Dynamic tests on the inter-modular and stub keys to support the blanket modules are being performed to measure the dynamic amplification factor (DAF). An in-service inspection (ISI) plan has been developed and R and D was launched for ISI. Conceptual design of the VV instrumentation has been developed. The VV baseline design was approved by the agreed notified body (ANB) in accordance with the French Nuclear Pressure Equipment Order procedure

Hydrogen/hydrocarbon explosions in the ITER vacuum vessel

International Nuclear Information System (INIS)

Goranson, P.L.

1992-01-01

The consequences of H 2 /hydrocarbon detonations in the vacuum vessel (torus) of the International Thermonuclear Experimental Reactor (ITER) have been studied. The most likely scenario for such a detonation involves a water leak into the torus and a vent of the torus to atmosphere, permitting the formation of an explosive fuel-air mixture. The generation of fuel gases and possible sources of air or oxygen are reviewed, and the severity and effects of specific fuel-air mixture explosions are evaluated. Detonation or deflagration of an explosive mixture could result in pressures exceeding the maximum allowable torus pressure. Further studies to examine the design details and develop an event-tree study of events following a gas detonation are recommended

Application of Module System for Processing a Large Capacity of Coal Steam Power Plant Flue gas by Electron Beam Machine

International Nuclear Information System (INIS)

Rukijatmo; Munawir Z, M.

2003-01-01

Conceptual design of SOx dan NOx flue gas treatment base on 25% of 400 M We capacity and 90% efficiency reduction of SOx, the electron beam machine will be utilized to performed the environment quality standard of air pollution. The technical specification of electron beam machine, processing system and chamber dimension should conformed to the regulation. The discussion is focused on the selection of electron beam machine type and the dimension of radiation vessel for perfect reaction and exact time processing. The design calculation is indicated that we need two electron beam machines of 500 mA, 800 kV installed in parallel and 3 up to 3.4 metres diameter, the speed of flue gas in the vessel around 16.4 up to 18.14 metre per second, 80% treatment of 0,7% sulphur content coal is conform to regulation on emission of flue gas environment, and only 50% of flue gas needed to be treated by 4 modular. (author)

Natural gas fuelling stations installation code

Energy Technology Data Exchange (ETDEWEB)

Barrigar, C; Burford, G; Adragna, M; Hawryn, S

2004-07-01

This Canadian Standard applies to natural gas fuelling stations that can be used for fleet and public dispensing operations. This document is divided into 11 sections that address the scope of the Standard; definitions and reference publications; general requirements; compressors; storage; dispensing; flow control devices; storage vessel dispatch and receiving; design, installation and testing of piping, tubing and fittings; and installation of vehicle refuelling appliances (VRAs) connected to storage piping. The most recent revision to the Standard includes requirements for indoor fuelling of natural gas vehicles. This Standard, like all Canadian Standards, was subject to periodic review and was most recently reaffirmed in 2004. tabs., figs.

Preconceptual design of the gas-phase decontamination demonstration cart

International Nuclear Information System (INIS)

Munday, E.B.

1993-12-01

Removal of uranium deposits from the interior surfaces of gaseous diffusion equipment will be a major portion of the overall multibillion dollar effort to decontaminate and decommission the gaseous diffusion plants. Long-term low-temperature (LTLT) gas-phase decontamination is being developed at the K-25 Site as an in situ decontamination process that is expected to significantly lower the decontamination costs, reduce worker exposure to radioactive materials, and reduce safeguard concerns. This report documents the preconceptual design of the process equipment that is necessary to conduct a full-scale demonstration of the LTLT method in accordance with the process steps listed above. The process equipment and method proposed in this report are not intended to represent a full-scale production campaign design and operation, since the gas evacuation, gas charging, and off-gas handling systems that would be cost effective in a production campaign are not cost effective for a first-time demonstration. However, the design presented here is expected to be applicable to special decontamination projects beyond the demonstration, which could include the Deposit Recovery Program. The equipment will therefore be sized to a 200 ft size 1 converter (plus a substantial conservative design margin), which is the largest item of interest for gas phase decontamination in the Deposit Recovery Program. The decontamination equipment will allow recovery of the UF 6 , which is generated from the reaction of ClF 3 with the uranium deposits, by use of NaF traps

Design and analysis of the vacuum vessel for RTO/RC-ITER

International Nuclear Information System (INIS)

Onozuka, M.; Ioki, K.; Johnson, G.; Kodama, T.; Sannazzaro, G.; Utin, Y.

2000-01-01

Recent progress in design and analysis of the vacuum vessel (VV) for the reduced technical objectives/reduced cost International Thermonuclear Experimental Reactor (RTO/RC-ITER) is presented. The basic VV design is similar to the previous ITER VV. However, because the back plate for the blanket modules could be eliminated, its previous functions could be transferred to the VV. For this option, the blanket modules are supported directly by the VV and the blanket coolant channels are structurally part of the VV double wall structure. In addition, a 'tight fitting' configuration is required to correctly position the modules' first wall. Although such modifications of the VV complicate its structure and increase its fabrication cost, the design of the VV is considered to be still feasible. The structural analyses of the VV have been conducted using several FE models of the VV, including global and local models. Although further assessment is required, based on the analyses performed to date, the structural aspects of the VV for the case without the back plate appear feasible

Design and analysis of the vacuum vessel for RTO/RC-ITER

Energy Technology Data Exchange (ETDEWEB)

Onozuka, M. E-mail: onozukm@itereu.de; Ioki, K.; Johnson, G.; Kodama, T.; Sannazzaro, G.; Utin, Y

2000-11-01

Recent progress in design and analysis of the vacuum vessel (VV) for the reduced technical objectives/reduced cost International Thermonuclear Experimental Reactor (RTO/RC-ITER) is presented. The basic VV design is similar to the previous ITER VV. However, because the back plate for the blanket modules could be eliminated, its previous functions could be transferred to the VV. For this option, the blanket modules are supported directly by the VV and the blanket coolant channels are structurally part of the VV double wall structure. In addition, a 'tight fitting' configuration is required to correctly position the modules' first wall. Although such modifications of the VV complicate its structure and increase its fabrication cost, the design of the VV is considered to be still feasible. The structural analyses of the VV have been conducted using several FE models of the VV, including global and local models. Although further assessment is required, based on the analyses performed to date, the structural aspects of the VV for the case without the back plate appear feasible.

