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Sample records for reactor building structure

  1. Vibration-damping structure for reactor building

    International Nuclear Information System (INIS)

    Kuno, Toshio; Iba, Chikara; Tanaka, Hideki; Kageyama, Mitsuru

    1998-01-01

    In a damping structure of a reactor building, an inner concrete body and a reactor container are connected by way of a vibration absorbing member. As the vibration absorbing member, springs or dampers are used. The inner concrete body and the reactor container each having weight and inherent frequency different from each other are opposed displaceably by way of the vibration absorbing member thereby enabling to reduce seismic input and reduce shearing force at least at leg portions. Accordingly, seismic loads are reduced to increase the grounding rate of the base thereby enabling to satisfy an allowable value. Therefore, it is not necessary to strengthen the inner concrete body and the reactor container excessively, the amount of reinforcing rods can be reduced, and the amount of a portion of the base buried to the ground can be reduced thereby enabling to constitute the reactor building easily. (N.H.)

  2. Nuclear reactor buildings

    International Nuclear Information System (INIS)

    Nagashima, Shoji; Kato, Ryoichi.

    1985-01-01

    Purpose: To reduce the cost of reactor buildings and satisfy the severe seismic demands in tank type FBR type reactors. Constitution: In usual nuclear reactor buildings of a flat bottom embedding structure, the flat bottom is entirely embedded into the rock below the soils down to the deck level of the nuclear reactor. As a result, although the weight of the seismic structure can be decreased, the amount of excavating the cavity is significantly increased to inevitably increase the plant construction cost. Cross-like intersecting foundation mats are embedded to the building rock into a thickness capable withstanding to earthquakes while maintaining the arrangement of equipments around the reactor core in the nuclear buildings required by the system design, such as vertical relationship between the equipments, fuel exchange systems and sponteneous drainings. Since the rock is hard and less deformable, the rigidity of the walls and the support structures of the reactor buildings can be increased by the embedding into the rock substrate and floor responsivity can be reduced. This enables to reduce the cost and increasing the seismic proofness. (Kamimura, M.)

  3. Life management for a non replaceable structure: the reactor building

    International Nuclear Information System (INIS)

    Torres, V.; Francia, L.

    1998-01-01

    Phase 1 of UNESA N.P.P. Lifetime Management Project identified and ranked important components, relative to plant life management. The list showed the Reactor Containment Structure in the third position, and thirteen concrete structures were among the list top twenty. Since the Reactor Containment Building, together with the Reactor Vessel, is the only non-replaceable plant component, and has a big impact on the plant technical life, there is an increasing interest on understanding its behavior to maintain structural integrity. This paper presents: a) IAEA (International Atomic Energy Agency) Coordinated Research Program experiences and studies. Under this Program, international experts address the most frequent degradation mechanisms affecting the containment building. b) IAEA Aging Management Program adapted to our plants. The paper addresses the aging mechanisms affecting the concrete structures, reinforcing steel and prestress systems as well as the aging management programs and the mitigation and control methods. Finally, this paper presents a new module called STRUCTURES, included in phase 2 of the above mentioned project, which will monitor and document the different aging mechanisms and management programs described in item b) regarding the Reactor Containment Building (concrete liner, post stressing system, anchor elements). This module will also support the Maintenance Rule related practices. (Author)

  4. Reactor building

    International Nuclear Information System (INIS)

    Ebata, Sakae.

    1990-01-01

    At least one valve rack is disposed in a reactor building, on which pipeways to a main closure valve, valves and bypasses of turbines are placed and contained. The valve rack is fixed to the main body of the building or to a base mat. Since the reactor building is designed as class A earthquake-proofness and for maintaining the S 1 function, the valve rack can be fixed to the building main body or to the base mat. With such a constitution, the portions for maintaining the S 1 function are concentrated to the reactor building. As a result, the dispersion of structures of earthquake-proof portion corresponding to the reference earthquake vibration S 1 can be prevented. Accordingly, the conditions for the earthquake-proof design of the turbine building and the turbine/electric generator supporting rack are defined as only the class B earthquake-proof design conditions. In view of the above, the amount of building materials can be saved and the time for construction can be shortened. (I.S.)

  5. Analysis of soil-structure interaction and floor response spectrum of reactor building for China advanced research reactor

    International Nuclear Information System (INIS)

    Rong Feng; Wang Jiachun; He Shuyan

    2006-01-01

    Analysis of Soil-Structure Interaction (SSI) and calculation of Floor Response Spectrum (FRS) is substantial for anti-seismic design for China Advanced Research Reactor (CARR) project. The article uses direct method to analyze the seismic reaction of the reactor building in considering soil-structure interaction by establishing two-dimensional soil-structure co-acting model for analyzing and inputting of seismic waves from three directions respectively. The seismic response and floor response spectrum of foundation and floors of the building under different cases have been calculated. (authors)

  6. Nuclear reactor building

    International Nuclear Information System (INIS)

    Oshima, Nobuaki.

    1991-01-01

    The secondary container in a nuclear reactor building is made of a transparent structure having a shielding performance such as lead glass, by which the inside of the secondary container can be seen without undergoing radiation exposure. In addition, an operator transportation facility capable of carrying about 5 to 10 operators at one time is disposed, and the side of the facility on the secondary container is constituted with a transparent material such as glass, to provide a structure capable of observing the inside of the secondary container. The ventilation and air conditioning in the operator's transportation facility is in communication with the atmosphere of a not-controlled area. Accordingly, operators at the outside of the reactor building can reach the operator's transportation facility without taking and procedures for entering the controlled area and without undergoing radiation exposure. The inside of the secondary container in the reactor building can be seen from various directions through the transparent structure having the shielding performance. (N.H.)

  7. Experimental and analytical studies on soil-structure interaction behavior of nuclear reactor building

    International Nuclear Information System (INIS)

    Tsushima, Y.

    1978-01-01

    The purpose of this study is to estimate damping effects due to soil-structure interaction by the dissipation of vibrational energy to the ground through the foundation in a building with a short fundamental period such as a nuclear reactor building. The author performed experimental and analytical studies on the vibrational characteristics of model steel structures ranging from one to four stories high erected on the rigid base and located on soil, which are simulated from the vibrational characteristics of a prototype reactor building: the former study is to obtain damping effects due to inner friction of steel frames and the latter to obtain radiation damping effects due to soil-structure interaction. The author also touches upon the results of experiments performed on a BWR-type reactor building in 1974, which showed damping ratios higher than 20% of those in fundamental modes. Then the author attempts to estimate the damping effects of the reactor building by his own method proposed in the report. Through these studies the author finally concludes that the experimental damping effects are remarkable in the lower modes by the energy dissipation and the analytical results show a fairly good fit to the experimental ones

  8. Structural analysis of reactor buildings with help of complete FE models

    International Nuclear Information System (INIS)

    Diaz, B.E.; Vaz, L.E.; Martha, L.F.R.; Costa, E.

    1984-01-01

    The reinforced concrete structures located within the steel containment shell of a Reactor Building are formed by highly complex structures subjected to a large amount of actions due to different causes. The analysis of this complex structure can be performed with help of small models, each one representing a part of the global structure. The interaction effects among the partial models are accounted for in approximate way. This approach has been used previously with entire success in the design of 1300 MW PWR nuclear power plants. However a new and entire different approach can be used in the design of these structures. The entire assembly of structural elements of the building is represented and analyzed with help of a single and very large FE model. This paper will present the main characteristics of this type of analysis as well as all the necessary procedures, which must be implemented for the proper data processing of the forces and the automatic reinforced concrete design of the structural elements of the Reactor Building. (Author) [pt

  9. Internal structure of reactor building for Madras Atomic Power Project

    International Nuclear Information System (INIS)

    Pandit, D.P.

    1975-01-01

    The structural configuration and analysis of structural elements of the internal structure of reactor building for the Madras Atomic Power Project has been presented. Two methods of analysis of the internal structure, viz. Equivalent Plane Frame and Finite Element Method, are explained and compared with the use of bending moments obtained. (author)

  10. Structural design of SBWR reactor building complex using microcomputers

    International Nuclear Information System (INIS)

    Mandagi, K.; Rajagopal, R.S.; Sawhney, P.S.; Gou, P.F.

    1993-01-01

    The design concept of Simplified Boiling Water Reactor (SBWR) plant is based on simplicity and passive features to enhance safety and reliability, improve performance, and increase economic viability. The SBWR utilizes passive systems such as Gravity Driven Core-Cooling System (GDCS) and Passive Containment Cooling System (PCCS). To suit these design features the Reactor Building (RB) complex of the SBWR is configured as an integrated structure consisting of a cylindrical Reinforced Concrete Containment Vessel (RCCV) surrounded by square reinforced concrete safety envelope and outer box structures, all sharing a common reinforced concrete basemat. This paper describes the structural analysis and design aspects of the RB complex. A 3D STARDYNE finite element model has been developed for the structural analysis of the complex using a PC Compaq 486/33L microcomputer. The structural analysis is performed for service and factored load conditions for the applicable loading combinations. The dynamic responses of containment structures due to pool hydrodynamic loads have been calculated by an axisymmetric shell model using COSMOS/M program. The RCCV is designed in accordance with ASME Section 3, Division 2 Code. The rest of the RB which is classified as Seismic Category 1 structure is designed in accordance with the ACI 349 Code. This paper shows that microcomputers can be efficiently used for the analysis and design of large and complex structures such as RCCV and Reactor Building complex. The use of microcomputers can result in significant savings in the computational cost compared with that of mainframe computers

  11. Seismic retrofitting of Apsara reactor building

    International Nuclear Information System (INIS)

    Reddy, G.R.; Parulekar, Y.M.; Sharma, A.; Rao, K.N.; Narasimhan, Rajiv; Srinivas, K.; Basha, S.M.; Thomas, V.S.; Soma Kumar, K.

    2006-01-01

    Seismic analysis of Apsara Reactor building was carried out and was found not meeting the current seismic requirements. Due to the building not qualifying for seismic loads, a retrofit scheme using elasto-plastic dampers is proposed. Following activities have been performed in this direction: Carried out detailed seismic analysis of Apsara reactor building structure incorporating proposed seismic retrofit. Demonstrating the capability of the retrofitted structure to with stand the earth quake level for Trombay site as per the current standards by analysis and by model studies. Implementation of seismic retrofit program. This paper presents the details of above aspects related to Seismic analysis and retrofitting of Apsara reactor building. (author)

  12. Seismic analysis of a PWR 900 reactor: study of reactor building with soil-structure interaction and evaluation of floor spectra

    International Nuclear Information System (INIS)

    Gantenbein, F.; Aguilar, J.

    1983-08-01

    The purpose of this paper is the evaluation of seismic response and floor spectra for a typical PWR 900 reactor building with respect to soil-structure interaction for soil stiffness). The typical PWR 900 reactor building consists of a concrete cylindrical external building and roof dome, a concrete internal structure (internals) on a common foundation mat as illustrated. The seismic response is obtained by SRSS method and floor spectra directly from ground spectrum and modal properties of the structure. Seismic responses and floor spectra computation is performed in the case of two different ground spectra: EDF spectrum (mean of oscillator spectra obtained from 8 californian records) normalized to 0.2 g, and DSN spectrum (typical of shallow seism) normalized to 0.3 g. The first section is devoted to internals' modelisation, the second one to the axisymmetric model of the reactor, the third one to the seismic response, the fourth one to floor spectra

  13. ITER [International Thermonuclear Experimental Reactor] reactor building design study

    International Nuclear Information System (INIS)

    Thomson, S.L.; Blevins, J.D.; Delisle, M.W.

    1989-01-01

    The International Thermonuclear Experimental Reactor (ITER) is at the midpoint of a two-year conceptual design. The ITER reactor building is a reinforced concrete structure that houses the tokamak and associated equipment and systems and forms a barrier between the tokamak and the external environment. It provides radiation shielding and controls the release of radioactive materials to the environment during both routine operations and accidents. The building protects the tokamak from external events, such as earthquakes or aircraft strikes. The reactor building requirements have been developed from the component designs and the preliminary safety analysis. The equipment requirements, tritium confinement, and biological shielding have been studied. The building design in progress requires continuous iteraction with the component and system designs and with the safety analysis. 8 figs

  14. Japanese contributions to containment structure, assembly and maintenance and reactor building for ITER

    International Nuclear Information System (INIS)

    Shibanuma, Kiyoshi; Honda, Tsutomu; Kanamori, Naokazu

    1991-06-01

    Joint design work on Conceptual Design Activity of International Thermonuclear Experimental Reactor (ITER) with four parties, Japan, the United States, the Soviet Union and the European Community began in April 1988 and was successfully completed in December 1990. In Japan, the home team was established in wide range of collaboration between JAERI and national institute, universities and heavy industries. The Fusion Experimental Reactor (FER) Team at JAERI is assigned as a core of the Japanese home team to support the joint Team activity and mainly conducted the design and R and D in the area of containment structure, remote handling and plant system. This report mainly describes the Japanese contribution on the ITER containment structure, remote handling and reactor building design. Main areas of contributions are vacuum vessel, attaching locks, electromagnetic analysis, cryostat, port and service line layout for containment structure, in-vessel handling equipment design and analysis, blanket handling equipment design and related short term R and D for assembly and maintenance, and finally reactor building design and analysis based on the equipment and service line layout and components flow during assembly and maintenance. (author)

  15. Structural safety of HDR reactor building during large scale vibration tests

    International Nuclear Information System (INIS)

    Stangenberg, F.; Zinn, R.

    1985-01-01

    In the second phase of the HDR investigations, a high shaker excitation of the building is planned using a large shaker which will be located on the operating floor and will be brought up to speed in a balanced condition and then unbalanced and decoupled from the drive system. With decreasing speed the shaker comes in resonance with the building frequencies and its energy is transferred to the building. In this paper the structural safety of the reactor building during the projected shaker tests is analysed. Dynamic response calculations with coupling between building and shaker by simultaneously integrating the equilibrium equations of both building and shaker are presented. The resulting building stresses, soil pressures etc. are compared with allowable values. (orig.)

  16. Reactor building for a nuclear reactor

    International Nuclear Information System (INIS)

    Haidlen, F.

    1976-01-01

    The invention concerns the improvement of the design of a liner, supported by a latticed steel girder structure and destined for guaranteeing a gastight closure for the plant compartments in the reactor building of a pressurized water reactor. It is intended to provide the steel girder structure on their top side with grates, being suited for walking upon, and to hang on their lower side diaphragms in modular construction as a liner. At the edges they may be sealed with bellows in order to avoid thermal stresses. The steel girder structure may at the same time serve as supports for parts of the steam pipe. (RW) [de

  17. Aircraft-crash-protected steel reactor building roof structure for the European market

    International Nuclear Information System (INIS)

    Posta, B.A.; Kadar, I.; Rao, A.S.

    1996-01-01

    This paper recommends the use of all steel roof structures for the reactor building of European Boiling Water Reactor (BWR) plants. This change would make the advanced US BWR designs more compatible with European requirements. Replacement of the existing concrete roof slab with a sufficiently thick steel plate would eliminate the concrete spelling resulting from a postulated aircraft crash, potentially damaging the drywell head or the spent fuel pool

  18. Structure of steel reactor building and construction method therefor

    International Nuclear Information System (INIS)

    Yamakawa, Toshikimi.

    1997-01-01

    The building of the present invention contains a reactor pressure vessel, and has double steel plate walls endurable to elevation of inner pressure and keeping airtightness, and shielding concretes are filled between the double steel plate walls. It also has empty double steel plate walls not filled with concretes and has pipelines, vent ducts, wirings and a support structures for attaching them between the double steel plate walls. It is endurable to a great inner pressure satisfactory and keeps airtightness by the two spaced steel plates. It can be greatly reduced in the weight, and can be manufactured efficiently with high quality in a plant by so called module construction, and the dimension of the entire of the reactor building can be reduced. It is constructed in a dock, transported on the sea while having the space between the two steel plate walls as a ballast tanks, placed in the site, and shielding concretes are filled between the double steel plate walls. The term for the construction can be reduced, and the cost for the construction can be saved. (N.H.)