Pressure vessel for nuclear reactors

International Nuclear Information System (INIS)

1975-01-01

The invention applies to a pressure vessel for nuclear reactors whose shell, made of cast metal segments, has a steel liner. This liner must be constructed to withstand all operational stresses and to be easily repairable. The invention solves this problem by installing the liner at a certain distance from the inner wall of the pressure vessel shell and by filling this clearance with supporting concrete. Both the concrete and the steel liner must have a lower prestress than the pressure vessel shell. In order to avoid damage to the liner when prestressing the pressure vessel shell, special connecting elements are provided which consist of welded-on fastening elements projecting into recesses in the cast metal segments of the pressure vessel. Their design is described in detail. (TK) [de

A continuum damage analysis of hydrogen attack in 2.25 Cr-1Mo vessel

DEFF Research Database (Denmark)

van der Burg, M.W.D.; van der Giessen, E.; Tvergaard, Viggo

1998-01-01

A micromechanically based continuum damage model is presented to analyze the stress, temperature and hydrogen pressure dependent material degradation process termed hydrogen attack, inside a pressure vessel. Hydrogen attack (HA) is the damage process of grain boundary facets due to a chemical...... reaction of carbides with hydrogen, thus forming cavities with high pressure methane gas. Driven by the methane gas pressure, the cavities grow, while remote tensile stresses can significantly enhance the cavitation rate. The damage model gives the strain-rate and damage rate as a function...... of the temperature, hydrogen pressure and applied stresses. The model is applied to study HA in a vessel wall, where nonuniform distributions of hydrogen pressure, temperature and stresses result in a nonuniform damage distribution over the vessel wall. Stresses inside the vessel wall first tend to accelerate...

Experiences in control system design aided by interactive computer programs: temperature control of the laser isotope separation vessel

International Nuclear Information System (INIS)

Gavel, D.T.; Pittenger, L.C.; McDonald, J.S.; Cramer, P.G.; Herget, C.J.

1985-01-01

A robust control system has been designed to regulate temperature in a vacuum vessel. The thermodynamic process is modeled by a set of nonlinear, implicit differential equations. The control design and analysis task exercised many of the computer-aided control systems design software packages, including MATLAB, DELIGHT, and LSAP. The working environment is a VAX computer. Advantages and limitations of the software and environment, and the impact on final controller design is discussed

Experiences in control system design aided by interactive computer programs: Temperature control of the laser isotope separation vessel

Science.gov (United States)

Gavel, D. T.; Pittenger, L. C.; McDonald, J. S.; Cramer, P. G.; Herget, C. J.

A robust control system has been designed to regulate temperature in a vacuum vessel. The thermodynamic process is modeled by a set of nonlinear, implicit differential equations. The control design and analysis task exercised many of the computer-aided control systems design software packages, including MATLAB, DELIGHT, AND LSAP. The working environment is a VAX computer. Advantages and limitations of the software and environment, and the impact on final controller design is discussed.

Nuclear powerplant with closed gas-cooling circuit

International Nuclear Information System (INIS)

Haferkamp, D.; Hodzic, A.; Winter, U.

1976-01-01

Disclosed is a nuclear power plant comprising a pressure-tight safety vessel surrounding the entire plant, an inner vessel of reinforced concrete, a high-temperature reactor contained in the inner vessel, a gas turbine assembly having a turbine and a high- and low-pressure compressor located in a horizontally oriented chamber below the reactor, a plurality of heat exchange units positioned in a plurality of vertically oriented pods spaced radially symmetrically about the reactor and suitable conduits for carrying the reactive gas between the system components. The conduits are arranged in generally horizontally and vertically oriented straight lines, and the conduits for carrying low-pressure gas comprise a horizontal system positioned beneath the turbine assembly having a plurality of coaxial connecting tubes, collectors and distributors as well as normal conduits, so that high pressure gas flows in the internal passage and low-pressure gas flows in the outer passage. 22 claims, 7 figures

Design Aspects of Wet Scrubber System

International Nuclear Information System (INIS)

Lee, Hyun-Chul; Bang, Young-suk; Jung, Woo-Young; Lee, Doo-Yong

2015-01-01

The water pool in the wet scrubber system has advantage to cope with decay heat based on the thermal hydraulic balance such as condensation and evaporation inside it. This study focuses on the design aspects of the wet scrubber system to estimate the required water pool mass during the mission time and size of the scrubbing tank including inner structures. The design of the wet scrubber system include the estimation of the required water mass during the mission time and sizing of the scrubber vessel to contain the water pool. The condensation due to the inlet steam and evaporation due to the steam and non-condensable gas superheat and decay heat from filtered fission products should be considered to estimate the water mass required to maintain its function during the mission time. On the other hand, the level swelling due to the noncondensable gas is another important design aspect on the sizing of the scrubber vessel and determination of the entry elevation of the filtration components such as the droplet separator or filter. The minimum water level based on the minimum collapsed water level should be higher than the exit of scrubber nozzle

Design Aspects of Wet Scrubber System

Energy Technology Data Exchange (ETDEWEB)

Lee, Hyun-Chul; Bang, Young-suk; Jung, Woo-Young; Lee, Doo-Yong [FNC Technology Co. Ltd., Yongin (Korea, Republic of)

2015-10-15

The water pool in the wet scrubber system has advantage to cope with decay heat based on the thermal hydraulic balance such as condensation and evaporation inside it. This study focuses on the design aspects of the wet scrubber system to estimate the required water pool mass during the mission time and size of the scrubbing tank including inner structures. The design of the wet scrubber system include the estimation of the required water mass during the mission time and sizing of the scrubber vessel to contain the water pool. The condensation due to the inlet steam and evaporation due to the steam and non-condensable gas superheat and decay heat from filtered fission products should be considered to estimate the water mass required to maintain its function during the mission time. On the other hand, the level swelling due to the noncondensable gas is another important design aspect on the sizing of the scrubber vessel and determination of the entry elevation of the filtration components such as the droplet separator or filter. The minimum water level based on the minimum collapsed water level should be higher than the exit of scrubber nozzle.

General design and main problems of a gas-heavy-water power reactor contained in a pressure vessel; Conception generale et principaux problemes d'un reacteur de puissance eau lourde-gaz contenu dans un caisson resistant

Energy Technology Data Exchange (ETDEWEB)

Roche, R.; Gaudez, J.C. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

In the framework of research carried out on a CO{sub 2}-cooled power reactor moderated by heavy water, the so-called 'pressure vessel' solution involves the total integration of the core, of the primary circuit (exchanges and blowers) and of the fuel handling machine inside a single, strong, sealed vessel made of pre-stressed concrete. A vertical design has been chosen: the handling 'attic' is placed above the core, the exchanges being underneath. This solution makes it possible to standardize the type of reactor which is moderated by heavy-water or graphite and cooled by a downward stream of carbon dioxide gas; it has certain advantages and disadvantages with respect to the pressure tube solution and these are considered in detail in this report. Extrapolation presents in particular.problems due specifically to the heavy water (for example its cooling,its purification, the balancing of the pressures of the heavy water and of the gas, the assembling of the internal structures, the height of the attic, etc. (authors) [French] Dans le cadre des etudes d'un reacteur de puissance modere a l'eau lourde et refroidi-au gaz carbonique, la solution dite 'en caisson' consiste en une integration totale du coeur, du circuit primaire (echangeurs et soufflantes) et du dispositif de manutention du combustible a l'interieur d'un meme caisson etanche et resistant en beton precontraint. La disposition envisagee est verticale; le grenier de manutention est dispose au-dessus du coeur, les echangeurs en dessous. Cette solution, qui permet d'uniformiser les types de reacteurs moderes a l'eau lourde ou au graphite et refroidis par une circulation descendante de gaz carbonique presente, par rapport a la solution a tube de force, des avantages et des inconvenients qui sont analyses dans cette etude. L'extrapolation pose, en particulier, des problemes specifiques a l'eau lourde (tels que son refroidissement, son epuration

Component design considerations for gas turbine HTGR waste-heat power plant

International Nuclear Information System (INIS)

McDonald, C.F.; Vrable, D.L.

1976-01-01

Component design considerations are described for the ammonia waste-heat power conversion system of a large helium gas-turbine nuclear power plant under development by General Atomic Company. Initial component design work was done for a reference plant with a 3000-MW(t) High-Temperature Gas-Cooled Reactor (HTGR), and this is discussed. Advanced designs now being evaluated include higher core outlet temperature, higher peak system pressures, improved loop configurations, and twin 4000-MW(t) reactor units. Presented are the design considerations of the major components (turbine, condenser, heat input exchanger, and pump) for a supercritical ammonia Rankine waste heat power plant. The combined cycle (nuclear gas turbine and waste-heated plant) has a projected net plant efficiency of over 50 percent. While specifically directed towards a nuclear closed-cycle helium gas-turbine power plant (GT-HTGR), it is postulated that the bottoming waste-heat cycle component design considerations presented could apply to other low-grade-temperature power conversion systems such as geothermal plants

46 CFR 90.10-16 - Industrial vessel.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Industrial vessel. 90.10-16 Section 90.10-16 Shipping... PROVISIONS Definition of Terms Used in This Subchapter § 90.10-16 Industrial vessel. This term means every vessel which by reason of its special outfit, purpose, design, or function engages in certain industrial...

Littoral Combat Vessels: Analysis and Comparison of Designs

National Research Council Canada - National Science Library

Christiansen, Bryan J

2008-01-01

.... The candidates are a Littoral Combat Ship with a surface warfare module, a National Security Cutter augmented with offensive and defensive weaponry, a "Sea Lance" inshore combat vessel, and a Combat...

X-ray phase contrast with injected gas for tumor microangiography

International Nuclear Information System (INIS)

Lundström, U; Larsson, D H; Burvall, A; Hertz, H M; Westermark, U K; Henriksson, M Arsenian

2014-01-01

We show that the microvasculature of mouse tumors can be visualized using propagation-based phase-contrast x-ray imaging with gas as the contrast agent. The large density difference over the gas–tissue interface provides high contrast, allowing the imaging of small-diameter blood vessels with relatively short exposure times and low dose using a compact liquid-metal-jet x-ray source. The method investigated is applied to tumors (E1A/Ras-transformed mouse embryonic fibroblasts) grown in mouse ears, demonstrating sub-15-µm-diameter imaging of their blood vessels. The exposure time for a 2D projection image is a few seconds and a full tomographic 3D map takes some minutes. The method relies on the strength of the vasculature to withstand the gas pressure. Given that tumor vessels are known to be more fragile than normal vessels, we investigate the tolerance of the vasculature of 12 tumors to gas injection and find that a majority withstand 200 mbar pressures, enough to fill 12-µm-diameter vessels with gas. A comparison of the elasticity of tumorous and non-tumorous vessels supports the assumption of tumor vessels being more fragile. Finally, we conclude that the method has the potential to be extended to the imaging of 15 µm vessels in thick tissue, including mouse imaging, making it of interest for, e.g., angiogenesis research. (paper)

Simulation of gas turbines operating in off-design condition

Energy Technology Data Exchange (ETDEWEB)

Walter, Arnaldo [Universidade Estadual de Campinas, SP (Brazil). Faculdade de Engenharia Mecanica. Dept. de Energia]. E-mail: walter@fem.unicamp.br

2000-07-01

In many countries thermal power plants based on gas turbines have been the main option for new investment into the electric system due to their relatively high efficiency and low capital cost. Cogeneration systems based on gas turbines have also been an important option for the electric industry. Feasibility studies of power plants based on gas turbine should consider the effect of atmospheric conditions and part-load operation on the machine performance. Doing this, an off-design procedure is required. A G T off-design simulation procedure is described in this paper. Ruston R M was used to validate the simulation procedure that, general sense, presents deviations lower than 2.5% in comparison to manufacturer's data. (author)

Design considerations for a gas microcontroller

Science.gov (United States)

Ritter, D. A.

1986-01-01

Some of the design problems that are now being addressed in consideration of a microcontroller for the upcoming GAS payload are discussed. Microcontrollers will be used to run the experiments and to collect and store the data from those experiments. Some of the requirements for a microcontroller are to be small, lightweight, have low power consumption, and high reliability. Some of the solutions that were developed to meet these design requirements are discussed. At present, the experiment is still in the design stage and the final design may change from what is presented here. The search for new integrated circuits which will do all that is needed all in one package continues.

POTENTIAL FOR HYDROGEN BUILDUP IN HANFORD SEALED AIR FILLED NUCLEAR STORAGE VESSELS

International Nuclear Information System (INIS)

HEY BE

2008-01-01

This calculation is performed in accordance with HNF-PRO-8259, PHMC Calculation Preparation and Issue and addresses the question as to whether a flammable mixture of hydrogen gas can accumulate in a Hanford sealed nuclear storage vessel where the only source of hydrogen is the moisture in the air that initially filled the vessel Of specific concern is nuclear fuel inside IDENT 69-Gs, placed in Core Component Containers (CCCs) located inside Interim Storage Vaults (ISVs) at the Plutonium Finishing Plant (PFP) The CCCs are to be removed from the ISVs and placed inside a Hanford Unirradiated Fuel Package (HUFP) for transport and interim storage. The repackaging procedures mandated that no plastics were permitted, all labels and tape were to be removed and the pins to be clean and inspected Loading of the fuel into the CCC/ISV package was permitted only if it was not raining or snowing. This was to preclude the introduction of any water The purpose was to minimize the presence of any hydrogenous material inside the storage vessels. The scope of NFPA 69, 'Standard on Explosion Prevention Systems', precludes its applicability for this case. The reactor fuel pins are helium bonded. The non-fuel pins, such as the pellet stacks, are also helium bonded. The fuel pellets were sintered at temperatures that preclude any residual hydrogenous material. Hydrogen gas can be formed from neutron and gamma radiolysis of water vapor. The radiolysis reaction is quite complex involving several intermediate radicals, and competing recombination reactions. Hydrogen gas can also be formed through corrosion. This analysis takes a simplistic approach and assumes that all water vapor present in the storage vessel is decomposed into hydrogen gas. Although the analysis is needed to specifically address HUFP storage of nuclear fuel, it is equally applicable to any sealed fuel storage vessel under the assumptions listed

LECOTELO - conceptual design, testings and realisation

International Nuclear Information System (INIS)

Ioan, M.; Hororoi, M.; Gutue, A.; Tudor, A.; Nistor, D.; Lebu, V.; Catana, A.; Ghita, G.; Pauna, E.; Cojocaru, V.; Tencu, V.