  19. Dynamic analysis of reactor containment building using axisymmetric finite element model

    International Nuclear Information System (INIS)

    Thakkar, S.K.; Dubey, R.N.

    1989-01-01

    The structural safety of nuclear reactor building during earthquake is of great importance in view of possibility of radiation hazards. The rational evaluation of forces and displacements in various portions of structure and foundation during strong ground motion is most important for safe performance and economic design of the reactor building. The accuracy of results of dynamic analysis is naturally dependent on the type of mathematical model employed. Three types of mathematical models are employed for dynamic analysis of reactor building beam model axisymmetric finite element model and three dimensional model. In this paper emphasis is laid on axisymmetric model. This model of containment building is considered a reinfinement over conventional beam model of the structure. The nuclear reactor building on a rocky foundation is considered herein. The foundation-structure interaction is relatively less in this condition. The objective of the paper is to highlight the significance of modelling of non-axisymmetric portion of building, such as reactor internals by equivalent axisymmetric body, on the structural response of the building

  20. Study on reactor building structure using ultrahigh strength materials, 1

    International Nuclear Information System (INIS)

    Ishimura, Kikuo; Odajima, Masahiro; Irino, Kazuo; Hashiba, Toshio.

    1991-01-01

    This study was promoted to be aimed at realization of the optimal nuclear reactor building structure of the future. As the first step, the study regarding ultrahigh strength reinforced concrete (abbr. RC) shear wall was selected. As the result of various tests, the application of ultrahigh strength RC shear walls was verified. The tests conducted were relevant to; ultrahigh strength concrete material tests; pure shear tests of RC flat panels; and bending shear tests and its simulation analysis of RC shear walls. (author)

  1. Aging management program of the reactor building concrete at Point Lepreau Generating Station

    Science.gov (United States)

    Aldea, C.-M.; Shenton, B.; Demerchant, M. M.; Gendron, T.

    2011-04-01

    In order for New Brunswick Power Nuclear (NBPN) to control the risks of degradation of the concrete reactor building at the Point Lepreau Generating Station (PLGS) the development of an aging management plan (AMP) was initiated. The intention of this plan was to determine the requirements for specific structural components of concrete of the reactor building that require regular inspection and maintenance to ensure the safe and reliable operation of the plant. The document is currently in draft form and presents an integrated methodology for the application of an AMP for the concrete of the reactor building. The current AMP addresses the reactor building structure and various components, such as joint sealant and liners that are integral to the structure. It does not include internal components housed within the structure. This paper provides background information regarding the document developed and the strategy developed to manage potential degradation of the concrete of the reactor building, as well as specific programs and preventive and corrective maintenance activities initiated.

  2. Reactor building with internal structure of which the movements are independent of those of the general raft and process for building these internal structures

    International Nuclear Information System (INIS)

    Hista, J.C.

    1982-01-01

    This reactor building includes a containment enclosure for the internal structures composed of a slab wedged on its periphery against the containment enclosure gusset and resting on the general raft by means of a peripheral bearing ring, a compressible layer being provided between the general raft and the slab [fr

  3. Study on reactor building structure using ultrahigh strength materials - Part 9: Summary of the study

    International Nuclear Information System (INIS)

    Tanaka, H.; Odajima, M.; Irino, K.; Hashiba, T.

    1993-01-01

    Considerations for longevity of nuclear facilities and ease of decommissioning are of great importance for future nuclear power plants. To this end, a concept of an optimal structural concept for nuclear reactor buildings has been studied: the main feature of this concept is to utilize large-sized, light weight prefabricated members with ultrahigh strength materials. The following two items have been selected to study the prospective structure: (1) Applicability of ultrahigh strength materials for reinforced concrete shear walls (2) Construction using large sized prefabricated members As the first step (1), material and structural tests using ultrahigh strength materials, and the subsequent analysis of those tests for reinforced concrete shear walls, has been conducted. The positive results of this study show a bright future for the use of ultrahigh strength materials for the reinforced concrete shear walls of nuclear reactor buildings. As the second step (2), tests on a mixed structure with precasted members have been conducted. Our results positively suggest the use of these materials and methods to improve prospective nuclear power plants. (author)

  4. Aging management program of the reactor building concrete at Point Lepreau Generating Station

    Directory of Open Access Journals (Sweden)

    Gendron T.

    2011-04-01

    Full Text Available In order for New Brunswick Power Nuclear (NBPN to control the risks of degradation of the concrete reactor building at the Point Lepreau Generating Station (PLGS the development of an aging management plan (AMP was initiated. The intention of this plan was to determine the requirements for specific structural components of concrete of the reactor building that require regular inspection and maintenance to ensure the safe and reliable operation of the plant. The document is currently in draft form and presents an integrated methodology for the application of an AMP for the concrete of the reactor building. The current AMP addresses the reactor building structure and various components, such as joint sealant and liners that are integral to the structure. It does not include internal components housed within the structure. This paper provides background information regarding the document developed and the strategy developed to manage potential degradation of the concrete of the reactor building, as well as specific programs and preventive and corrective maintenance activities initiated.

  5. Air conditioning device for reactor buildings

    International Nuclear Information System (INIS)

    Kikuchi, Shiro.

    1982-01-01

    Purpose: To decrease the opening areas of pipe lines for an air conditioning device at the portions passing through the shielding walls of a reactor building for a FBR type reactor, as well as reduce the size of the building. Constitution: Airs in the building for containing reactor are liquefied in an air liquefying mechanism. The liquefied airs are sent by way of pipe lines to each of evaporators, wherein each of the chambers are cooled because of latent heat of evaporation and evaporated airs are released to each of the chambers. The airs released to each of the chambers are collected into an exhaust chamber and sent by way of a duct to the air liquefying mechanism and liquefied again. Since the volume of the liquefied airs may be smaller than the amount conventionally required for usual cooled airs, the pipe lines passing through the shielding walls of the building can be of smaller diameter. This can decrease the opening areas of the pipe lines at the portions passing through the walls of the shieldings and, since the opening areas are smaller, the structure of the radiation shieldings can be simplified in these portions. Further, since the space of the pipe lines in the building is reduced extremely, the size of the building can be reduced. (Moriyama, K.)

  6. Nuclear reactor melt-retention structure to mitigate direct containment heating

    International Nuclear Information System (INIS)

    Tutu, N.K.; Ginsberg, T.; Klages, J.R.

    1991-01-01

    This patent describes a nuclear reactor melt-retention structure that functions to retain molten core material within a melt retention chamber to mitigate the extent of direct containment heating. The structure being adapted to be positioned within or adjacent to a pressurized or boiling water nuclear reactor containment building at a location such that at least a portion of the melt retention structure is lower than and to one side of the nuclear reactor pressure vessel, and such that the structure is adjacent to a gas escape channel means that communicates between the reactor cavity and the containment building of the reactor. It comprises a melt-retention chamber, wall means defining a passageway extending between the reactor cavity underneath the reactor pressure vessel and one side of the chamber, the passageway including vent means extending through an upper wall portion thereof. The vent means being in communication with the upper region of the reactor containment building, whereby gas and steam discharged from the reactor pressure vessel are vented through the passageway and vent means into the gas-escape channel means and the reactor containment building

  7. Method of decommissioning nuclear reactor building by utilizing sea water buyoancy

    International Nuclear Information System (INIS)

    Iwashima, Sumio; Ogoshi, Shigeru; Kobari, Shin-ichi.

    1989-01-01

    Upon dismantling nuclear reactor buildings, peripheral yards are excavated and channels leading to sea shore are formed. Since the outer walls of the reactor buildings are made of iron-reinforced concretes, the opening poritons are grouted with concretes to attain a tightly such closed structure that radioactive wastes, etc. in the inside are not flown out upon reactor discommisioning. Peripheral buildings at relatively low level of radiation contaminations are dismantled and withdrawn. The fundations of the nuclear reactor buildings were dug out and jacked to separate base rocks and the reactor buildings. Then, sea water is introduced into the water channels to entirely float up the buildings. A water gate is disposed in the water channel on the side of sea shore to control the level of sea water. The buildings are moved and guided to the sea shore and towed to a site optimum as a permanent storage area and then burried in that place. The operation period for the discommissioning work can greatly be shortened and the radiation dose and the amount of the wastes can be reduced. (T.M.)

  8. Earthquake response analysis of embedded reactor building considering soil-structure separation and nonlinearity of soil

    International Nuclear Information System (INIS)

    Ichikawa, T.; Hayashi, Y.; Nakai, S.

    1987-01-01

    In the earthquake response analysis for a rigid and massive structure as a nuclear reactor building, it is important to estimate the effect of soil-structure interaction (SSI) appropriately. In case of strong earthquakes, the nonlinearity, such as the wall-ground separation, the base mat uplift of sliding, makes the behavior of the soil-structure system complex. But, if the nuclear reactor building is embedded in a relatively soft ground with surface layer, the wall-ground separation plays the most important role in the response of soil-structure system. Because, it is expected that the base uplift and slide would be less significant due to the effect of the embedment, and the wall-ground friction is usually neglected in design. But, the nonlinearity of ground may have some effect on the wall-ground separation and the response of the structure. These problems have been studied by use of FEM. Others used joint elements between the ground and the structure which does not resist tensile force. Others studied the effect of wall-ground separation with non-tension springs. But the relationship between the ground condition and the effect of the separation has not been clarified yet. To clarify the effect the analyses by FE model and lumped mass model (sway-rocking model) are performed and compared. The key parameter is the ground profile, namely the stiffness of the side soil

  9. Coupling of impedance functions to nuclear reactor building for soil-structure interaction analysis

    International Nuclear Information System (INIS)

    Danisch, R.; Delinic, K.; Trbojevic, V.M.

    1991-01-01

    Finite element model of a nuclear reactor building is coupled to complex soil impedance functions and soil-structure-interaction analysis is carried out in frequency domain. In the second type of analysis applied in this paper, soil impedance functions are used to evaluate equivalent soil springs and dashpots of soil. These are coupled to the structure model in order to carry out the time marching analysis. Three types of soil profiles are considered: hard, medium and soft. Results of two analyzes are compared on the same structural model. Equivalent soil springs and dashpots are determined using new method based on the least square approximation. (author)

  10. Structure of pool in reactor building

    International Nuclear Information System (INIS)

    Yokoyama, Shigeki.

    1997-01-01

    Shielding walls made of iron-reinforced concrete having a metal liner including two body walls rigidly combined to the upper surface of a reactor container are disposed at least to one of an equipment pool or spent fuel storage pool in a reactor building. A rack for temporarily placing an upper lattice plate is detachably attached at least above one of a steam dryer or a gas/liquid separator temporarily placed in the temporary pool, and the height from the bottom portion to the upper end of the shielding wall is determined based on the height of an upper lattice plate temporary placed on the rack and the water depth required for shielding radiation from the upper lattice plate. An operator's exposure on the operation floor can be reduced by the shielding wall, and radiation dose from the spent fuels is reduced. The increase of the height of a pool guarder enhances bending resistance as a ceiling. In addition, the total height of them is made identical with the depth of the spent fuel storage pool thereby enabling to increase storage area for spent fuels. (N.H.)

  11. Elastic-plastic dynamic analysis of a reactor building

    International Nuclear Information System (INIS)

    Umemura, Hajime; Tanaka, Hiroshi.

    1976-01-01

    The basic characteristics of the dynamic response of a reactor building to severe earthquake ground motion are very important for the evaluation of the safety of nuclear plant systems. A computer program for elastic-plastic dynamic analysis of reactor buildings using lumped mass models is developed. The box and cylindrical walls of boiling water reactor buildings are treated as vertical beams. The nonlinear moment-rotation and shear force-shear deformation relationships of walls are based in part upon the experiments of prototype structures. The geometrical non-linearity of the soil rocking spring due to foundation separation is also considered. The nonlinear equation of motion is expressed in incremental form using tangent stiffness matrices, following the algorithm developed by E.L. Wilson et al. The damping matrix in the equation is formulated as the combination of the energy evaluation method and Penzien-Wilson's approach to accomodate the different characteristics of soil and building damping. The analysis examples and the comparison of elastic and elastic-plastic analysis results are presented. (auth.)

  12. Effect of modeling of super-structure on the behaviour of reactor building raft

    International Nuclear Information System (INIS)

    Mondal, A.; Singh, A.K.; Roy, Raghupati; Verma, U.S.P.; Warudkar, A.S.

    2003-01-01

    The behaviour of the reactor building raft was studied when the stiffness of the super-structural elements is included in the analysis as compared to the results of conventional analysis ignoring the stiffness of the super-structural elements. The effect of the stiffness of the super-structures on the loss of contact of the raft under seismic environment was also investigated. In order to study the effect of horizontal springs on the behaviour of the raft particularly near the stressing gallery under seismic environment, a separate study has been carried out considering a 3D model consisting of solid elements supported on both horizontal and vertical springs. The model was analysed for all the forces applied at the top of the raft and the analysis results were compared with those of shell model. The following conclusions are drawn: (i) Idealisation of the reactor building raft using shell elements is adequate for estimating the design forces/moments on the raft. The design forces/moments obtained from FE model consisting of solid elements closely matches with those obtained from FE model with shell elements. Idealisation of the RB raft using shell elements will also reduce the problem size and the related computational efforts. (ii) The stiffness of the super-structure has significant effect on the behaviour of the raft. Consideration of the stiffness of the super structure reduces the design forces/moments significantly and hence, modelling of the stiffness of the super structure is necessary for economical design. (iii) Modelling of horizontal stiffness of the raft in terms of horizontal springs at the interface of the raft and the rock does not have significant effect on the behaviour of the raft and as such, is not required to be considered in the FE model. However, it is necessary to ensure adequate factor of safety against the overall stability of the raft

  13. Response characteristics of reactor building on weathered soft rock ground

    International Nuclear Information System (INIS)

    Hirata, Kazuta; Tochigi, Hitoshi

    1991-01-01

    The purpose of this study is to investigate the seismic stability of nuclear power plants on layered soft bedrock grounds, focusing on the seismic response of reactor buildings. In this case, the soft bedrock grounds refer to the weathered soft bedrocks with several tens meter thickness overlaying hard bedrocks. Under this condition, there are two subjects regarding the estimation of the seismic response of reactor buildings. One is the estimation of the seismic response of surface ground, and another is the estimation of soil-structure interaction characteristics for the structures embedded in the layered grounds with low impedandce ratio between the surface ground and the bedrock. Paying attention to these subjects, many cases of seismic response analysis were carried out, and the following facts were clarified. In the soft rock grounds overlaying hard bedrocks, it was proved that the response acceleration was larger than the case of uniform hard bedrocks. A simplified sway and rocking model was proposed to consider soil-structure interaction. It was proved that the response of reactor buildings was small when the effect of embedment was considered. (K.I.)

  14. Reactor building

    International Nuclear Information System (INIS)

    Maruyama, Toru; Murata, Ritsuko.

    1996-01-01

    In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)

  15. 3-dimensional finite element modelling of reactor building internal structure for static analysis

    International Nuclear Information System (INIS)

    Joshi, M.H.; Reddy, V.J.; Kushwaha, H.S.; Reddy, G.R.; Karandikar, G.V.

    1991-01-01

    a) Thin shell element gives fairly accurate results when compared to 3-D Brick element for the type of structure and loading in Reactor Building. b) The maximum element size is fixed from model 3(c) i.e. 2.0 m. c) Openings with size smaller than 0.5 m can be neglected without affecting the results very much. d) For any such problem, the methodology described in this paper can be used to take rational decisions which will ensure reasonable accuracy. (author)

  16. Considerations on safety against seismic excitations in the project of reactor auxiliary building and control building in nuclear power plants

    International Nuclear Information System (INIS)

    Santos, S.H.C.; Castro Monteiro, I. de

    1986-01-01

    The seismic requests to be considered in the project of main buildings of a nuclear power plant are discussed. The models for global seismic analysis of nuclear power plant structures, as well as models for global strength distribution are presented. The models for analysing reactor auxiliary building and control building, which together with the reactor building and turbine building form the main energy generation complex in a nuclear power plant, are described. (M.C.K.) [pt

  17. Models test on dynamic structure-structure interaction of nuclear power plant buildings

    International Nuclear Information System (INIS)

    Kitada, Y.; Hirotani, T.