2013-01-01

A synthesis of all Computer Aid Design (CAD) and Computational Fluid Dynamics (CFD) works for LEad COrrosion TEsting LOop (LECOTELO) facility is presented. This facility was conceived to assure all conditions requested by corrosion/erosion tests in pure hot lead for different materials of interest for Lead cooled Fast Reactor (LFR). The main vessel will be able to receive at least 36 material samples; each of them must be swept on both sides by a lead flow at a very well known speed. The main circuit is composed by the following components: the centrifugal pump, flow-meters, heat exchanger, pre-heater, testing vessel, melting vessel, gas system, void system, electrical heaters, valves, etc. The main circuit has a capacity of 12 litters (0,012 m3). (authors)

The effects of fission gas release on PWR fuel rod design and performance

International Nuclear Information System (INIS)

Leech, W.J.; Kaiser, R.S.

1980-01-01

The purpose of this investigation was to determine the effects of fission gas release on PWR fuel rod design and performance. Empirical models were developed from fission gas release data. Fission gas release during normal operation is a function of burnup. There is little additional fission gas release during anticipated transients. The empirical models were used to evaluate Westinghouse fuel rod designs. It was determined that fission gas release is not a limiting parameter for obtaining rod average burnups in the range of 50,000 to 60,000 MWD/MTU. Fission gas release during anticipated transients has a negligible effect on the margins to rod design limits. (author)

Optimal Fuzzy and Dynamics Design of Ecological Sandwich Panel Vessel Roofs

Directory of Open Access Journals (Sweden)

Heikki Martikka

2011-01-01

Full Text Available In this study the basic engineering principles, goals, and constraints are all combined with fuzzy methodology and applied to optimally design sandwich panel circular plate roofs for large vessels loaded statically and dynamically. These panels are made up of two stiff, strong veneer skins separated by vertical and peripheral stiffener plates. Advantages are high strength, lightweight, and sustainability. In the present approach, first the goals and constraints of the end user are identified and expressed as decision variables which are formulated using the engineering variables for materials, geometry, and function. Then same consistent fuzzy satisfaction functions are formed over the desired ranges to suit the customer's desires. The risk of extreme dynamic loadings exciting resonance is studied by natural frequency and mode analysis by FEM and analytical models. The results show the most critical locations and give guidelines for innovative remedies of the concept before detailed FEM analyses to finalize the design.

Electrical swing adsorption gas storage and delivery system

Science.gov (United States)

Judkins, Roddie R.; Burchell, Timothy D.

1999-01-01

Systems and methods for electrical swing natural gas adsorption are described. An apparatus includes a pressure vessel; an electrically conductive gas adsorptive material located within the pressure vessel; and an electric power supply electrically connected to said adsorptive material. The adsorptive material can be a carbon fiber composite molecular sieve (CFCMS). The systems and methods provide advantages in that both a high energy density and a high ratio of delivered to stored gas are provided.

Limiting Factors for External Reactor Vessel Cooling

International Nuclear Information System (INIS)

Cheung, F.B.

2005-01-01

The method of external reactor vessel cooling (ERVC) that involves flooding of the reactor cavity during a severe accident has been considered a viable means for in-vessel retention (IVR). For high-power reactors, however, there are some limiting factors that might adversely affect the feasibility of using ERVC as a means for IVR. In this paper, the key limiting factors for ERVC have been identified and critically discussed. These factors include the choking limit for steam venting (CLSV) through the bottleneck of the vessel/insulation structure, the critical heat flux (CHF) for downward-facing boiling on the vessel outer surface, and the two-phase flow instabilities in the natural circulation loop within the flooded cavity. To enhance ERVC, it is necessary to eliminate or relax these limiting factors. Accordingly, methods to enhance ERVC and thus improve margins for IVR have been proposed and demonstrated, using the APR1400 as an example. The strategy is based on using two distinctly different methods to enhance ERVC. One involves the use of an enhanced vessel/insulation design to facilitate steam venting through the bottleneck of the annular channel. The other involves the use of an appropriate vessel coating to promote downward-facing boiling. It is found that the use of an enhanced vessel/insulation design with bottleneck enlargement could greatly facilitate the process of steam venting through the bottleneck region as well as streamline the resulting two-phase motions in the annular channel. By selecting a suitable enhanced vessel/insulation design, not only the CLSV but also the CHF limits could be significantly increased. In addition, the problem associated with two-phase flow instabilities and flow-induced mechanical vibration could be minimized. It is also found that the use of vessel coatings made of microporous metallic layers could greatly facilitate downward-facing boiling on the vessel outer surface. With vessel coatings, the local CHF limits at

Calculation of aerodynamics of aerosol filter designs for cleaning of heavy liquid metal cooler reactor gas loops

International Nuclear Information System (INIS)

Valery P Melnikov; Pyotr N Martynov; Albert K Papovyants; Ivan V Yagodkin

2005-01-01

Full text of publication follows: One of the basic performances of aerosol filters is the aerodynamic resistance to the flow of gaseous medium to be cleaned. Calculation of the aerodynamics of aerosol filters in reference to the gas loops of reactor installations with heavy liquid metal coolant (HLMC) allows the design of the structural components of filters to be optimized to provide minimum initial resistance values. It is established that owing to various factors aerosol particles of different concentration and disperse composition are present always in the gas spaces of heavy liquid metal cooled reactor gas loops. To prevent the negative effect of aerosols on the equipment of the gas loops, it is reasonable to use filters of multistep design with sections of preliminary and fine cleaning to catch micron and submicron particles, respectively. A computer program and technique have been developed to evaluate the aerodynamics of folded aerosol filters for different parameters of their structural components, taking account of the aerosol spectrum and concentration. The algorithm of the calculation is presented by the example of a two-step design assembled in single vessel; the filter dimensions and pattern of the air flow to be cleaned are determined under the given boundary conditions. The evaluation of the aerodynamic resistance of filters was performed with consideration for local resistances and resistances of all the structural components of the filter (sudden constriction, expansion, the flow in air channels, filtering material and so on). Correlations have been derived for the resistance of air channels, filtering materials of preliminary and fine cleaning sections as a function of such parameters as the section depth (50-500 mm), the height of separators (3,5-20 mm), the filtering surface area (1,5-30 m 2 ). Based on the calculation results, the auto-similarity domain was brought out for the minimal values of filter resistances as a function of the ratio of

Acrylic vessel cleaning tests

International Nuclear Information System (INIS)

Earle, D.; Hahn, R.L.; Boger, J.; Bonvin, E.