    1999-01-01

    A reactor building of an NPP (nuclear power plant) is generally constructed closely adjacent to a turbine building and other buildings such as the auxiliary building, and in increasing numbers of NPPs, multiple plants are being planned and constructed closely on a single site. In these situations, adjacent buildings are considered to influence each other through the soil during earthquakes and to exhibit dynamic behaviour different from that of separate buildings, because those buildings in NPP are generally heavy and massive. The dynamic interaction between buildings during earthquake through the soil is termed here as 'dynamic cross interaction (DCI)'. In order to comprehend DCI appropriately, forced vibration tests and earthquake observation are needed using closely constructed building models. Standing on this background, Nuclear Power Engineering Corporation (NUPEC) had planned the project to investigate the DCI effect in 1993 after the preceding SSI (soil-structure interaction) investigation project, 'model tests on embedment effect of reactor building'. The project consists of field and laboratory tests. The field test is being carried out using three different building construction conditions, e.g. a single reactor building to be used for the comparison purposes as for a reference, two same reactor buildings used to evaluate pure DCI effects, and two different buildings, reactor and turbine building models to evaluate DCI effects under the actual plant conditions. Forced vibration tests and earthquake observations are planned in the field test. The laboratory test is planned to evaluate basic characteristics of the DCI effects using simple soil model made of silicon rubber and structure models made of aluminum. In this test, forced vibration tests and shaking table tests are planned. The project was started in April 1994 and will be completed in March 2002. This paper describes an outline and the summary of the current status of this project. (orig.)

  18. Seismic analysis of a reactor building with eccentric layout

    International Nuclear Information System (INIS)

    Itoh, T.; Deng, D.Z.F.; Lui, K.

    1987-01-01

    Conventional design for a reactor building in a high seismic area has adopted an essentially concentric layout in response to fear of excessive torsional effect due to horizontal seismic load on an eccentric plant. This concentric layout requirement generally results in an inflexible arrangement of the plant facilities and thus increases the plant volume. This study is performed to investigate the effect of eccentricity on the overall seismic structural response and to provide technical information in this regard to substantiate the volume reduction of the overall power plant. The plant layout is evolved from the Bechtel standard plan of a PWR plant by integrating the reactor building and the auxiliary building into a combined building supported on a common basemat. This plant layout is optimized for volume utilization and to reduce the length of piping systems. The mass centers at various elevations of the combined building do not coincide with the rigidity center (RC) of the respective floor and the geometric center of the basemat, thus creating an eccentric response of the building in a seismic environment. Therefore, the torsional effects of the structure have to be taken into account in the seismic analysis

  19. Determination of the NPP Cernavoda reactor building seismic response

    International Nuclear Information System (INIS)

    Krutzik, N.J.; Rotaru, I.; Bobei, M.; Mingiuc, C.; Serban, V.

    1997-01-01

    Seismic input for systems and equipment installed in buildings depends on: - the seismic movement in free field on site; - the building movement in the soil; - the building deflection. The percentage of the 3 movements for the system and equipment input, depends on the position of the systems and equipment inside the building as well on the type of the foundation soil. The type of the foundation soil is important because if it is stiff it transfers a lot of energy to the building, energy which amplify the movement of the building on the top. If the foundation soil is soft, it accommodates the overall movement of the building in the soil, amplifying the movement to lower levels and the building response is attenuated if a resonance phenomenon between the whole building movement and the seismic excitation does not exist. This input is given with the design floor response spectra (FRS), in the logarithmic scale and seismic anchor movement (SAM). The design floor response spectra for NPP Cernavoda U1 Nuclear Building were determined in several stages starting with simple models (STICK type) without twisting movement and ending with detailed 3-dimensional models. From the point of view of dynamic behavior, the Reactor Building can be considered to be made up of 4 sub-structures: the containment building, internal structures containing separate elements such as the reactor vault, the fuel transfer structure and itself. Each sub-structure has its own movement (some of the structures present also some local effects) which combines with the overall movement of the building in the soil and the seismic excitation produce the total movement so that the response spectrum for each point of the sub-structure is specific. One should note that for structures which also show the twisting effect, the selection of the points on the floor, for the determination on the response spectra, is an engineering decision so that the response should be relevant for the equipment installed on the

  20. Industrial structure at research reactor suppliers

    International Nuclear Information System (INIS)

    Roegler, H.-J.; Bogusch, E.; Friebe, T.

    2001-01-01

    Due to the recent joining of the forces of Framatome S. A. from France and the Nuclear Division of Siemens AG Power Generation (KWU) from Germany to a Joint Venture named Framatome Advanced Nuclear Power S.A.S., the issue of the necessary and of the optimal industrial structure for nuclear projects as a research reactor is, was discussed internally often and intensively. That discussion took place also in the other technical fields such as Services for NPPs but also in the field of interest here, i. e. Research Reactors. In summarizing the statements of this presentation one can about state that: Research Reactors are easier to build than NPPs, but not standardised; Research Reactors need a wide spectrum of skills and experiences; to design and build Research Reactors needs an experienced team especially in terms of management and interfaces; Research Reactors need background from built reference plants more than from operating plants; Research Reactors need knowledge of suitable experienced subsuppliers. Two more essential conclusions as industry involved in constructing and upgrading research reactors are: Research Reactors by far are more than a suitable core that generates a high neutron flux; every institution that designs and builds a Research Reactor lacks quality or causes safety problems, damages the reputation of the entire community

  1. Measurement of Narora reactor building relative settlement

    International Nuclear Information System (INIS)

    Deo, P.M.; Pande, K.C.; Patwardhan, H.S.

    1977-01-01

    The civil construction of the reactor building of Narora Atomic Power Project has a special problem. The stability of the structure is liable to settlement as this location falls in seismic zone. To obviate the possibility of large scale unequal settlements, the reactor building is founded on a 4 meter thick rigid raft concreted in three layers, at a depth of 13 meters below ground. Stainless steel tanks will be embedded at 17 locations to measure relative settlements. The relative elevation difference will be detected by electrical probes when the water level in any one of the tanks touches the tip of the probes. The design envisages a maximum permissible unequal settlements of about 10 mm. over a period of 20 years. (K.B.)

  2. A seismic design of nuclear reactor building structures applying seismic isolation system in a seismicity region-a feasibility case study in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kubo, Tetsuo [The University of Tokyo, Tokyo (Japan); Yamamoto, Tomofumi; Sato, Kunihiko [Mitsubishi Heavy Industries, Ltd., Kobe (Japan); Jimbo, Masakazu [Toshiba Corporation, Yokohama (Japan); Imaoka, Tetsuo [Hitachi-GE Nuclear Energy, Ltd., Hitachi (Japan); Umeki, Yoshito [Chubu Electric Power Co. Inc., Nagoya (Japan)

    2014-10-15

    A feasibility study on the seismic design of nuclear reactor buildings with application of a seismic isolation system is introduced. After the Hyogo-ken Nanbu earthquake in Japan of 1995, seismic isolation technologies have been widely employed for commercial buildings. Having become a mature technology, seismic isolation systems can be applied to NPP facilities in areas of high seismicity. Two reactor buildings are discussed, representing the PWR and BWR buildings in Japan, and the application of seismic isolation systems is discussed. The isolation system employing rubber bearings with a lead plug positioned (LRB) is examined. Through a series of seismic response analyses using the so-named standard design earthquake motions covering the design basis earthquake motions obtained for NPP sites in Japan, the responses of the seismic isolated reactor buildings are evaluated. It is revealed that for the building structures examined herein: (1) the responses of both isolated buildings and isolating LRBs fulfill the specified design criteria; (2) the responses obtained for the isolating LRBs first reach the ultimate condition when intensity of motion is 2.0 to 2.5 times as large as that of the design-basis; and (3) the responses of isolated reactor building fall below the range of the prescribed criteria.

  3. On-site experimental dynamic analysis for evaluating the soil-structure interaction and the seismic behaviour of the Italian PEC fast reactor building

    International Nuclear Information System (INIS)

    Casirati, M.; Castoldi, A.; Panzeri, P.; Pezzoli, P.; Martelli, A.; Masoni, P.; Brancati, V.

    1988-01-01

    The paper describes the on-site dynamic tests carried out on the PEC fast reactor building, using various excitation methods (two eccentric back-rotating-mass mechanical vibrator, blasting in bore-hole, hydraulic actuators at the building foundations). It points out the purposes of the four tests campaigns performed at various construction stages and reports the main experimental results. These results concern both the design safety margins and the data for the validation of the three-dimensional numerical model of the reactor building, including soil-structure interaction phenomena. (author)

  4. Characterization of the Three Mile Island Unit-2 reactor building atmosphere prior to the reactor building purge

    International Nuclear Information System (INIS)

    Hartwell, J.K.; Mandler, J.W.; Duce, S.W.; Motes, B.G.

    1981-05-01

    The Three Mile Island Unit-2 reactor building atmosphere was sampled prior to the reactor building purge. Samples of the containment atmosphere were obtained using specialized sampling equipment installed through penetration R-626 at the 358-foot (109-meter) level of the TMI-2 reactor building. The samples were subsequently analyzed for radionuclide concentration and for gaseous molecular components (O 2 , N 2 , etc.) by two independent laboratories at the Idaho National Engineering Laboratory (INEL). The sampling procedures, analysis methods, and results are summarized

  5. Reactor building seismic analysis of a PWR type - NPP

    International Nuclear Information System (INIS)

    Kakubo, Masao

    1983-01-01

    Earthquake engineering studies raised up in Brazil during design licensing and construction phases of Almirante Alvaro Alberto NPP, units 1 and 2. State of art of soil - structure interaction analysis with particular reference to the impedance function calculation analysis with particular reference to the impedance function calculation of a group of pile is presented in this M.Sc. Dissertation, as an example the reactor building dynamic response of a 1325 MWe NPP PWR type is calculated. The reactor building is supported by a pile foundation with 2002 end bearing piles. Upper and lower bound soil parameters are considered in order to observe their influence on dynamic response of structure. Dynamic response distribution on pile heads show pile-soil-pile interaction effects. (author)

  6. Nonlinear seismic response analysis of an embedded reactor building based on the substructure approach

    International Nuclear Information System (INIS)

    Hasegawa, M.; Ichikawa, T.; Nakai, S.; Watanabe, T.

    1987-01-01

    A practical method to calculate the elasto-plastic seismic response of structures considering the dynamic soil-structure interaction is presented. The substructure technique in the time domain is utilized in the proposed method. A simple soil spring system with the coupling effects which are usually evaluated by the impedance matrix is introduced to consider the soil-structure interaction for embedded structures. As a numerical example, the response of a BWR-MARK II type reactor building embedded in the layered soil is calculated. The accuracy of the present method is verified by comparing its numerical results with exact solutions. The nonlinear behaivor and the soil-structure interaction effects on the response of the reactor building are also discussed in detail. It is concluded that the present method is effective for the aseismic design considering both the material nonlinearity of the nuclear reactor building and the dynamic soil-structure interaction. (orig.)

  7. Model tests and numerical analysis on restoring force characteristics of reactor buildings

    International Nuclear Information System (INIS)

    Uchiyama, Y.; Suzuki, S.; Akino, K.

    1987-01-01

    Seismic shear walls of nuclear reactor buildings are composed of cylindrical, truncated cone-shape, box-shape, irregular polygonal walls or its combination and they are generally heavily reinforced concrete (RC) walls. So the elasto-plastic behaviors of those RC structures in ultimate regions have many unsolved and may be considered as especially important factors for explaining nonlinear response of nuclear reactor buildings. Following these research demands, the authors have prepared a nonlinear F.E.M. code called ''SANREF'' and made an extensive study for the restoring force characteristics of the inner concrete structures (I/C) of a PWR-type containment vessel and the principal seismic shear walls of a BWR-type reactor building by some series of reduced model tests and simulation analysis for the tests results. The detailed objectives of this study can be summarized as follows: (1) Examine the effectiveness of the configurations of shear walls, reinforcement ratios, shear span ratios (M/Qd) and vertical axial stress by ''partial model test'' which simulates some independent shear walls of the PWR-type and BWR-type reactor buildings. (2) Obtain fundamental data of restoring force characteristics of the complex shaped RC structures by ''composite model test'' which models are composed of the partial model test specimens. (3) Verify the applicability of analytical methods and constitutive modelings in SANREF code for complex shaped RC structures through nonlinear simulation analysis for the composite model test

  8. Seismic response analysis of nuclear reactor buildings under consideration of soil-structure interaction with torsional behavior

    International Nuclear Information System (INIS)

    Mizuno, N.; Iida, T.; Tsushima, Y.; Araki, T.; Nojima, O.

    1977-01-01

    In this paper, the seismic response analysis is described in detail for estimating the soil-structure interaction effects with the torsional behavior. The analytical method is firstly shown for estimating the stiffness of reactor building by the bending-shear and torsion theory of the thin wall sections in regard to the behavior of structure. The three-dimensional behavior of structure can be obtained more briefly and simply by the proposed method. Secondly, the dynamical soil-foundation coefficient for estimating the dissipation of vibrational energy on the ground is derived by H. Tajimi's theory which is based on a solution of the propagation of seismic waves caused by point excitation on the surface of the elastic half-space medium. The above results give the vibrational impedances of the soil-foundation corresponding to the static soil coefficient, which is defined to the excitation force in the frequency domain. In order to analyze to the equivalues of reactor building, the authors thirdly attempt to approximate the dynamic soil-foundation coefficient as the frequency transfer function of displacement. The complex damping is used for more suitably estimating the elastic structural damping effects of structure. The regression analysis of many degrees of freedom is fourthly attempted for estimating the natural periods annd equivalent viscous damping ratios directly from the experimental results by the forced vibrational test performed in 1974. The analytical results are finally shown for simulating and comparing with the above-mentioned experimental results

  9. Study on reactor building structure using ultrahigh strength materials - Part 6: Tests for joints of SC-frames and PCa-panels

    International Nuclear Information System (INIS)

    Uchiyama, T.; Ishimura, K.; Takahashi, T.; Kei, T.

    1993-01-01

    A mixed structure composed of reinforced concrete precast panels and frames of steel beams and concrete filled steel tube columns using ultrahigh strength materials was proposed for reactor buildings. The paper describes the structural characteristics of the high tension bolt joints between the panels and the frames. (author)

  10. Effects of non-uniform embedments on earthquake responses of nuclear reactor building

    International Nuclear Information System (INIS)

    Koyanagi, Y.; Okamoto, S.; Yoshida, K.; Inove, H.

    1989-01-01

    The nuclear reactor buildings have the portion embedded in soil. In the seismic design of such structures, it is essential to consider the effects of the embedment on the earthquake response. Most studies on these effects, however, assume the uniform embedment, i.e. the depth of the embedment is constant, which is convenient for the design and analysis. The behavior of the earthquake response considering the three-dimensional aspects of non-uniform embedment has not been made clear yet. In this paper, the authors evaluate the effects of the non-uniform embedment in an inclined ground surface on the earthquake response of a nuclear reactor building as illustrated. A typical PWR type reactor building is chosen as an analysis structure model. Four different types of embedment are set up for the comparison study. The three-dimensional analysis is carried out considering the geometry of embedment

  11. Seismic simulation analysis of nuclear reactor building by soil-building interaction model

    International Nuclear Information System (INIS)

    Muto, K.; Kobayashi, T.; Motohashi, S.; Kusano, N.; Mizuno, N.; Sugiyama, N.

    1981-01-01

    Seismic simulation analysis were performed for evaluating soil-structure interaction effects by an analytical approach using a 'Lattice Model' developed by the authors. The purpose of this paper is to check the adequacy of this procedure for analyzing soil-structure interaction by means of comparing computed results with recorded ones. The 'Lattice Model' approach employs a lumped mass interactive model, in which not only the structure but also the underlying and/or surrounding soil are modeled as descretized elements. The analytical model used for this study extends about 310 m in the horizontal direction and about 103 m in depth. The reactor building is modeled as three shearing-bending sticks (outer wall, inner wall and shield wall) and the underlying and surrounding soil are divided into four shearing sticks (column directly beneath the reactor building, adjacent, near and distant columns). A corresponding input base motion for the 'Lattice Model' was determined by a deconvolution analysis using a recorded motion at elevation -18.5 m in the free-field. The results of this simulation analysis were shown to be in reasonably good agreement with the recorded ones in the forms of the distribution of ground motions and structural responses, acceleration time histories and related response spectra. These results showed that the 'Lattice Model' approach was an appropriate one to estimate the soil-structure interaction effects. (orig./HP)

  12. A SEISMIC DESIGN OF NUCLEAR REACTOR BUILDING STRUCTURES APPLYING SEISMIC ISOLATION SYSTEM IN A HIGH SEISMICITY REGION –A FEASIBILITY CASE STUDY IN JAPAN-

    Directory of Open Access Journals (Sweden)

    TETSUO KUBO

    2014-10-01

    Full Text Available A feasibility study on the seismic design of nuclear reactor buildings with application of a seismic isolation system is introduced. After the Hyogo-ken Nanbu earthquake in Japan of 1995, seismic isolation technologies have been widely employed for commercial buildings. Having become a mature technology, seismic isolation systems can be applied to NPP facilities in areas of high seismicity. Two reactor buildings are discussed, representing the PWR and BWR buildings in Japan, and the application of seismic isolation systems is discussed. The isolation system employing rubber bearings with a lead plug positioned (LRB is examined. Through a series of seismic response analyses using the so-named standard design earthquake motions covering the design basis earthquake motions obtained for NPP sites in Japan, the responses of the seismic isolated reactor buildings are evaluated. It is revealed that for the building structures examined herein: (1 the responses of both isolated buildings and isolating LRBs fulfill the specified design criteria; (2 the responses obtained for the isolating LRBs first reach the ultimate condition when intensity of motion is 2.0 to 2.5 times as large as that of the design-basis; and (3 the responses of isolated reactor building fall below the range of the prescribed criteria.

  13. Investigation of base isolation for fast breeder reactor building

    International Nuclear Information System (INIS)

    Morishita, M.; Kobatake, M.; Ohta, K.; Okada, Y.

    1989-01-01

    Achievement of great rationalization for seismic-resistant design of equipment system is necessary and indispensable from the viewpoints of economical and structural validity for a fast breeder reactor to be made practical. The method of reducing seismic loads on the building and equipment by application of base isolation may be an effective method, but in application to nuclear facilities, it will become necessary to examine the feasibility to actual design considering the severe seismic design requirements in Japan. With these considerations as the background, the authors carried out analytical studies from various viewpoints such as restoring force characteristics of base isolation device, influence of input earthquake motion, soil-structure interaction in base- isolated structure, etc. in case of providing base isolation system for a fast breeder reactor building. Based on these analytical studies, vibration tests on a base-isolated structure using a triaxial shaking table and simulation analyses of the tests were performed attempting to verify the effectiveness of the base isolation system and appropriateness of the analysis method. Results are presented

  14. Forced vibration tests and simulation analyses of a nuclear reactor building. Part 2: simulation analyses

    International Nuclear Information System (INIS)

    Kuno, M.; Nakagawa, S.; Momma, T.; Naito, Y.; Niwa, M.; Motohashi, S.

    1995-01-01

    Forced vibration tests of a BWR-type reactor building. Hamaoka Unit 4, were performed. Valuable data on the dynamic characteristics of the soil-structure interaction system were obtained through the tests. Simulation analyses of the fundamental dynamic characteristics of the soil-structure system were conducted, using a basic lumped mass soil-structure model (lattice model), and strong correlation with the measured data was obtained. Furthermore, detailed simulation models were employed to investigate the effects of simultaneously induced vertical response and response of the adjacent turbine building on the lateral response of the reactor building. (author). 4 refs., 11 figs

  15. Nuclear reactor melt-retention structure to mitigate direct containment heating

    Science.gov (United States)

    Tutu, Narinder K.; Ginsberg, Theodore; Klages, John R.

    1991-01-01

    A light water nuclear reactor melt-retention structure to mitigate the extent of direct containment heating of the reactor containment building. The structure includes a retention chamber for retaining molten core material away from the upper regions of the reactor containment building when a severe accident causes the bottom of the pressure vessel of the reactor to fail and discharge such molten material under high pressure through the reactor cavity into the retention chamber. In combination with the melt-retention chamber there is provided a passageway that includes molten core droplet deflector vanes and has gas vent means in its upper surface, which means are operable to deflect molten core droplets into the retention chamber while allowing high pressure steam and gases to be vented into the upper regions of the containment building. A plurality of platforms are mounted within the passageway and the melt-retention structure to direct the flow of molten core material and help retain it within the melt-retention chamber. In addition, ribs are mounted at spaced positions on the floor of the melt-retention chamber, and grid means are positioned at the entrance side of the retention chamber. The grid means develop gas back pressure that helps separate the molten core droplets from discharged high pressure steam and gases, thereby forcing the steam and gases to vent into the upper regions of the reactor containment building.

  16. Dynamic response of aircraft impact of a reactor building with protective shell on independent foundation

    International Nuclear Information System (INIS)

    Constantopoulos, I.V.; Vardanega, C.; Attalla, I.

    1981-01-01

    Aircraft impact loading can penalize significantly the design of the equipment in a conventional containment building. An alternative scheme was developed in an attempt to reduce the aircraft impact response. A preliminary study was carried out to investigate the feasibility of the alternative scheme. This study was made in such perspective and for the purpose of comparing the response to aircraft impact of a standard reactor building, to that of a reactor building having an independently founded outer shell. In the second scheme, the outer shell is meant to receive the aircraft impact, so that the load will be transmitted to the reactor building internals only by way of the structure-soil-structure system. In both cases, the aircraft impact was postulated to occur on a linear single degree of freedom oscillator which modeled, approximately, the plastification of the impact area. The soil was considered as a half-space with properties corresponding to a medium stiff soil, and modeled by lumped soil springs and dashpots. The reactor internals, inner shell and protective outer shell were modeled with beam elements and concentrated inertias. In modeling the coupled system, soil-structure interaction and structure-to-structure interaction through the soil were represented by a global stiffness matrix corresponding to the three degrees the freedom of each foundation, i.e. horizontal, vertical and rocking. (orig./HP)

  17. Seismic response of reactor building on alluvial soil by direct implicit integration

    International Nuclear Information System (INIS)

    Thakkar, S.K.; Dinkar, A.K.

    1983-01-01

    The evaluation of seismic response of a reactor building is a complex problem. A study has been made in this paper of seismic response of a reactor building by direct implicit integration method. The direct implicit integration methods besides being unconditionally stable have the merit of including response of higher modes without much effort. A reactor building consisting of external shell, internal shell, internals and raft is considered to be resting on alluvium. The complete building including the foundation is idealized by axisymmetric finite elements. The structure is analyzed separately for horizontal and vertical components of ground motion using harmonic analysis. Total response is found by superposition of two responses. The variation of several parameters, such as soil stiffness, embedment depth, inertia of foundation, viscous boundary and damping on seismic response is studied. The structural response is seen to depend significantly on the soil stiffness and damping. The seismic response is observed to be less sensitive to embedment depth and inertia of foundation. The vertical accelerations on the raft, boiler room floor slab and dome due to vertical ground motions are quite appreciable. The viscous boundary is seen to alter structural response in significantly compared to rigid boundaries in a larger mesh and its use appears to be promising in absorbing energy of body waves when used with direct implicit integration method. (orig.)

  18. Experimental and analytical studies for a BWR nuclear reactor building. Evaluation of soil-structure interaction behaviour

    International Nuclear Information System (INIS)

    Mizuno, N.; Tsushima, Y.

    1975-01-01

    This paper evaluates the spatial characteristics of dynamic properties, especially soil-structure interaction behaviour, of the BWR nuclear building by experimental and analytical studies. It is well known that the damping effects in soil-structure interaction are remarkable on the building with short periods by the dissipation of vibrational energy to the soil. The authors have previously reported an analytical method for estimating the damping effects the properties of which are characterized as follows: 1) The complex damping is used, because the so-called structural damping may be more suitable for estimating the damping effects of an elastic structure. 2) H. Tajimi's theory is used for estimating the dynamical soil-foundation stiffness with the dissipation of vibrational energy on the elastic half-space soil. In this paper, an approximate explanation is presented in regard to the more developmental mathematical method for estimating the damping effects than the above-mentioned previous method, which is 'Modes Superposition Method for Multi-Degrees of Freedom System' with the constant complex stiffness showing the structural damping effects and the dynamical soil-foundation stiffness approximated by the linear or quadratic functions of the eigenvalues. An approximate explanation is presented in regard to the experimental results of the No. 1 reactor building (BWR) of Hamaoka Nuclear Power Station, The Chubu Electric Power Co., Ltd. (Auth.)

  19. Pressure suppression device for nuclear reactor building

    International Nuclear Information System (INIS)

    Ikegame, Noboru.

    1992-01-01

    In a nuclear reactor building, there are disposed cooling coils connected to an air supply duct at the outside of the building, an air supply blower, an air supply duct having the top end opened, an exhaustion duct having the top end opened and a bypassing pipeline interposed between the exhaustion duct and the air supply duct on the side of the inlet of the cooling coils. In the reactor building, when a radioactive material leakage accident should occur, an isolation valve is closed to isolate the building from the outside. Further, bypassing isolation valve is opened to form a closed cooling circuit by the cooling coils, the air supply blower and the air supply duct, the exhaustion duct and the bypassing pipeline in the reactor building. With such a constitution, since air as the atmosphere in the building is circulated through the closed cooling circuit and cooled by the cooling coils, the temperature is not elevated. Accordingly, since the pressure elevation of the atmosphere in the building is suppressed, the atmosphere containing radioactive materials do not flow out of the building. (I.N.)

  20. Study on vertical seismic response model of BWR-type reactor building

    International Nuclear Information System (INIS)

    Konno, T.; Motohashi, S.; Izumi, M.; Iizuka, S.

    1993-01-01

    A study on advanced seismic design for LWR has been carried out by the Nuclear Power Engineering Corporation (NUPEC), under the sponsorship of the Ministry of International Trade and Industry (MITI) of Japan. As a part of the study, it has been investigated to construct an accurate analytical model of reactor buildings for a seismic response analysis, which can reasonably represent dynamic characteristics of the building. In Japan, vibration models of reactor buildings for horizontal ground motion have been studied and examined through many simulation analyses for forced vibration tests and earthquake observations of actual buildings. And now it is possible to establish a reliable horizontal vibration model on the basis of multi-lumped mass and spring model. However, vertical vibration models have not been so much studied as horizontal models, due to less observed data for vertical motions. In this paper, the vertical seismic response models of a BWR-type reactor building including soil-structure interaction effect are numerically studied, by comparing the dynamic characteristics of (1) three dimensional finite element model, (2) multi-stick lumped mass model with a flexible base-mat, (3) multi-stick lumped mass model with a rigid base-mat and (4) single-stick lumped mass model. In particular, the BWR-type reactor building has the long span truss roof which is considered to be one of the critical members to vertical excitation. The modelings of the roof trusses are also studied

  1. Hanford B Reactor Building Hazard Assessment Report

    International Nuclear Information System (INIS)

    Griffin, P. W.

    1999-01-01

    The 105-B Reactor (hereinafter referred to as B Reactor) is located in the 100 Area of the Hanford Site near Richland, Washington. The B Reactor is one of nine plutonium production reactors that were constructed in the 1940s during the Cold War Era. Construction of the B Reactor began June 7, 1943, and operation began on September 26, 1944. The Environmental Restoration Contractor was requested by RL to provide an assessment/characterization of the B Reactor building to determine and document the hazards that are present and could pose a threat to the environment and/or to individuals touring the building. This report documents the potential hazards, determines the feasibility of mitigating the hazards, and makes recommendations regarding areas where public tour access should not be permitted

  2. Experimental and analytical studies for a BWR nuclear reactor building evaluation of soil-structure interaction behavior

    International Nuclear Information System (INIS)

    Mizuno, N.; Tsushima, Y.

    1975-01-01

    The purpose of this paper is to evaluate the spatial characteristics of dynamic properties, especially soil-structure interaction behavior, or the BWR nuclear reactor building by experimental and analytical studies. An analytical method (SMIRT-1 Paper K 2/4) for estimating the damping effects is reported. The complex damping is used, because the so-called structural damping may be more suitable for estimating the damping effects of an elastic structure. H. Tajimi's theory is used for estimating the dynamical soil-foundation stiffness with the dissipation of vibrational energy on the elastic half-space soil. An approximate explanation is presented in regard to the more developmental mathematical method for estimating the damping effects than the above-mentioned previous method, which is 'Modes Superposition Method for Multi-Degrees of Freedom System' with the constant complex stiffness showing the structural damping effects and the dynamical soil-foundation stiffness approximated by the linear or quadratic functions of the eigenvalues. Next, an approximate explanation is presented in regard to the experimental results of the No.1 reactor building (BWR) of Hamaoka Nuclear Power Station, The Chubu Electric Power Co., Ltd. The regression analyses of the experimental resonance curves by one degree system show that the critical damping ratio is larger than the 0.10 used in the design for the fundamental natural period. It is attempted to simulate the experimental results by the above-mentioned method. The simulated model is a fourty-eight degrees of freedom spring mass system because of the eight masses for the eight floors including the base foundation and the six degrees of freedom for a mass

  3. Confirmatory Survey Results for the Reactor Building Dome Upper Structural Surfaces, Rancho Saco Nuclear Generating Station, Herald, California

    International Nuclear Information System (INIS)

    Wade C. Adams

    2006-01-01

    Results from a confirmatory survey of the upper structural surfaces of the Reactor Building Dome at the Rancho Seco Nuclear Generating Station (RSNGS) performed by the Oak Ridge Institute for Science and Education for the NRC. Also includes results of interlaboratory comparison analyses on several archived soil samples that would be provided by RSNGS personnel. The confirmatory surveys were performed on June 7 and 8, 2006

  4. Incorporating higher order WINKLER springs with 3-D finite element model of a reactor building for seismic SSI analysis

    International Nuclear Information System (INIS)

    Ermutlu, H.E.

    1993-01-01

    In order to fulfill the seismic safety requirements, in the frame of seismic requalification activities for NPP Muehleberg, Switzerland, detailed seismic analysis performed on the Reactor Building and the results are presented previously. The primary objective of the present investigation is to assess the seismic safety of the reinforced concrete structures of reactor building. To achieve this objective requires a rather detailed 3-D finite element modeling for the outer shell structures, the drywell, the reactor pools, the floor decks and finally, the basemat. This already is a complicated task, which enforces need for simplifications in modelling the reactor internals and the foundation soil. Accordingly, all internal parts are modelled by vertical sticks and the Soil Structure Interaction (SSI) effects are represented by sets of transitional and higher order rotational WINKLER springs, i.e. avoiding complicated finite element SSI analysis. As a matter of fact, the availability of the results of recent investigations carried out on the reactor building using diversive finite element SSI analysis methods allow to calibrate the WINKLER springs, ensuring that the overall SSI behaviour of the reactor building is maintained

  5. Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

    1989-10-01

    A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs

  6. Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

    1989-10-01

    A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs.