1997-01-01

The acrylic vessel as constructed is dirty. The dirt includes blue tape, Al tape, grease pencil, gemak, the glue or residue form these tapes, finger prints and dust of an unknown composition but probably mostly acrylic dust. This dirt has to be removed and once removed, the vessel has to be kept clean or at least to be easily cleanable at some future stage when access becomes much more difficult. The authors report on the results of a series of tests designed: (a) to prepare typical dirty samples of acrylic; (b) to remove dirt stuck to the acrylic surface; and (c) to measure the optical quality and Th concentration after cleaning. Specifications of the vessel call for very low levels of Th which could come from tape residues, the grease pencil, or other sources of dirt. This report does not address the concerns of how to keep the vessel clean after an initial cleaning and during the removal of the scaffolding. Alconox is recommended as the cleaner of choice. This acrylic vessel will be used in the Sudbury Neutrino Observatory

Radial gas turbine design

Energy Technology Data Exchange (ETDEWEB)

Krausche, S.; Ohlsson, Johan

1998-04-01

The objective of this work was to develop a program dealing with design point calculations of radial turbine machinery, including both compressor and turbine, with as few input data as possible. Some simple stress calculations and turbine metal blade temperatures were also included. This program was then implanted in a German thermodynamics program, Gasturb, a program calculating design and off-design performance of gas turbines. The calculations proceed with a lot of assumptions, necessary to finish the task, concerning pressure losses, velocity distribution, blockage, etc., and have been correlated with empirical data from VAT. Most of these values could have been input data, but to prevent the user of the program from drowning in input values, they are set as default values in the program code. The output data consist of geometry, Mach numbers, predicted component efficiency etc., and a number of graphical plots of geometry and velocity triangles. For the cases examined, the error in predicted efficiency level was within {+-} 1-2% points, and quite satisfactory errors in geometrical and thermodynamic conditions were obtained Examination paper. 18 refs, 36 figs

Power-to-Gas: storing surplus electrical energy. A design study

NARCIS (Netherlands)

Buchholz, O.S.; van der Ham, Aloysius G.J.; Veneman, Rens; Brilman, Derk Willem Frederik; Kersten, Sascha R.A.

2014-01-01

In this work a conceptual design of a Power-to-Gas (PtG) process for storing electrical energy in form of synthetic natural gas (SNG) of gas grid quality H is presented. The combination with a conventional lignite fired power plant (LPP) was investigated for possible improvement of its economic

Brazilian liquefied natural gas terminals: from the conception to the operation

Energy Technology Data Exchange (ETDEWEB)

Santos, Renata N.R. dos; Lemos, Marcelo C. de; Silva, Marcos Jose M. da; Borges, Jorge Luiz P.; Soares, Fabio L.; Grieco, Eduardo P.; Melchior, Jose Carlos [PETROBRAS Transporte S.A. (TRANSPETRO), Rio de Janeiro, RJ (Brazil)

2009-12-19

To guarantee more flexibility and safety in supplying natural gas to the Brazilian market, PETROBRAS designed and constructed two liquefied natural gas (LNG) regasification terminals. These terminals are unprecedented in the world since they are the only ones to use vessels that have been adapted to store LNG and re gasify the product onboard and to adopt the model of transferring LNG from a supply ship to a regasification vessel via cryogenic arms. Due to these different characteristics, TRANSPETRO, the PETROBRAS subsidiary in charge of operating these terminals, had to prepare itself not only to deal with a new product but also to operate terminals that have no counterparts in the world. This article gives an overview of the LNG supply chain and of the new Brazilian LNG Terminals from their conception to the preparation process to operate them. (author)

Development of advanced manufacturing technologies for low cost hydrogen storage vessels

Energy Technology Data Exchange (ETDEWEB)

Leavitt, Mark [Quantum Fuel Systems Technologies Worldwide, Inc., Irvine, CA (United States); Lam, Patrick [Boeing Research and Technology (BR& T), Seattle, WA (United States)

2014-12-29

The U.S. Department of Energy (DOE) defined a need for low-cost gaseous hydrogen storage vessels at 700 bar to support cost goals aimed at 500,000 units per year. Existing filament winding processes produce a pressure vessel that is structurally inefficient, requiring more carbon fiber for manufacturing reasons, than would otherwise be necessary. Carbon fiber is the greatest cost driver in building a hydrogen pressure vessel. The objective of this project is to develop new methods for manufacturing Type IV pressure vessels for hydrogen storage with the purpose of lowering the overall product cost through an innovative hybrid process of optimizing composite usage by combining traditional filament winding (FW) and advanced fiber placement (AFP) techniques. A numbers of vessels were manufactured in this project. The latest vessel design passed all the critical tests on the hybrid design per European Commission (EC) 79-2009 standard except the extreme temperature cycle test. The tests passed include burst test, cycle test, accelerated stress rupture test and drop test. It was discovered the location where AFP and FW overlap for load transfer could be weakened during hydraulic cycling at 85°C. To design a vessel that passed these tests, the in-house modeling software was updated to add capability to start and stop fiber layers to simulate the AFP process. The original in-house software was developed for filament winding only. Alternative fiber was also investigated in this project, but the added mass impacted the vessel cost negatively due to the lower performance from the alternative fiber. Overall the project was a success to show the hybrid design is a viable solution to reduce fiber usage, thus driving down the cost of fuel storage vessels. Based on DOE’s baseline vessel size of 147.3L and 91kg, the 129L vessel (scaled to DOE baseline) in this project shows a 32% composite savings and 20% cost savings when comparing Vessel 15 hybrid design and the Quantum

Design validation and performance of closed loop gas recirculation system

International Nuclear Information System (INIS)

Kalmani, S.D.; Majumder, G.; Mondal, N.K.; Shinde, R.R.; Joshi, A.V.

2016-01-01

A pilot experimental set up of the India Based Neutrino Observatory's ICAL detector has been operational for the last 4 years at TIFR, Mumbai. Twelve glass RPC detectors of size 2 × 2 m 2 , with a gas gap of 2 mm are under test in a closed loop gas recirculation system. These RPCs are continuously purged individually, with a gas mixture of R134a (C 2 H 2 F 4 ), isobutane (iC 4 H 10 ) and sulphur hexafluoride (SF 6 ) at a steady rate of 360 ml/h to maintain about one volume change a day. To economize gas mixture consumption and to reduce the effluents from being released into the atmosphere, a closed loop system has been designed, fabricated and installed at TIFR. The pressure and flow rate in the loop is controlled by mass flow controllers and pressure transmitters. The performance and integrity of RPCs in the pilot experimental set up is being monitored to assess the effect of periodic fluctuation and transients in atmospheric pressure and temperature, room pressure variation, flow pulsations, uniformity of gas distribution and power failures. The capability of closed loop gas recirculation system to respond to these changes is also studied. The conclusions from the above experiment are presented. The validations of the first design considerations and subsequent modifications have provided improved guidelines for the future design of the engineering module gas system.