  7. Effect of elevated temperatures on heavy concrete structural strength in Qinshan phase 3 CANDU 6 reactor buildings

    International Nuclear Information System (INIS)

    Alikhan, S.; Khan, A.F.; Chen, S.

    2005-01-01

    Heavy concrete is commonly used inside the Qinshan Phase 3 CANDU 6 reactor buildings for radiation shielding functions in order to provide access to key areas during reactor operation. In some cases, the heavy concrete elements are also structural elements. Concerns have been raised about the functional performance of the heavy concrete structural elements, specifically the primary heat transport pump (PHTS) supporting slabs, surrounding the feeder cabinets when subjected to elevated temperatures between 42 degree C and 121 degree C and their corresponding temperature gradients on a long-term basis during the normal operation of the plant. This paper presents the results of a test investigation on the strength of heavy concrete under elevated temperature conditions being experienced by the heavy concrete structural elements around the feeder cabinet to confirm that these structural elements meet their functional requirements. The loading conditions consist subjecting the specimens to the elevated temperatures and temperature gradient noted during commissioning, including the effect of epoxy coating. The heavy concrete mix proportion and materials of the test samples (ilmenite aggregate and Portland cement) are identical to those used for heavy concrete structural elements surrounding the feeder cabinet. Subsequent to the confirmation of the functional requirements of the heavy concrete structural elements, alarm limits are recommended for these structural elements. (authors)

  8. Lightning protection system analysis at Multipurpose Reactor G A. Siwabessy building

    International Nuclear Information System (INIS)

    Teguh-Sulistyo

    2003-01-01

    Analysis to the part of lightning protection system at Multi Purpose Reactor GA Siwabessy (RSG-GAS) have been done. Observation examined the damage of some part of the earthing system caused by human error of chemically system. The analysis performed some assumptions and simulations to the points of lightning stroke. From this analysis obtained that the reactor building do not have vertical finial which can protect effectively to the whole reactor building and auxiliary building. Installing some new finials at some places are needed to protect building therefore the reactor building and auxiliary building well safe from lighting stroke

  9. Dynamic analysis of a reactor building on alluvial soil

    International Nuclear Information System (INIS)

    Arya, A.S.; Chandrasekaran, A.R.; Paul, D.K.; Warudkar, A.S.

    1977-01-01

    The reactor building consists of reinforced concrete internal framed structure enclosed in double containment shells of prestressed and reinforced concrete all resting on a common massive raft. The external cylindrical shell is capped by a spherical dome while the internal shell carries a cellular gird slab. The building is partially buried under ground. The soil consists of alluvial going to 1000 m depth. The site lies in a moderate seismic zone. The paper presents the dynamic analysis of the building including soil-structure interaction. The mathematical model consists of four parallel, suitably interconnected struxtures, namely inner containment, outer containment, internal frame and the calandria vault. Each one of the parallel structures consists of lumped-mass beam elements. The soil below the raft and on the sides of outer containment shell is represented by elastic springs in both horizontal and vertical directions. The various assumpions required to be made in developing the mathematical model are briefly discussed in the paper. (Auth.)

  10. Damage of reactor buildings occurred at the Fukushima Daiichi accident. Focusing on sequence leading to hydrogen explosions

    International Nuclear Information System (INIS)

    Naito, Masanori

    2011-01-01

    Fukushima Daiichi accident discharged enormous radioactive materials confined inside into the environment due to hydrogen explosions occurred at reactor buildings and forced many people to live the refugee life. This article described overview of Great East Japan Earthquake, specifications of Fukushima Daiichi nuclear power plants, sequence of plant status after earthquake occurrence and computerized simulation of plant behavior of Unit 1 leading to core melt and hydrogen explosion. Simulation results with estimated and assumed conditions showed water level decreased to bottom of reactor core after 4 hrs and 15 minutes passed, core melt started after 6 hrs and 49 minutes passed, failure of core support plate after 7 hrs and 18 minutes passed and through failure of penetration at bottom of pressure vessel after 7 hrs and 25 minutes passed. Hydrogen concentration at operating floor of reactor building of Unit 1 would be 15% accumulated and the pressure would amount to about 5 bars after hydrogen explosion if reactor building did not rupture with leak-tight structure. Since reactor building was not pressure-proof structure, walls of operating floor would rupture before 5 bars attained. (T. Tanaka)

  11. Nuclear Capacity Building through Research Reactors

    International Nuclear Information System (INIS)

    2017-01-01

    Four Instruments: •The IAEA has recently developed a specific scheme of services for Nuclear Capacity Building in support of the Member States cooperating research reactors (RR) willing to use RRs as a primary facility to develop nuclear competences as a supporting step to embark into a national nuclear programme. •The scheme is composed of four complementary instruments, each of them being targeted to specific objective and audience: Distance Training: Internet Reactor Laboratory (IRL); Basic Training: Regional Research Reactor Schools; Intermediate Training: East European Research Reactor Initiative (EERRI); Group Fellowship Course Advanced Training: International Centres based on Research Reactors (ICERR)

  12. Simulation of hydrogen deflagration and detonation in a BWR reactor building

    International Nuclear Information System (INIS)

    Manninen, M.; Silde, A.; Lindholm, I.; Huhtanen, R.; Sjoevall, H.

    2002-01-01

    A systematic study was carried out to investigate the hydrogen behaviour in a BWR reactor building during a severe accident. BWR core contains a large amount of Zircaloy and the containment is relatively small. Because containment leakage cannot be totally excluded, hydrogen can build up in the reactor building, where the atmosphere is normal air. The objective of the work was to investigate, whether hydrogen can form flammable and detonable mixtures in the reactor building, evaluate the possibility of onset of detonation and assess the pressure loads under detonation conditions. The safety concern is, whether the hydrogen in the reactor building can detonate and whether the external detonation can jeopardize the containment integrity. The analysis indicated that the possibility of flame acceleration and deflagration-to-detonation transition (DDT) in the reactor building could not be ruled out in case of a 20 mm 2 leakage from the containment. The detonation analyses indicated that maximum pressure spike of about 7 MPa was observed in the reactor building room selected for the analysis

  13. Earthquake response of nuclear reactor buildings deeply embedded in soil

    International Nuclear Information System (INIS)

    Masao, T.; Takasaki, Y.; Hirasawa, M.; Okajima, M.; Yamamoto, S.; Kawata, E.; Koori, Y.; Ochiai, S.; Shimizu, N.

    1980-01-01

    This paper is concerned with experimental and analytical studies to investigate dynamic behavior of deeply embedded structures such as nuclear reactor buildings. The principal points studied are as follows: (1) Examination of stiffness and radiation damping effects according to embedded depth, (2) verification for distributions of earth pressure according to embedded depth, (3) differences of response characteristics during oscillation according to embedded depth, and (4) proposal of an analytical method for seismic design. Experimental studies were performed by two ways: forced vibration test, and earthquake observation against a rigid body model embedded in soil. Three analytical procedures were performed to compare experimental results and to examine the relation between each procedure. Finally, the dynamic behavior for nuclear reactor buildings with different embedded depths were evaluated by an analytical method. (orig.)

  14. Ventilation system in the RA reactor building - design specifications

    International Nuclear Information System (INIS)

    Badrljica, R.

    1984-09-01

    Protective role of the ventilation system of nuclear facilities involve construction of ventilation barriers which prevent release of radioactive particulates or gases, elimination od radioactive particulates and gases from the air which is released from contaminated zones into the reactor environment. Ventilation barriers are created by dividing the building into a number of ventilation zones with different sub pressure compared to the atmospheric pressure. The RA reactor building is divided into four ventilation zones. First zone is the zone of highest risk. It includes reactor core with horizontal experimental channels, underground rooms of the primary coolant system (D 2 O), helium system, hot cells and the space above the the reactor core. Second zone is the reactor hall and the room for irradiated fuel storage. The third zone includes corridors in the basement, ground floor and first floor where the probability of contamination is small. The fourth zone includes the annex where the contamination risk is low. There is no have natural air circulation in the reactor building. Ventilators for air input and outlet maintain the sub pressure in the building (pressure lower than the atmospheric pressure). This prevents release of radioactivity into the atmosphere [sr

  15. Hydrogen behavior in a large-dry pressurized water reactor containment building during a severe accident

    International Nuclear Information System (INIS)

    Hsu Wensheng; Chen Hungpei; Hung Zhenyu; Lin Huichen

    2014-01-01

    Following severe accidents in nuclear power plants, large quantities of hydrogen may be generated after core degradation. If the hydrogen is transported from the reactor vessel into the containment building, an explosion might occur, which might threaten the integrity of the building; this can ultimately cause the release of radioactive materials. During the Fukushima Daiichi nuclear accident in 2011, the primary containment structures remained intact but contaminated fragments broke off the secondary containment structures, which disrupted mitigation activities and triggered subsequent explosions. Therefore, the ability to predict the behavior of hydrogen after severe accidents may facilitate the development of effective nuclear reactor accident management procedures. The present study investigated the behavior of hydrogen in a large-dry pressurized water reactor (PWR). The amount of hydrogen produced was calculated using the Modular Accident Analysis Program. The hydrogen transport behavior and the effect of the explosion on the PWR containment building were simulated using the Flame Acceleration Simulator. The simulation results showed that the average hydrogen volume fraction is approximately 7% in the containment building and that the average temperature is 330 K. The maximum predicted pressure load after ignition is 2.55 bar, which does not endanger the structural integrity of the containment building. The results of this investigation indicate that the hydrogen mitigation system should be arranged on both the upper and lower parts of the containment building to reduce the impact of an explosion. (author)

  16. Overview of the Westinghouse Small Modular Reactor building layout

    Energy Technology Data Exchange (ETDEWEB)

    Cronje, J. M. [Westinghouse Electric Company LLC, Centurion (South Africa); Van Wyk, J. J.; Memmott, M. J. [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the third in a series of four papers, which describe the design and functionality of the Westinghouse SMR. It focuses in particular upon the plant building layout and modular design of the Westinghouse SMR. In the development of small modular reactors, the building layout is an area where the safety of the plant can be improved by applying new design approaches. This paper will present an overview of the Westinghouse SMR building layout and indicate how the design features improve the safety and robustness of the plant. The Westinghouse SMR is designed with no shared systems between individual reactor units. The main buildings inside the security fence are the nuclear island, the rad-waste building, the annex building, and the turbine building. All safety related equipment is located in the nuclear island, which is a seismic class 1 building. To further enhance the safety and robustness of the design, the reactor, containment, and most of the safety related equipment are located below grade on the nuclear island. This reduces the possibility of severe damage from external threats or natural disasters. Two safety related ultimate heat sink (UHS) water tanks that are used for decay heat removal are located above grade, but are redundant and physically separated as far as possible for improved safety. The reactor and containment vessel are located below grade in the center of the nuclear island. The rad-waste and other radioactive systems are located on the bottom floors to limit the radiation exposure to personnel. The Westinghouse SMR safety trains are completely separated into four unconnected quadrants of the building, with access between quadrants only allowed

  17. Incoherent SSI Analysis of Reactor Building using 2007 Hard-Rock Coherency Model

    International Nuclear Information System (INIS)

    Kang, Joo-Hyung; Lee, Sang-Hoon

    2008-01-01

    Many strong earthquake recordings show the response motions at building foundations to be less intense than the corresponding free-field motions. To account for these phenomena, the concept of spatial variation, or wave incoherence was introduced. Several approaches for its application to practical analysis and design as part of soil-structure interaction (SSI) effect have been developed. However, conventional wave incoherency models didn't reflect the characteristics of earthquake data from hard-rock site, and their application to the practical nuclear structures on the hard-rock sites was not justified sufficiently. This paper is focused on the response impact of hard-rock coherency model proposed in 2007 on the incoherent SSI analysis results of nuclear power plant (NPP) structure. A typical reactor building of pressurized water reactor (PWR) type NPP is modeled classified into surface and embedded foundations. The model is also assumed to be located on medium-hard rock and hard-rock sites. The SSI analysis results are obtained and compared in case of coherent and incoherent input motions. The structural responses considering rocking and torsion effects are also investigated

  18. Effects of embedment including slip and separation on seismic SSI response of a nuclear reactor building

    International Nuclear Information System (INIS)

    Saxena, Navjeev; Paul, D.K.

    2012-01-01

    Highlights: ► Both the slip and separation of reactor base reduce with increase in embedment. ► The slip and separation become insignificant beyond 1/4 and 1/2 embedment respectively. ► The stresses in reactor reduce significantly upto 1/4 embedment. ► The stress reduction with embedment is more pronounced in case of tensile stresses. ► The modeling of interface is important beyond 1/8 embedment as stresses are underestimated otherwise. - Abstract: The seismic response of nuclear reactor containment building considering the effects of embedment, slip and separation at soil–structure interface requires modeling of the soil, structure and interface altogether. Slip and separation at the interface causes stress redistribution in the soil and the structure around the interface. The embedment changes the dynamic characteristics of the soil–structure system. Consideration of these aspects allows capturing the realistic response of the structure, which has been a research gap and presented here individually as well as taken together. Finite element analysis has been carried out in time domain to attempt the highly nonlinear problem. The study draws important conclusions useful for design of nuclear reactor containment building.

  19. Seismic stability analyses of various reactor buildings on quaternary deposit

    International Nuclear Information System (INIS)

    Takeuchi, Y.; Tsutagawa, M.; Asakura, S.; Katoh, T.; Tomura, H.; Uchiyama, S.; Koyama, M.; Oguro, E.; Akino, K.; Iizuka, S.; Hayashi, M.

    1993-01-01

    Many nuclear power plants have been built on Quaternary deposits in Europe and U.S.A., however, Japanese basic policy is to construct the reactor building and other auxiliary buildings on a bed rock which are important to safety, because large earthquakes are postulated to occur. Being limited bed rock sites in Japan, it has become necessary to increase possible place for nuclear power plant in order to cope with the middle and long term siting problems. For the purpose of establishing the draft of guideline on seismic design of reactor building on the Quaternary sand and gravel deposit in Japan, foundation soil stability and seismic resistance of the reactor building and plant equipment have been investigated and studied from 1983 to 1998. The studies have shown the following: 1) The response rotation angles of both common light weight basement (CL) and step basement (ES) plants during the earthquake reduce to 1/2 of the BR plant value, and the bearing pressure between the basement and the soil of improved plant are reduced as well; (2) every structure built on quaternary sand and gravel deposit, having 400m/s shear velocity, maintains enough seismic resistance, because the shear stress caused in the wall is small. The maximum shear strain of soil below the basemat of BR-BWR, which suffers the largest bearing pressure, is 1.1x10 -9 , but it can be said that the soil has enough stability according to the past soil tests for the Quaternary sand and gravel deposit that had been done by authors

  20. Prediction of hydrogen distribution in the reactor building in CANDU6 plant

    International Nuclear Information System (INIS)

    Jin, Y.; Song, Y.

    2008-01-01

    The CANDU plants have a lot of zircaloy. The fuel cladding, calandria tubes and pressure tubes are made of zircaloy. The zircaloy can be oxidized and hydrogen is generated during severe accident progression. The detonation or deflagration to detonation transition (DDT) due to hydrogen combustion may occur if the local hydrogen concentration or global hydrogen concentration exceeds certain value. The detonation may result in the rupture of the reactor building. The inside of the reactor building of CANDU plants is complex. So prediction of hydrogen distribution in the reactor building is important. This prediction is made using ISAAC code and GOTHIC code. ISAAC code partitioned the reactor building in to 7 compartments. GOTHIC code modeled the CANDU6 reactor building using 12 nodes. The hydrogen concentrations in the various compartments in the reactor building are compared. GOTHIC code slightly underpredicts hydrogen concentration in the F/M rooms than ISAAC code, but trend is same. The hydrogen concentration in the boiler room and the moderator room shows almost same as for both codes. (author)

  1. The Hanford Site N Reactor buildings task identification and evaluation of historic properties

    International Nuclear Information System (INIS)

    Stapp, D.C.; Marceau, T.E.