Gas storing and processing device

International Nuclear Information System (INIS)

Kobayashi, Yoshihiro; Takano, Yosoko.

1988-01-01

Purpose: To increase the gas injection processing performance and obtain stable accumulation layers by increasing the thickness of the accumulation layers of amorphous alloy. Constitution: The gas storing processing device comprises a cylindrical vessel constituting an outer cathode for introducing gases to be processed, an inner cathode in which transition metal material and rare earth metal material as a sputtering target disposed in the vessel are combined by way of insulating material, an anode cover disposed to the upper portion of the vessel and an anode bottom disposed at the bottom thereof. It is adapted such that DC high voltage sources are connected respectively to the outer and the inner cathodes and sputtering voltage can be applied, removed and controlled independently to the transition metal and the rare earth metal of the inner cathode. This enables to control the composition ratio of the accumulation layers of amorphous alloy formed to the surface of the outer cathode, thereby enabling operation related with the gas injection ratio. (Sekiya, K.)

Off-gas recirculation system for nuclear reactors

International Nuclear Information System (INIS)

Eppler, M.; Lade, H.J.

1975-01-01

According to the invention, it is suggested to provide a buffer vessel in the ring main of the off-gas recirculation system for off-gases of a nuclear reactor to which all chambers or vessels which may contain radioactively contaminated gases are connected, within the connection line to outside air. This is to prevent the immediate release of an appreciable amount of gas to the outside air due to pressure variations conditioned by the sequence of operations - e.g. on the filling of the coolant storage. After the improvement, the released gas may be reduced to the amount of gas corresponding to the leakage gas flow entering the ring mains system. (TK) [de

Report of Task Group on Ex-Vessel Thermal-Hydraulics Corium/concrete interactions and combustible gas distribution in large dry containments

International Nuclear Information System (INIS)

1987-11-01

The Task Group on Ex-Vessel Thermal-Hydraulics was established by the PWG 2 to address the physical processes that occur in the ex-vessel phase of severe accidents, to study their impact on containment loading and failure, and to assess the available calculation methods. This effort is part of an overall CSNI effort to come to an international understanding of the issues involved. The Task Group decided to focus its initial efforts on the Large Dry Containment used extensively to contain the consequences of postulated (design basis) accidents in Light Water Reactors (LWR). Although such containments have not been designed with explicit consideration of severe accidents, recent assessments indicate a substantial inherent capability for these accidents. The Task Group has examined the loads likely to challenge the integrity of the containment, and considered the calculation of the containment's response. This report is the outcome of this effort

Flexible Composite-Material Pressure Vessel

Science.gov (United States)

Brown, Glen; Haggard, Roy; Harris, Paul A.

2003-01-01

A proposed lightweight pressure vessel would be made of a composite of high-tenacity continuous fibers and a flexible matrix material. The flexibility of this pressure vessel would render it (1) compactly stowable for transport and (2) more able to withstand impacts, relative to lightweight pressure vessels made of rigid composite materials. The vessel would be designed as a structural shell wherein the fibers would be predominantly bias-oriented, the orientations being optimized to make the fibers bear the tensile loads in the structure. Such efficient use of tension-bearing fibers would minimize or eliminate the need for stitching and fill (weft) fibers for strength. The vessel could be fabricated by techniques adapted from filament winding of prior composite-material vessels, perhaps in conjunction with the use of dry film adhesives. In addition to the high-bias main-body substructure described above, the vessel would include a low-bias end substructure to complete coverage and react peak loads. Axial elements would be overlaid to contain damage and to control fiber orientation around side openings. Fiber ring structures would be used as interfaces for connection to ancillary hardware.

Design review report for rotary mode core sample truck (RMCST) modifications for flammable gas tanks, preliminary design

International Nuclear Information System (INIS)

Corbett, J.E.

1996-02-01

This report documents the completion of a preliminary design review for the Rotary Mode Core Sample Truck (RMCST) modifications for flammable gas tanks. The RMCST modifications are intended to support core sampling operations in waste tanks requiring flammable gas controls. The objective of this review was to validate basic design assumptions and concepts to support a path forward leading to a final design. The conclusion reached by the review committee was that the design was acceptable and efforts should continue toward a final design review

Test results on direct containment heating by high-pressure melt ejection into the Surtsey vessel: The TDS test series

International Nuclear Information System (INIS)

Allen, M.D.; Blanchat, T.K.; Pilch, M.M.

1994-08-01

The Technology Development and Scoping (TDS) test series was conducted to test and develop instrumentation and procedures for performing steam-driven, high-pressure melt ejection (HPME) experiments at the Surtsey Test Facility to investigate direct containment heating (DCH). Seven experiments, designated TDS-1 through TDS-7, were performed in this test series. These experiments were conducted using similar initial conditions; the primary variable was the initial pressure in the Surtsey vessel. All experiments in this test series were performed with a steam driving gas pressure of ≅ 4 MPa, 80 kg of lumina/iron/chromium thermite melt simulant, an initial hole diameter of 4.8 cm (which ablated to a final hole diameter of ≅ 6 cm), and a 1/10th linear scale model of the Surry reactor cavity. The Surtsey vessel was purged with argon ( 2 ) to limit the recombination of hydrogen and oxygen, and gas grab samples were taken to measure the amount of hydrogen produced

Design and implementation of motion planning of inspection and maintenance robot for ITER-like vessel

Energy Technology Data Exchange (ETDEWEB)

Wang, Hesheng; Lai, Yinping [Department of Automation, Shanghai Jiao Tong University, Shanghai 200240 (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China (China); Chen, Weidong, E-mail: wdchen@sjtu.edu.cn [Department of Automation, Shanghai Jiao Tong University, Shanghai 200240 (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China (China); Cao, Qixin [Institute of Robotics, Shanghai Jiao Tong University, Shanghai 200240 (China)

2015-12-15

Robot motion planning is a fundamental problem to ensure the robot executing the task without clashes, fast and accurately in a special environment. In this paper, a motion planning of a 12 DOFs remote handling robot used for inspecting the working state of the ITER-like vessel and maintaining key device components is proposed and implemented. Firstly, the forward and inverse kinematics are given by analytic method. The work space and posture space of this manipulator are both considered. Then the motion planning is divided into three stages: coming out of the cassette mover, moving along the in-vessel center line, and inspecting the D-shape section. Lastly, the result of experiments verified the performance of the motion design method. In addition, the task of unscrewing/screwing the screw demonstrated the feasibility of system in function.

UK regulatory aspects of prestressed concrete pressure vessels for gas-cooled reactor nuclear power stations

International Nuclear Information System (INIS)

Watson, P.S.