    1996-01-01

    The New Production Reactor complex at Hanford (hereafter referred to as N Reactor) is proposed to be deactivated, decommissioned, and demolished in the coming years. Recognizing that the Hanford Site has been important to the nation, state, and local community, a task was funded to examine the effects that these activities may have on the historic properties of N Reactor. The objectives of the N Reactor buildings task were to identify potential historic properties at N Reactor, to complete Historic Property Inventory forms for all structures considered eligible and ineligible for listing in the National Register of Historic Places, and to prepare a Memorandum of Agreement that identifies the measures required to mitigate any adverse effects

  2. Dynamic analysis of a reactor building on alluvial soil

    International Nuclear Information System (INIS)

    Arya, A.S.; Chandrasekaran, A.R.; Paul, D.K.

    1977-01-01

    The reactor building consists of reinforced concrete internal framed structure enclosed in double containment shells of prestressed and reinforced concrete all resting on a common massive raft. The external cylindrical shell is capped by a spherical dome while the internal shell carries a cellular grid slab. The building is partially buried under ground. The soil consists of alluvial going to 1000 m depth. The site lies in a moderate seismic zone. The paper presents the dynamic analysis of the building including soil-structure interaction. The mathematical model consists of four parallel, suitably interconnected structures, namely inner containment, outer containment, internal frame and the calandria vault. Each one of the parallel structures consists of lumped-mass beam elements. The soil below the raft and on the sides of outer containment shell is represented by elastic springs in both horizontal and vertical directions. The various assumptions required to be made in developing the mathematical model are briefly discussed in the paper. Transfer matrix technique has been used to determine the frequencies and mode shapes. The deformations due to bending, shear and effect of the rotary inertia have been included. Various alternatives of laterally interconnecting the internals and the shells have been examined and the best alternative from earthquake considerations has been obtained. In the study, the effect of internal structure flexibility and Calandria vault flexibility on the whole building have been studied. The resulting base raft motion and the structural timewise response of all floors have been determined for the design basis (safe shutdown) earthquake by mode superposition

  3. Venting krypton-85 from the Three Mile Island Unit 2 reactor building

    International Nuclear Information System (INIS)

    Burton, H.M.

    1981-01-01

    To permit the less restricted access to the reactor building necessary to maintain instrumentation and equipment, and to proceed towad the total decontamination of the facility, General Public Utilities, operators of the facility referred to hereafter as GPU, asked the United States Nuclear Regulatory Commission, or NRC, for permission to remove the 85 Kr from the reactor building by venting it to the environment. GPU supported their request with the Safety Analysis and Environmental Assessment Report on the proposed reactor building venting plan. On June 12, 1980, after seven months of licensing deliberations and numerous public hearings, the NRC granted GPU's request. The actual venting took place between June 28 and July 11, 1980. This report presents an overview of the detailed effort involved in the TMI-2 reactor building venting program. The findings reported here are condensed from a published report entitled TMI-2 Reactor Building Purge--Kr-85 Venting

  4. Earthquake-proof supporting structure in reactor vessel

    International Nuclear Information System (INIS)

    Sakurai, Akio; Sekine, Katsuhisa; Madokoro, Manabu; Katoono, Shin-ichi; Konno, Mutsuo; Suzuki, Takuro.

    1990-01-01

    Conventional earthquake-proof structure comprises a vessel vibration stopper integrated to a reactor vessel, powder for restricting the horizontal displacements, a safety vessel surrounds the outer periphery of the reactor vessel and a safety vessel vibration stopper integrated therewith, which are fixed to buildings. However, there was a problem that a great amount of stresses are generated in the base of the reactor vessel vibration stopper due to reaction of the powders which restrict thermal expansion. In order to remarkably reduce the reaction of the powers, powders are charged into a spaces formed between each of the reactor vessel vibration stopper, the safety vessel vibration stopper and the flexible member disposed between them. According to this constitution, the reactor vessel vibration stopper does not undergo a great reaction of the powers upon thermal expansion of the reactor vessel to moderate the generated stresses, maintain the strength and provide earthquake-proof supporting function. (N.H.)

  5. Evaluation for rigidity of box construction of nuclear reactor building

    International Nuclear Information System (INIS)

    Yamakawa, Tetsuo

    1979-01-01

    A huge box-shaped structure (hereafter, called box construction) of reinforced concrete is presently utilized as the reactor building structure in nuclear power plants. Evaluation of the rigidity of the huge box construction is required for making a vibration analysis model of nuclear reactor buildings. It is necessary to handle the box construction as the plates to which the force in plane is applied. This paper describes that the bending theory in elementary beam theory is equivalent to a peculiar, orthogonally anisotropic plate, the shearing rigidity and film rigidity in y direction of which are put to infinity and the Poisson's ratio is put to zero, viewed from the two-dimensional theory of elasticity. The form factor of 1.2 for shearing deformation in rectangular cross section was calculated from the parabolic distribution of shearing stress intensity, and it is the maximum value. The factor is equal to 1.2 for slender beams, but smaller than 1.2 for short and thick beams, having tendency to converge to 1.0. The non-conformity of boundary conditions regarding the shearing force at the both ends of cantilevers does not affect very seriously the evaluation of shearing rigidity. From the above results, it was found that the application of the theory to the box construction was able to give the rigidity evaluation with sufficient engineering accuracy. The theory can also be applied to the evaluation of tube type ultrahigh buildings. (Wakatsuki, Y.)

  6. Characteristics of Soil Structure Interaction for Reactor Building of Kashiwazaki-Kariwa Nuclear Power Plant

    International Nuclear Information System (INIS)

    Gil, Moon Joo; Jung, Rae Young; Hyun, Chang Hun; Kim, Moon Soo; Lim, Nam Hyoung

    2010-01-01

    On 16 July 2007, the Nigataken-chuetsu-oki earthquake registering a moment magnitude of 6.8 occurred at a depth of about 15 km. As a result of this earthquake, noticeable shaking exceeding the design ground motion was measured at the Tokyo Electric Power Company (TEPCO) Kashiwazaki-Kariwa Nuclear Power Station (KKN), the biggest nuclear power plant in the world, located at about 16 km away from the epicenter. This earthquake triggered a fire at an electrical transformer and insignificant damage on some parts of facilities. This event gave an impulse to study on the damage and safety margin of nuclear power plant due to the strong earthquake exceeding design basis. As a part of those efforts, KARISMA (KAshiwazaki-Kariwa Research Initiative for Seismic Margin Assessment) benchmark study was launched by the IAEA in terms of an international collaborative research. The main objectives of this research are to estimate the structural behavior and to evaluate the seismic margin of reactor building considering the effects of Soil-Structure Interaction (SSI). This paper presents verification of structural model developed here and validation of soil foundation characteristics through soil-column analysis. It has also been demonstrated that the spring constants and damping coefficient obtained from impedance analysis represent well the soil foundation characteristics

  7. Study on the hydrogen explosion risk at reactor building during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    JNES carried out analysis on the hydrogen mixing and explosion at reactor building with CFD code and explosion analysis code to evaluate what exactly has happened at the reactor buildings of the Fukushima Daiichi NPS. Based on the MELCOR severe accident analysis results of Fukushima Daiichi Unit 1 and Unit 3, sensitivity study using the CFD code FLUENT was carried out on the parameter of the release rate, total mass of hydrogen gas, the release path between reactor building and PCV, and so on. Then an analysis using AUTODYN code was carried out to investigate the explosion at the reactor building of Unit 4 as well as Unit 1 and, Unit 3. With those analysis results it became possible to estimate the leaked path and the total amount of leaked hydrogen gas from PCV to reactor building. (author)

  8. Building reactor operator sustain expert system with C language integrated production system

    International Nuclear Information System (INIS)

    Ouyang Qin; Hu Shouyin; Wang Ruipian

    2002-01-01

    The development of the reactor operator sustain expert system is introduced, the capability of building reactor operator sustain expert system is discussed with C Language Integrated Production System (Clips), and a simple antitype of expert system is illustrated. The limitation of building reactor operator sustain expert system with Clips is also discussed

  9. Assessment of the seismic resistance of a ventilation stack on a reactor building

    International Nuclear Information System (INIS)

    Makovicka, Daniel; Makovicka, Daniel

    2005-01-01

    The paper analyzes the seismic resistance of a ventilation stack on a reactor building, including the possible reserves of increasing the resistance. Structures of this type are highly sensitive to seismic loads, as the tuning of the stack (the spectrum of its lowest natural frequencies) corresponds with the frequency spectrum of excitation due to seismic effects. The purpose of the paper is to present an example of an actual structure to show the character of the response of the structure, and the participation of the individual frequency components of the response in the overall stress and strain state of a structure of this type. The methodology for a numerical analysis of the structure is also given. The load of the stack proper is modified by the transfer characteristics of the building. In engineering practice, the system is usually divided into two subsystems: the building with the sub-base, and the stack proper. The level of justification for the application of this simplification depends on the distance of the natural frequencies of the stack from the natural frequencies of the building. Finally, the paper deals with possible errors in determining the actual seismic resistance of the stack structure

  10. Response of a NPP reactor building under seismic action with regard to different soil properties

    International Nuclear Information System (INIS)

    Wagenknecht, E.

    1987-01-01

    The object of this investigation is the response of a reactor building on seismic action with systematic variation of the soil stiffness. A thin-walled orthotropic containment shell on varying heavy and rigid foundations is regarded as calculation model. The soil stiffness is simulated by meand of spring elements for horizontal translation and for rocking motions of the building. By the response spectra method the loads of the containment shell are calculated for a horizontal seismic excitation. The investigation is aimed at determining the influence of differentiated soil stiffnesses on the containment action effects and at recognizing the causes for the occuring effects. The results are thoroughly represented by selected quantities of the building's response, the effects from the soil-structure interaction are discussed and the causes of the effects cleary explained. Apossibility is provided for determining critical soil stiffnesses which cause a siginificat intensification effect. The results of the investigations show that both the soil stiffness and structural configuration of the reactor building particulary in case of the substructure being heavy and rigid, exert a decisive on the loading of the superstructure. (orig.)

  11. Radionuclide distribution in TMI-2 reactor building basement liquids and solids

    International Nuclear Information System (INIS)

    Horan, J.T.; McIsaac, C.V.; Keefer, D.G.

    1984-01-01

    As a result of the TMI-2 accident, approximately 2.46 x 10 6 L of contaminated water were released to the Reactor Building basement. The principal fission product release pathway from the damaged core was through the reactor coolant system (RCS) to the pressurizer, through the pressure-operated relief valve (PORV) on the pressurizer to the Reactor Coolant Drain Tank (RCDT), and then through the RCDT rupture disk to the Reactor Building basement. Since August 1979, a number of efforts have been made to determine the location, quantity, and composition of fission products released to the Reactor Building basement. These efforts have included sampling of the basement water and solids, the basement sump pump recirculation line, the RCDT, and visual surveys using a closed circuit television (CCTV) system. The analysis of basement samples has provided data on the physical and radioisotopic characteristics of the liquids and solids. This paper describes the sample collection techniques and discusses radiochemical analyses results

  12. Updating of a dynamic finite element model from the Hualien scale model reactor building

    International Nuclear Information System (INIS)

    Billet, L.; Moine, P.; Lebailly, P.

    1996-08-01

    The forces occurring at the soil-structure interface of a building have generally a large influence on the way the building reacts to an earthquake. One can be tempted to characterise these forces more accurately bu updating a model from the structure. However, this procedure requires an updating method suitable for dissipative models, since significant damping can be observed at the soil-structure interface of buildings. Such a method is presented here. It is based on the minimization of a mechanical energy built from the difference between Eigen data calculated bu the model and Eigen data issued from experimental tests on the real structure. An experimental validation of this method is then proposed on a model from the HUALIEN scale-model reactor building. This scale-model, built on the HUALIEN site of TAIWAN, is devoted to the study of soil-structure interaction. The updating concerned the soil impedances, modelled by a layer of springs and viscous dampers attached to the building foundation. A good agreement was found between the Eigen modes and dynamic responses calculated bu the updated model and the corresponding experimental data. (authors). 12 refs., 3 figs., 4 tabs

  13. Study of vibration analysis for nuclear reactor building

    International Nuclear Information System (INIS)

    Hirashima, Shin-ichi

    1978-01-01

    The mutual interference between the contiguous buildings with separate foundations and also that between the outer wall under the ground and the foundation bottom of the building were taken into consideration for the vibration analysis with spring-mass system. For two contiguous foundations of buildings it was attempted to represent the static mutual interference by a spring-mass system model. The theoretical analysis formulas are shown for the combination of the vertical movement and rocking motion, and for the interfering forces between the foundation and the outer wall of a building. The method of extending the model to dynamic one is explained. Several spring constants utilized in the analysis were obtained, for example, for mutual interference springs regarding vertical motion, mutual interfering springs for the foundation and the outer wall of a building and the mutual interference springs concerning horizontal movement. These models and analysis were applied to the BWR-MARK II-1100 MW nuclear reactor building and the contiguous turbine building. The structures and level relations of two buildings are shown, and the spring-mass system model for these buildings is expressed. The masses of about 20, the weights, the rotating inertia, the sectional moment of inertia, the spring constant and the damping coefficient for each mass are tabulated. As the results, the peak displacements occur at 2.556 Hz, 6.918 Hz, 10.43 Hz and 13.85 Hz. The damping coefficient is large and about 10 - 30% at the lower order modes. The calculated and the measured vibration characteristics for the BWR plant buildings are not much different, and this spring-mass system model is verified to be adequate. (Nakai, Y.)

  14. Fuel transporting device in nuclear reactor

    International Nuclear Information System (INIS)

    Inoue, Tatsumi.

    1975-01-01

    Object: To obtain a support structure of an excellent quakeproof property for a fuel transporting device provided for the transportation of fuel between a reactor building and an auxiliary building in a pressure tube reactor or the like. Structure: The structure comprises an oblique transfer chute loosely penetrating the reactor building, reactor container and auxiliary building, a transfer chute support outer cylinder surrounding the transfer chute and having one end coupled to the transfer chute and other end coupled to the container, flexible seal members respectively provided on the reactor building side and on the auxiliary building side and surrounding the transfer chute and a slidable support supported on the side of the auxiliary building such that it can be in frictional contact with the outer periphery of the transfer chute. With this construction, the relative displacements of various parts caused by an earthquake or the like can be absorbed by the support outer cylinder, flexible seals and slidable support. (Ikeda, J.)

  15. Earthquake proof device for nuclear power plant building

    International Nuclear Information System (INIS)

    Okada, Yasuo.

    1991-01-01

    The structure of the present invention enables three dimensional vibration proof, i.e., in horizontal and vertical directions of a reactor container building. That is, each of the reactor container building and a reactor auxiliary building is adapted as an independent structure. The periphery of the reactor container building is surrounded by the reactor auxiliary building. The reactor auxiliary building is supported against the ground by way of a horizontal vibration proof device. The reactor container building is supported against the ground by way of a three-dimensional vibration proof device that prevents vibrations in both of the horizontal directions, and the vertical directions. The reactor container building is connected to the auxiliary building by way of a vertical vibration proof device. With such a constitution, although the reactor container building is vibration proof against both of the horizontal and the vertical vibrations, the vertical vibration proofness is an extension of inherent vertical vibration period. Accordingly, the head of the building undergoes rocking vibrations. However, since the reactor container building is connected to the reactor auxiliary building, the rocking vibrations are prevented by the reactor auxiliary building. As a result, safety upon occurrence of an earthquakes can be ensured. (I.S.)

  16. A Wireless Monitoring System for Cracks on the Surface of Reactor Containment Buildings.

    Science.gov (United States)

    Zhou, Jianguo; Xu, Yaming; Zhang, Tao

    2016-06-14

    Structural health monitoring with wireless sensor networks has been increasingly popular in recent years because of the convenience. In this paper, a real-time monitoring system for cracks on the surface of reactor containment buildings is presented. Customized wireless sensor networks platforms are designed and implemented with sensors especially for crack monitoring, which include crackmeters and temperature detectors. Software protocols like route discovery, time synchronization and data transfer are developed to satisfy the requirements of the monitoring system and stay simple at the same time. Simulation tests have been made to evaluate the performance of the system before full scale deployment. The real-life deployment of the crack monitoring system is carried out on the surface of reactor containment building in Daya Bay Nuclear Power Station during the in-service pressure test with 30 wireless sensor nodes.

  17. Seismic safety of building structures of NPP Kozloduy III

    International Nuclear Information System (INIS)

    Varbanov, G.I.; Kostov, M.K.; Stefanov, D.D.; Kaneva, A.D.

    2005-01-01

    In the proposed paper is presented a general summary of the analyses carried out to evaluate the dynamic behavior and to assess the seismic safety of some safety related building structures of NPP Kozloduy. The design seismic loads for the site of Kozloduy NPP has been reevaluated and increased during and after the construction of investigated Units 5 and 6. Deterministic and probabilistic approaches are applied to assess the seismic vulnerability of the investigated structures, taking into account the newly defined seismic excitations. The presented results show sufficient seismic safety for the studied critical structures and good efficiency of the seismic upgrading. The applicability of the investigated structures at sites with some higher seismic activities is discussed. The presented study is dealing mainly with the civil structures of the Reactor building, Turbine hall, Diesel Generator Station and Water Intake Structure. (authors)

  18. Reactor core structure

    International Nuclear Information System (INIS)

    Higashinakagawa, Emiko; Sato, Kanemitsu.

    1992-01-01

    Taking notice on the fact that Fe based alloys and Ni based alloys are corrosion resistant in a special atmosphere of a nuclear reactor, Fe or Ni based alloys are applied to reactor core structural components such as fuel cladding tubes, fuel channels, spacers, etc. On the other hand, the neutron absorption cross section of zirconium is 0.18 barn while that of iron is 2.52 barn and that of nickel is 4.6 barn, which amounts to 14 to 25 times compared with that of zirconium. Accordingly, if the reactor core structural components are constituted by the Fe or Ni based alloys, neutron economy is lowered. Since it is desirable that neutrons contribute to uranium fission with least absorption to the reactor core structural components, the reactor core structural components are constituted with the Fe or Ni based alloys of good corrosion resistance only at a portion in contact with reactor water, that is, at a surface portion, while the main body is constituted with zircalloy in the present invention. Accordingly, corrosion resistnace can be kept while keeping small neutron absorption cross section. (T.M.)

  19. Cassette blanket and vacuum building: key elements in fusion reactor maintenance

    International Nuclear Information System (INIS)

    Werner, R.W.

    1977-01-01

    The integration of two concepts important to fusion power reactors is discussed. The first concept is the vacuum building which improves upon the current fusion reactor designs. The second concept, the use of the cassette blanket within the vacuum building environment, introduces four major improvements in blanket design: cassette blanket module, zoning concept, rectangular blanket concept, and internal tritium recovery

  20. Assessment of structural materials inside the reactor pool of the Dalat research reactor

    International Nuclear Information System (INIS)

    Nguyen Nhi Dien; Luong Ba Vien; Nguyen Minh Tuan; Trang Cao Su

    2010-01-01

    Originally the Dalat Nuclear Research Reactor (DNRR) was a 250-kW TRIGA MARK II reactor, started building from early 1960s and achieved the first criticality on February 26, 1963. During the 1982-1984 period, the reactor was reconstructed and upgraded to 500kW, and restarted operation on March 20, 1984. From the original TRIGA reactor, only the pool liner, beam ports, thermal columns, and graphite reflector have been remained. The structural materials of pool liner and other components of TRIGA were made of aluminum alloy 6061 and aluminum cladding fuel assemblies. Some other parts, such as reactor core, irradiation rotary rack around the core, vertical irradiation facilities, etc. were replaced by the former Soviet Union's design with structural materials of aluminum alloy CAV-1. WWR-M2 fuel assemblies of U-Al alloy 36% and 19.75% 235 U enrichment and aluminum cladding have been used. In its original version, the reactor was setting upon an all-welded aluminum frame supported by four legs attached to the bottom of the pool. After the modification made, the new core is now suspended from the top of the pool liner by means of three aluminum concentric cylindrical shells. The upper one has a diameter of 1.9m, a length of 3.5m and a thickness of 10mm. This shell prevents from any visual access to the upper part of the pool liner, but is provided with some holes to facilitate water circulation in the 4cm gap between itself and the reactor pool liner. The lower cylindrical shells act as an extracting well for water circulation. As reactor has been operated at low power of 500 kW, it was no any problem with degradation of core structural materials due to neutron irradiation and thermal heat, but there are some ageing issues with aluminum liner and other structures (for example, corrosion of tightening-up steel bolt in the fourth beam port and flood of neutron detector housing) inside the reactor pool. In this report, the authors give an overview and assessment of

  1. On the response of a reactor building and its equipment to aircraft crash

    International Nuclear Information System (INIS)

    Larsson, G.; Lundsager, P.

    1977-01-01

    The present study investigates the dynamic response of the ASEA-ATOM BWR 75 reactor building in terms of response spectra at significant locations considering various aircraft and points of load application. In the first part of the study a total of 21 forcing functions, most of them from the open literature and including the commonly used standard functions, have been studied with respect to documentation, consistency and frequency content. Since none of the forcing functions have been experimentally verified, their validity must be assessed mainly by judging the structural models and assumptions used in their derivation and by checking their consistency. In the second part, linear dynamical models of various degrees of detailedness have been investigated regarding their capacity to describe the behavior of the reactor building under this high frequency loading. The most detailed model consists of plane stress finite elements for every significant wall and floor. In the third part of the study the effects of a number of parameters on the response of the building are investigated. The parameters include the points of attack, damping values, soil spring stiffness as well as different forcing functions of various frequency contents. The reponse is displayed as response spectra and member forces for characteristic locations. The results serve as a basis for development of standardized design floor response spectra and for the structural verification of the bui

  2. Behavior and ultimate strength of an inner concrete structure of a nuclear reactor building subjected to thermal and seismic loads

    International Nuclear Information System (INIS)

    Omatsuzawa, K.; Suzuki, Y.; Sato, M.; Takeda, T.; Yamaguchi, T.; Yoshioka, K.; Nakayama, T.; Furuya, N.; Kawaguchi, T.; Koike, K.; Naganuma, K.

    1987-01-01

    Heating tests and heating-plus-seismic-loading tests at high temperature (T max = 175 0 C) were conducted using various concrete structural members such as beams, cylindrical walls, H-section walls, and 1/10-scale models of the inner concrete (I/C) structure in a fast breeder reactor (FBR) building. Concrete subjected to high temperature exceeding 100 0 C has a tendency to have lower Young's modulus and to shrink. As these material constants are temperature-dependent, the thermal stress occurring within the concrete structure is smaller than the values usually obtained by normal crack analysis methods. Although thermal stresses and cracks exert marked influences on the behaviors of the structures during the earlier stages of loading, they hardly affect the ultimate bending and shear strengths. Specifically, as a result of I/C model tests, it was made clear that the ultimate strength of the structure is considerably greater than the design loads under combined thermal and seismic loading conditions. (orig./HP)

  3. Study of seismic responses of Candu-3 reactor building using isolator bearings

    International Nuclear Information System (INIS)

    Biswas, J.K.

    1992-01-01

    Seismic isolator bearings are known to increase reliability, reduce cost and increase the potential sitings for nuclear power plants located in regions of high seismicity. High seismic activities in Canada occur mainly in the western coast, the Grand Banks and regions of Quebec along the St. Lawrence river. In Canada, nuclear power plants are located in Ontario, Quebec and New Brunswick where the seismicity levels are low to moderate. Consequently, seismic isolator bearings have not been used in the existing nuclear power plants in Canada. The present paper examines the effect of using seismic isolator bearings in the design for the new CANDU3 which would be suitable for regions having high seismicity. The CANDU3 Nuclear Power Plant is rated at 450 MW of net output power and is a smaller version of its predecessor CANDU6 successfully operating in Canada and abroad. The design of CANDU3 is being developed by AECL CANDU. Advanced technologies for design, construction and plant operation have been utilized. During the conceptual development of the CANDU3 design, various design options including the use of isolator bearings were considered. The present paper presents an overview of seismic isolation technology and summarizes the analytical work for predicting the seismic behavior of the CANDU3 reactor building. A lumped-parameter dynamic model for the reactor building is used for the analysis. The characteristics of the bearings are utilized in the analysis work. The time-history modal analysis has been used to compute the seismic responses. Seismic responses of the reactor building with and without isolator bearings are compared. The isolator bearings are found to reduce the accelerations of the reactor building. As a result, a lower level of seismic qualification for components and systems would be required. The use of these bearings however increases rigid body seismic displacements of the structure requiring special considerations in the layout and interfaces for

  4. RA reactor building and installations; Zgrada 'RA' i instalacije

    Energy Technology Data Exchange (ETDEWEB)

    Badrljica, R; Sanovic, V; Skoric, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1985-08-15

    RA reactor building is made of reinforced concrete and bricks. It is closed facility with a limited number of controlled openings, doors and windows. The site of the building is 100 m above the sea level, 20 m above the mean Danube level and 8 m above the level of the neighbouring stream Mlaka. The building consists of three parts: central prismatic part, annex - surrounding the central part and the sanitary corridor. The biggest space is the reactor hall. In addition to the detailed description and drawings of the reactor building this documents includes design specifications of: electrical installation, water supply system, sewage system, ventilation and heating, gas and compressed air systems. A separate chapter is devoted to fire protection. Zgrada reaktora RA izgradjena je od armiranog betona i opeke, kao zatvoreni objekat ogranicenog broja kontolisanih otvora, sa ogranicenim brojem vrata i prozora. Plato na kojem je zgrada izgradjena nalazi se na 100 m nadmorske visine, na 20 m iznad srednjeg vodostaja Dunava i 8 m iznad nivoa obliznjeg potoka Mlaka. Zgrada se sastoji iz tri dela: sredisnjeg prizmaticnog dela, aneksa - prstenastog okvira sredisnog dela i sanitarnog propusnika. Pojedinacno najveci prostor zauzima reaktorska hala. Pored detaljnog opisa i plana zgrade, ovaj dokument sadrzi projekat elektricne instalacije, projekat vodovoda i kanalizacije, ventilacije i grejanja, instalacije gasa i komprimovanog vazduha. Posebno poglavlje posveceno je protivpozarnoj zastiti.

  5. Numerical and on-site experimental dynamic analysis of the Italian PEC fast reactor building

    International Nuclear Information System (INIS)

    Castoldi, A.; Muzzi, F.; Orsi, R.; Panzeri, P.; Pezzoli, P.; Ruggeri, G.; Martelli, A.; Masoni, P.; Brancati, V.

    1988-01-01

    On-site dynamic tests and three-dimensional numerical analysis have been performed by ISMES on behalf of ENEA on the building of the Italian PEC fast reactor test facility. These studies aimed at evaluating the safety margins in the PEC reactor seismic analysis and at providing data for the optimization of the PEC seismic monitoring system. The paper describes the on-site dynamic tests carried out using various excitation methods (two eccentric back-rotating-mass mechanical vibrator, blasting in bore-hole and hydraulic actuators at the building foundations). It highlights the purposes of the four tests campaigns performed at various construction stages and reports the main experimental results. In connection with the experimental tests, a detailed 3D finite element model was set up for fixed base analysis; from the results of the 3D model a simplified equivalent model of the structure was then derived for soil-structure interaction analysis. The mathematical model was validated and calibrated by using the results of the experimental dynamic tests. The main numerical results and the comparisons with the experimental data are presented. (author)

  6. Computed versus measured response of HDR reactor building in large scale shaking tests

    International Nuclear Information System (INIS)

    Werkle, H.; Waas, G.

    1987-01-01

    The earthquake resistant design of NPP structures and their installations is commonly based on linear analysis methods. Nonlinear effects, which may occur during strong earthquakes, are approximately accounted for in the analysis by adjusting the structural damping values. Experimental investigations of nonlinear effects were performed with an extremely heavy shaker at the decommissioned HDR reactor building in West Germany. The tests were directed by KfK (Nuclear Research Center Karlsruhe, West Germany) and supported by several companies and institutes from West Germany, Switzerland and the USA. The objective was the dynamic repsonse behaviour of the structure, piping and components to strong earthquake-like shaking including nonlinear effects. This paper presents some results of safety analyses and measurements, which were performed prior and during the test series. It was intended to shake the building up to a level where only a marginal safety against global structural failure was left

  7. Effects of different SSI parameters on the floor response spectra of a nuclear reactor building

    International Nuclear Information System (INIS)

    Kabir, A.F.; Bolourchi, S.; Maryak, M.E.

    1991-01-01

    The effects of several critical soil-structure interaction (SSI) parameters on the floor response spectra (FRS) of a typical nuclear reactor building have been examined. These parameters are computation of soil impedance functions using different approaches, scattering effects (reductions in ground motion due to embedment and rigidity of building foundation) and strain dependency of soil dynamic properties. This paper reports that the significant conclusions of the study, which are applicable to a deeply embedded very rigid nuclear reactor building, are as follows: FRS generated without considering scattering effects are highly conservative; differences between FRS, generated considering strain-dependency of soil dynamic properties, and those generated suing low-strain values, are not significant; and the lumped-parameter approach of SSI calculations, which only uses a single value of soil shear modulus in impedance calculations, may not be able to properly compute the soil impedances for a soil deposit with irregularly varying properties with depth

  8. RESRAD-Build: A model to estimate dose from contaminated structures. Innovative technology summary report

    International Nuclear Information System (INIS)

    1998-12-01

    The RESRAD-BUILD model is an exposure pathway and analysis code used to determine whether radiologically contaminated buildings and structures can be free released for a specific land use (e.g., residential or industrial). The model provides estimates of dose to a hypothetical receptor from the structure. The RESRAD-BUILD technology can calculate dose from variety of site-specific hypothetical scenarios, decay-time intervals, and radionuclides. When using the RESRAD-BUILD code, specific project assumptions must be developed with the appropriate regulatory agencies, especially the cleanup criteria and the exposure scenario to be used. The C Reactor demonstration of RESRAD-BUILD modeled hypothetical future use of below grade portions of the reactor building complex. A residential exposure scenario with a cleanup criteria of 15 mrem/yr above background (Environmental Protection Agency [EPA] draft guidance) was used to coordinate decommissioning with adjacent ongoing remedial actions conducted in accordance with an existing Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA) Record of Decision. This paper gives a description of the technology and discusses its performance, applications, cost, regulatory and policy issues, and lessons learned

  9. RESRAD-BUILD: A model to estimate dose from contaminated structures. Innovative technology summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-12-01

    The RESRAD-BUILD model is an exposure pathway and analysis code used to determine whether radiologically contaminated buildings and structures can be free released for a specific land use (e.g., residential or industrial). The model provides estimates of dose to a hypothetical receptor from the structure. The RESRAD-BUILD technology can calculate dose from variety of site-specific hypothetical scenarios, decay-time intervals, and radionuclides. When using the RESRAD-BUILD code, specific project assumptions must be developed with the appropriate regulatory agencies, especially the cleanup criteria and the exposure scenario to be used. The C Reactor demonstration of RESRAD-BUILD modeled hypothetical future use of below grade portions of the reactor building complex. A residential exposure scenario with a cleanup criteria of 15 mrem/yr above background (Environmental Protection Agency [EPA] draft guidance) was used to coordinate decommissioning with adjacent ongoing remedial actions conducted in accordance with an existing Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA) Record of Decision. This paper gives a description of the technology and discusses its performance, applications, cost, regulatory and policy issues, and lessons learned.