1990-01-01

Safety assessment principles for nuclear power plants and for nuclear chemical plants demand application of best proven techniques, recognised standards, adequacy margins, inspection and maintenance of all the components including prestressed concrete pressure vessels. In service inspection of prestressed concrete pressure vessels includes: concrete surface examination; anchorage inspection; tendon load check; tendon material examination; foundation settlement and tilt; log-term deformation; vessel temperature excursions; coolant loss; top cap deflection. Hartlepool and Heysham 1 power plants prestress shortfall problem is discussed. Main recommendations can be summarised as follows: at all pressure vessel stations prestress systems should be calibrated in a manner which results in all load bearing components being loaded in a representative manner; at all pressure vessel stations load measurements during calibration should be verified by a redundant and diverse system

Summary of design of nuclear vessels and piping to ASME III (NB, NC, ND) and vessels to BS 5500

International Nuclear Information System (INIS)

Harrop, L.P.

1992-01-01

There is a hierarchy of design code requirements for pressurised components, starting with non-nuclear codes as the minimum and progressing through the ASME III nuclear Classes 3, 2, 1. In establishing and assessing the safety justifications of nuclear plants it is important to have an appreciation of the gradation of requirements in the ASME III design rules and how these go beyond non-nuclear component design rules. There are two broad aspects to the structural integrity of pressurised components, namely the achievement of integrity and the demonstration of integrity. The technical requirements of design codes are associated with achieving integrity while the documentary aspects are usually associated with demonstrating integrity. In practice documents also have a part in achieving integrity in the communication of information between different organisations and personnel involved in the design process. It is not possible to assign simple numerical measures to the relative integrity afforded by non-nuclear codes and the three Classes of ASME III. Instead it is necessary to compare the different requirements of the rules for the various technical and documentary aspects. This paper summarises the most important technical and documentary aspects of the three Classes of the ASME III Code for vessels and the non-nuclear code BS 5500. A similar summary is also provided for the three Classes of ASME III rules for piping. The intention is that the paper provides a basis for appreciating the relative integrity afforded by these various rules. (author)

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.

Design and Structural Analysis for the Vacuum Vessel of Superconducting Tokamak JT-60SC

International Nuclear Information System (INIS)

Kudo, Y.; Sakurai, S.; Masaki, K.; Urata, K.; Sasajima, T.; Matsukawa, M.; Sakasai, A.; Ishida, S.

2003-01-01

A modification of the JT-60 is planned to be a superconducting tokamak (JT-60SC) in order to establish steady-state operation of high beta plasma for 100 s, and to ensure the applicability of ferritic steel as a reduced activation material for reactor relevant break-even class plasmas. This paper describes the detailed design of the vacuum vessel, which has a unique structure for cost effective manufacturing, as well as structural analysis results for a feasibility study

Presentation summary: Gas Turbine - Modular Helium Reactor (GT-MHR)

International Nuclear Information System (INIS)

2001-01-01

Numerous prototypes and demonstration plants have been constructed and operated beginning with the Dragon plant in the early 1960s. The MHTGR was the U.S. developed modular plant and underwent pre application review by NRC. The GT-MHR represents a further refinement on this concept with the steam cycle being replaced by a closed loop gas turbine (Brayton) cycle. Modular gas reactors and the GT-MHR represent a fundamental shift in reactor design and safety philosophy. The reactor system is contained in a 3 vessel, side-by-side arrangement. The reactor and a shutdown cooling system are in one vessel, and the gas turbine based power conversion system, including the generator, in a second parallel vessel. A more detailed look at the system shows the compact arrangement of gas turbine, compressors, recuperator, heat exchanges, and generator. Fueled blocks are stacked in three concentric rings with inert graphite blocks making up the inner and outer reflectors. Operating control rods are located outside the active core while startup control rods and channels for reserve shutdown pellets are located near the core center. Ceramic coated fuel is the key to the GT-MHR's safety and economics. A kernel of Uranium oxycarbide (or UO 2 ) is placed in a porous carbon buffer and then encapsulated in multiple layers of pyrolytic carbon and silicon carbide. These micro pressure vessels withstand internal pressures of up to 2,000 psi and temperatures of nearly 2,000 C providing extremely resilient containment of fission products under both normal operating and accident conditions. The fuel particles are blended in carbon pitch, forming fuel rods, and then loaded into holes within large graphite fuel elements. Fuel elements are stacked to form the core. Fuel particle testing in has repeatedly demonstrated the high temperature resilience of coated particle fuel to temperature approaching 2,000 C. As an conservative design goal, GT-MHR has been sized to keep maximum fuel temperatures

Design considerations and data for gas-insulated high voltage structures

International Nuclear Information System (INIS)

Hopkins, D.B.

1975-11-01

This paper is intended to benefit the person faced with the occasional task of designing gas insulated high-voltage structures or spark gaps and who must decide upon the proper geometry, spacings, gas type, and pressure for reliable voltage-holding. An approach is presented along with a summary of how various factors affect voltage breakdown. The design procedures described apply to situations where the influence of nearby insulators is negligible. The accuracy of the data is estimated to be within 10 to 15 percent, a value usually attained in practice only when one follows the cautionary advice discussed in the paragraphs on materials preparation, gas properties, and conditioning

Depressurization as a means of leak checking large vacuum vessels

International Nuclear Information System (INIS)

Callis, R.W.; Langhorn, A.; Petersen, P.I.; Ward, C.; Wesley, J.

1985-01-01

A common problem associated with large vacuum vessels used in magnetic confinement fusion experiments is that leak checking is hampered by the inaccessibility to most of the vacuum vessel surface. This inaccessibility is caused by the close proximity of magnetic coils, diagnostics and, for those vessels that are baked, the need to completely surround the vessel with a thermal insulation blanket. These obstructions reduce the effectiveness of the standard leak checking method of using a mass spectrometer and spraying a search gas such as helium on the vessel exterior. Even when the presence of helium is detected, its entry point into the vessel cannot always be pinpointed. This paper will describe a method of overcoming this problem. By slightly depressurizing the vessel, an influx of helium through the leak is created. The leak site can then be identified by personnel within the vessel using standard sniffing procedures. There are two conditions which make this method of leak checking practical. First, the vessel need only be depressurized 2 psi, thus allowing personnel inside to perform the sniffing operation. Second, the sniffing probe used (Leybold--Heraus ''Quick Test'') could detect a change in helium concentration as small as 100 ppb, which allows for faster scanning of the vessel inferior. Use of this technique to find an elusive 10 -3 Torrxl/s leak in the Doublet III tokamak vacuum vessel will be presented

The impact of vessel speed reduction on port accidents.

Science.gov (United States)

Chang, Young-Tae; Park, Hyosoo

2016-03-19

Reduced-speed zones (RSZs) have been designated across the world to control emissions from ships and prevent mammal strikes. While some studies have examined the effectiveness of speed reduction on emissions and mammal preservation, few have analyzed the effects of reduced ship speed on vessel safety. Those few studies have not yet measured the relationship between vessel speed and accidents by using real accident data. To fill this gap in the literature, this study estimates the impact of vessel speed reduction on vessel damages, casualties and frequency of vessel accidents. Accidents in RSZ ports were compared to non-RSZ ports by using U.S. Coast Guard data to capture the speed reduction effects. The results show that speed reduction influenced accident frequency as a result of two factors, the fuel price and the RSZ designation. Every $10 increase in the fuel price led to a 10.3% decrease in the number of accidents, and the RSZ designation reduced vessel accidents by 47.9%. However, the results do not clarify the exact impact of speed reduction on accident casualty. Copyright © 2016 Elsevier Ltd. All rights reserved.

In-vessel maintenance remote manipulator system

International Nuclear Information System (INIS)

Jimenez, E.

1978-01-01

The radiation environment within the Tokamak Fusion Test Reactor (TFTR) vacuum vessel necessitates the development of a Remote Manipulator System (RMS) to perform required periodic inspection and maintenance tasks. The RMS must be able to perform dexterous operations and handle loads that exceed human capabilities. The limited size of the access ports on the TFTR vacuum vessel and the performance profile, defined by the various handling requirements, present unique design constraints. The design approach and formulation of a RMS configuration which satisfies TFTR requirements is presented herein

Cold source vessel development for the advanced neutron source

Energy Technology Data Exchange (ETDEWEB)

Williams, P.T.; Lucas, A.T. [Oak Ridge National Lab., TN (United States)

1995-09-01

The Advanced Neutron Source (ANS), in its conceptual design phase at Oak Ridge National Laboratory (ORNL), will be a user-oriented neutron research facility that will produce the most intense flux of neutrons in the world. Among its many scientific applications, the productions of cold neutrons is a significant research mission for the ANS. The cold neutrons come from two independent cold sources positioned near the reactor core. Contained by an aluminum alloy vessel, each cold source is a 410 mm diameter sphere of liquid deuterium that functions both as a neutron moderator and a cryogenic coolant. With nuclear heating of the containment vessel and internal baffling, steady-state operation requires close control of the liquid deuterium flow near the vessel`s inner surface. Preliminary thermal-hydraulic analyses supporting the cold source design are being performed with multi-dimensional computational fluid dynamics simulations of the liquid deuterium flow and heat transfer. This paper presents the starting phase of a challenging program and describes the cold source conceptual design, the thermal-hydraulic feasibility studies of the containment vessel, and the future computational and experimental studies that will be used to verify the final design.

Heat exchanger for transfering heat produced in a high temperature reactor to an intermediate circuit gas

International Nuclear Information System (INIS)

Barchewitz, E.; Baumgaertner, H.

1985-01-01

The invention is concerned with improving the arrangement of a heat exchanger designed to transfer heat from the coolant gas circuit of a high temperature reactor to a gas which is to be used for a process heat plant. In the plant the material stresses are to be kept low at high differential pressures and temperatures. According to the invention the tube bundles designed as boxes are fixed within the heat exchanger closure by means of supply pipes having got loops. For conducting the hot gas the heat exchanger has got a central pipe leading out of the reactor vessel through the pod closure and having got only one point of fixation, lying in this closure. Additional advantageous designs are mentioned. (orig./PW)

The chemistry of tributyl phosphate at elevated temperatures in the Plutonium Finishing Plant Process Vessels

International Nuclear Information System (INIS)

Barney, G.S.; Cooper, T.D.

1994-01-01

Potentially violent chemical reactions of the tributyl phosphate solvent used by the Plutonium Finishing Plant at the Hanford Site were investigated. There is a small probability that a significant quantity of this solvent could be accidental transferred to heated process vessels and react there with nitric acid or plutonium nitrate also present in the solvent extraction process. The results of laboratory studies of the reactions show that exothermic oxidation of tributyl phosphate by either nitric acid or actinide nitrates is slow at temperatures expected in the heated vessels. Less than four percent of the tributyl phosphate will be oxidized in these vented vessels at temperatures between 125 degrees C and 250 degrees C because the oxidant will be lost from the vessels by vaporization or decomposition before the tributyl phosphate can be extensively oxidized. The net amounts of heat generated by oxidation with concentrated nitric acid and with thorium nitrate (a stand-in for plutonium nitrate) were determined to be about -150 and -220 joules per gram of tributyl phosphate initially present, respectively. This is not enough heat to cause violent reactions in the vessels. Pyrolysis of the tributyl phosphate occurred in these mixtures at temperatures of 110 degrees C to 270 degrees C and produced mainly 1-butene gas, water, and pyrophosphoric acid. Butene gas generation is slow at expected process vessel temperatures, but the rate is faster at higher temperatures. At 252 degrees C the rate of butene gas generated was 0.33 g butene/min/g of tributyl phosphate present. The measured heat absorbed by the pyrolysis reaction was 228 J/g of tributyl phosphate initially present (or 14.5 kcal/mole of tributyl phosphate). Release of flammable butene gas into process areas where it could ignite appears to be the most serious safety consideration for the Plutonium Finishing Plant

The chemistry of tributyl phosphate at elevated temperatures in the Plutonium Finishing Plant Process Vessels

Energy Technology Data Exchange (ETDEWEB)

Barney, G.S.; Cooper, T.D.

1994-06-01

Potentially violent chemical reactions of the tributyl phosphate solvent used by the Plutonium Finishing Plant at the Hanford Site were investigated. There is a small probability that a significant quantity of this solvent could be accidental transferred to heated process vessels and react there with nitric acid or plutonium nitrate also present in the solvent extraction process. The results of laboratory studies of the reactions show that exothermic oxidation of tributyl phosphate by either nitric acid or actinide nitrates is slow at temperatures expected in the heated vessels. Less than four percent of the tributyl phosphate will be oxidized in these vented vessels at temperatures between 125{degrees}C and 250{degrees}C because the oxidant will be lost from the vessels by vaporization or decomposition before the tributyl phosphate can be extensively oxidized. The net amounts of heat generated by oxidation with concentrated nitric acid and with thorium nitrate (a stand-in for plutonium nitrate) were determined to be about -150 and -220 joules per gram of tributyl phosphate initially present, respectively. This is not enough heat to cause violent reactions in the vessels. Pyrolysis of the tributyl phosphate occurred in these mixtures at temperatures of 110{degrees}C to 270{degrees}C and produced mainly 1-butene gas, water, and pyrophosphoric acid. Butene gas generation is slow at expected process vessel temperatures, but the rate is faster at higher temperatures. At 252{degrees}C the rate of butene gas generated was 0.33 g butene/min/g of tributyl phosphate present. The measured heat absorbed by the pyrolysis reaction was 228 J/g of tributyl phosphate initially present (or 14.5 kcal/mole of tributyl phosphate). Release of flammable butene gas into process areas where it could ignite appears to be the most serious safety consideration for the Plutonium Finishing Plant.

Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230