  10. Dismantling method for reactor pressure vessel and system therefor

    International Nuclear Information System (INIS)

    Hayashi, Makoto; Enomoto, Kunio; Kurosawa, Koichi; Saito, Hideyo.

    1994-01-01

    Upon dismantling of a reactor pressure vessel, a containment building made of concretes is disposed underground and a spent pressure vessel is contained therein, and incore structures are contained in the spent pressure vessel. Further, a plasma-welder and a pressing machine are disposed to a pool for provisionally placing reactor equipments in the reactor building for devoluming the incore structures by welding and compression. An overhead-running crane and rails therefor are disposed on the roof and the outer side of the reactor building for transporting the pressure vessel from the reactor building to the containment building. They may be contained in the containment building after incorporation of the incore structures into the pressure vessel at the outside of the reactor building. For the devoluming treatment, a combination of cutting, welding, pressing and the like are optically conducted. A nuclear power plant can be installed by using a newly manufactured nuclear reactor, with no requirement for a new site and it is unnecessary to provide a new radioactive waste containing facility. (N.H.)

  11. Seismic resistance design of nuclear power plant building structures in Japan

    International Nuclear Information System (INIS)

    Kitano, Takehito

    1997-01-01

    Japan is one of the countries where earthquakes occur most frequently in the world and has incurred a lot of disasters in the past. Therefore, the seismic resistance design of a nuclear power plant plays a very important role in Japan. This report describes the general method of seismic resistance design of a nuclear power plant giving examples of PWR and BWR type reactor buildings in Japan. Nuclear facilities are classified into three seismic classes and is designed according to the corresponding seismic class in Japan. Concerning reactor buildings, the short-term allowable stress design is applied for the S1 seismic load and it is confirmed that the structures have a safety margin against the S2 seismic load. (J.P.N.)

  12. Seismic resistance design of nuclear power plant building structures in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kitano, Takehito [Kansai Electric Power Co., Inc., Osaka (Japan)

    1997-03-01

    Japan is one of the countries where earthquakes occur most frequently in the world and has incurred a lot of disasters in the past. Therefore, the seismic resistance design of a nuclear power plant plays a very important role in Japan. This report describes the general method of seismic resistance design of a nuclear power plant giving examples of PWR and BWR type reactor buildings in Japan. Nuclear facilities are classified into three seismic classes and is designed according to the corresponding seismic class in Japan. Concerning reactor buildings, the short-term allowable stress design is applied for the S1 seismic load and it is confirmed that the structures have a safety margin against the S2 seismic load. (J.P.N.)

  13. Decontamination and radioactivity measurement on building surfaces related to dismantling of Japan power demonstration reactor (JPDR)

    International Nuclear Information System (INIS)

    Hatakeyama, Mutsuo; Tachibana, Mitsuo; Yanagihara, Satoshi

    1997-12-01

    In the final stage of dismantling activities for decommissioning a nuclear power plant, building structures have to be demolished to release the site for unrestricted use. Since building structures are generally made from massive reinforced concrete materials, it is not a rational way to treat all concrete materials arising from its demolition as radioactive waste. Segregation of radioactive parts from building structures is therefore indispensable. The rational procedures were studied for demolition of building structures by treating arising waste as non-radioactive materials, based on the concept established by Nuclear Safety Commission, then these were implemented in the following way by the JPDR dismantling demonstration project. Areas of the JPDR facilities are categorized into two groups : possibly contaminated areas, and possibly non-contaminated areas, based on the document of the reactor operation. Radioactivity on the building surfaces was then measured to confirm that the qualitative categorization is reasonable. After that, building surfaces were decontaminated in such a way that the contaminated layers were removed with enough margin to separate radioactive parts from non-radioactive building structures. Thought it might be possible to demolish the building structures by treating arising waste as non-radioactive materials, confirmation survey for radioactivity was conducted to show that there is no artificial radioactive nuclides produced by operation in the facility. This report describes the procedures studied on measurement of radioactivity and decontamination, and the results of its implementation in the JPDR dismantling demonstration project. (author)

  14. SPECIFICATIONS FOR HIGH TEMPERATURE LATTICE TEST REACTOR BUILDING 318 PROJECT CAH-100

    Energy Technology Data Exchange (ETDEWEB)

    Vitro Engineering Company

    1964-07-15

    This is the specifications for the High Temperature Lattice Test Reactor Building 318 and it is divided into the following 21 divisions or chapters: (1) Excavating, Backfill & Grading; (2) Reinforced Concrete; (3) Masonry; (4) Structural Steel & Miscellaneous Metal Items, Contents - Division V; (5) Plumbing, Process & Service Piping; (6) Welding; (7) Insulated Metal Siding; (8) Roof Decks & Roofing; (9) Plaster Partitions & Ceiling; (10) Standard Doors, Windows & Hardware; (11) Shielding Doors; (12) Sprinkler System & Fire Extinguishers, Contents - Division XIII; (13) Heating, Ventilating & Air Conditioning; (14) Painting, Protective Coating & Floor Covering, Contents - Division XV; (15) Electrical; (16) Communications & Alarm Systems; (17) Special Equipment & Furnishings; (18) Overhead Bridge Crane; (19) Prefabricated Steel Building; (20) Paved Drive; and (21) Landscaping & Irrigation Sprinklers.

  15. Seismic strengthening of overhead roads between reactor buildings of WWER-1000 MW type NPP

    International Nuclear Information System (INIS)

    Stoyanov, G.; Jordanov, M.

    2005-01-01

    This paper presents results obtained during the upgrading design of overhead roads (OHR) between WWER-1000 MW Reactor Units at Kozloduy NPP. In order to avoid the deficiencies of OHR seismic capacity different approaches were developed based on the site and structure specifics. Overhead roads are precasted RC structures. They consist of pedestrian gallery and pipeline RC box, connecting reactor buildings with auxiliary building. They are mounted at approximately 10 m above ground level. The overhead roads are evaluated at their as-is status and a seismic upgrading of the structure is designed. The analysis of the upgraded structure is performed for Review Level Earthquake (RLE). Soil-Structure Interaction (SSI) effects are taken into account through equivalent soil springs with frequency adjusted stiffnesses. The upgraded structure is checked for conformance with the specially developed technical design specification based on International, US and Bulgarian standards and codes, taking into account site specific conditions. The general approach is consistent with up-to-date practice for evaluation and upgrade of nuclear power plant facilities. The existing site conditions and Owner's requirements are taken into account during development of the upgrading. The proposed upgrading measures can be divided in two major categories global and local. Special attention is paid to improvement of the ductile behavior of the structure through new detailing and upgrading of existing connection. These measures are grouped in two final design concepts and after a comparative study one of them is chosen for implementation. For the upgraded structure response spectra are derived at locations where equipment is attached. (authors)

  16. Model test on interaction of reactor building and soil. Part 1

    International Nuclear Information System (INIS)

    Iguchi, M.; Akino, K.; Kiva, Y.

    1989-01-01

    Theoretical and experimental studies on the effects of dynamic interaction between structures and soil have been carried out in recent years. Most of the dynamic tests, however, have been conducted using comparatively small-scale models. In order to evaluate the effects of soil-structure interaction for rigid structure such as reactor building, a series of tests, including forced vibration test and earthquake observations, was carried out. Large-scale models constructed on an actual soil were used. These tests included forced vibration tests on individual foundations, on foundations with superstructures, on cross interaction through the soil between adjacent structures. Tests on the embedded effects of foundation, on artificial ground-shaking, on large amplitude excitation, and aging effects in soil properties were performed. This paper describes the results of forced vibration tests and analyses of cross interaction through the soil between adjacent structures

  17. Reactor building integrity testing: A novel approach at Gentilly 2 - principles and methodology

    International Nuclear Information System (INIS)

    Collins, N.; Lafreniere, P.

    1991-01-01

    In 1987, Hydro-Quebec embarked on an ambitious development program to provide the Gentilly 2 nuclear power station with an effective, yet practical reactor building Integrity Test. The Gentilly 2 Integrity Test employs an innovative approach based on the reference volume concept. It is identified as the Temperature Compensation Method (TCM) System. This configuration has been demonstrated at both high and low test pressure and has achieved extraordinary precision in the leak rate measurement. The Gentilly 2 design allows the Integrity Test to be performed at a nominal 3 kPa(g) test pressure during an (11) hour period with the reactor at full power. The reactor building Pressure Test by comparison, is typically performed at high pressure 124 kPa(g)) in a 7 day window during an annual outage. The Integrity Test was developed with the goal of demonstrating containment availability. Specifically it was purported to detect a leak or hole in the 'bottled-up' reactor building greater in magnitude than an equivalent pipe of 25 mm diameter. However it is considered feasible that the high precision of the Gentilly 2 TCM System Integrity Test and a stable reactor building leak characteristic will constitute sufficient grounds for the reduction of the Pressure Test frequency. It is noted that only the TCM System has, to this date, allowed a relevant determination of the reactor building leak rate at a nominal test pressure of 3 kPa(g). Classical method tests at low pressure have lead to inconclusive results due to the high lack of precision

  18. Development of remote decontamination technologies improving internal environment of reactor buildings at Fukushima Daiichi Nuclear Power Station

    International Nuclear Information System (INIS)

    Hotta, Koji; Hayashi, Hirotada; Sakai, Hitoshi

    2016-01-01

    The reactor buildings at the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Co., Inc., which was seriously damaged by the Great East Japan Earthquake of March 11, 2011, have been highly contaminated by radioactive materials. To safely and efficiently advance the processes related to the forthcoming decommissioning of the reactors, it is necessary to improve the hazardous environment inside the reactor buildings. During the more than four years that have elapsed since the Great East Japan Earthquake, Toshiba has been implementing various measures to reduce the ambient dose rates inside the reactor buildings through decontamination work and participation in a national project for the development of remote decontamination technologies for reactor buildings. A variety of vehicles and technologies to support decontamination work have been developed through these activities, and are significantly contributing to improvement of the environment inside the reactor buildings. (author)

  19. Method of constructing reactor buildings

    International Nuclear Information System (INIS)

    Hyuga, Takenori; Nagai, Fumio; Akutsu, Masayoshi.

    1985-01-01

    Purpose: To shorten the construction period for LMFBR type reactors, as well as smoothly introduce high pressure steams generated in concretes upon loss of coolant accidents to the outside of the system. Method: After disposing a liner plate as a chamber lining of reactor buildings, heat insulation materials having steam discharge channels at the outer surface are attached to the outside of the liner plate and, further, an organic films are disposed to the outside of the heat insulation materials. Then, concretes are spiked to the outside of the organic films using the liner plate and the heat insulation material as the mold for concretes. In this way, the construction period can be shortened by utilizing the liner plate and the heat insulation materials as the mold for concretes, as well as steams at high temperature resulted in the concretes upon loss of coolant accidents can smoothly be discharged to the outside of the system. (Moriyama, K.)

  20. Study on the leak rate test for HANARO reactor building

    International Nuclear Information System (INIS)

    Choi, Y. S.; Kim, Y. K.; Kim, M. J.; Park, J. M.; Woo, J. S.

    2002-01-01

    The reactor building of HANARO adopts the confinement concept, which allows a certain amount of air leakage. In order to restrict the air leakage through the confinement boundary, negative pressure of at least 2.5 mmWG is maintained in normal operating condition while maintaining 25 mmWG of negative pressure in abnormal condition, the inside air filtered by a train of charcoal filter is released to the atmosphere through the stack. In this situation, if the emergency ventilation system is not operable, the reactor building is isolated from the outside then the trapped air inside will be leaked out through the building by ground release concept. As the leak rate may be affected by an effect of wind velocity outside the reactor building, the air tightness of confinement should be maintained to limit the leak rate below the allowable value. The local leak rate test method was used since the beginning of the commissioning until July 1999. However it has been pointed out as a defect that the method is so susceptible to the change of temperature and atmospheric pressure during testing. For more accurate leak rate testing, we have introduced a new test method. We have periodically carried out the new leak rate testing and the results indicate that the bad effect by the temperature and atmospheric pressure change is considerably reduced, which gives more stable leak rate measurement

  1. Parliament votes against building fifth power reactor

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    After a heated three-day debate, Finland's parliament voted on September 24 to reject the proposal to build the country's fifth nuclear power reactor. As predicted, the vote was close: 107 voted against more nuclear power, 90 were in favor, two members of the 200-seat parliament were not present, and the speaker did not vote

  2. Prediction of prestressing losses for long term operation of nuclear reactor buildings

    Directory of Open Access Journals (Sweden)

    Thillard G.

    2011-04-01

    Full Text Available Prestressed concrete is used in nuclear reactor buildings to guarantee containment and structural integrity in case of an accident. Monitoring and operating experience over 40 years has shown that prestressing losses can be much greater than the design estimation based on the usual standard laws. A method was developed to determine the realistic residual prestress level in structures, in particular for those where no embedded instrumentation was installed, taking into account in situ measurement results rather than design characteristics. The results can enable the owner to justify extending the lifespan while guaranteeing adequate safety and to define and plan adequate maintenance actions.

  3. Calculation of prefabricated part of WWR-K reactor building

    International Nuclear Information System (INIS)

    Belyashova, N.N.; Aptikaev, F.F.; Kopnichev, Yu.F.

    1998-01-01

    According of factual characteristics a strength and deformation of over-land part of carrier constructions under construction movement is defined. Direct dynamical calculation of design elements under action of inertial loads from supports shifts shows, that seismic stability of enclosing construction is not ensured. Possibly practically total collapse of coating construction is possibly, under which following levels of damages of internal design constructions of reactor central room have been forecasted: 1. Fall of destroyed design construction on reactor vessel in time moment (1.56-1.59 s) after coming to building of earthquake seismic waves of 10 balls. 2. It is possibly cracks formation in radial direction in lower part of reactor cap, but destroying of cap does not incident; 3. It is possibly cracks formation within stretched concrete zone of reactor construction at the mark from - 0.859 up to 0.100. Destroy of concrete's compressive zone of reactor construction have not being expected. 4. Collapse of reactor first contour coating constructions have not being expected

  4. Vibration-proof FBR type reactor

    International Nuclear Information System (INIS)

    Kawamura, Yutaka.

    1992-01-01

    In a reactor container in an FBR type reactor, an outer building and upper and lower portions of a reactor container are connected by a load transmission device made of a laminated material of rubber and steel plates. Each of the reactor container and the outer building is disposed on a lower raft disposed on a rock by way of a vibration-proof device made of a laminated material of rubber and steel plates. Vibration-proof elements for providing vertical eigen frequency of the vibration-proof system comprising the reactor building and the vibration-proof device within a range of 3Hz to 5Hz are used. That is, the peak of designed acceleration for response spectrum in the horizontal direction of the reactor structural portions is shifted to side of shorter period from the main frequency region of the reactor structure. Alternatively, rigidity of the vibration-proof elements is decreased to shift the peak to the side of long period from the main frequency region. Designed seismic force can be greatly reduced both horizontally and vertically, to reduce the wall thickness of the structural members, improve the plant economy and to ensure the safety against earthquakes. (N.H.)

  5. Culham conceptual Tokamak reactor MkII. Conceptual layout of buildings for a twin reactor power station

    International Nuclear Information System (INIS)

    Guthrie, J.A.S.; Harding, N.H.

    1981-01-01

    This paper discusses the conceptual design of the nuclear complex of a 2400 MWe twin fusion reactor power station utilising common services and a single containment building. The design is based upon environmental and maintenance logistical requirements, the provision of adequate storage, workshop and construction facilities and the constraints imposed by the geometry of the main and auxiliary reactor coolant systems. (author)

  6. Damage Curves of a Nuclear Reactor Structure exposed to Air Blast Loading

    International Nuclear Information System (INIS)

    Brandys, I.; Ornai, D.; Ronen, Y.

    2014-01-01

    Nuclear Power Plant (NPP) radiological hazards due to accidental failure or deliberated attacks are of most concern due to their destructive and global consequences: large area contaminations, injuries, exposure to ionizing radiation (which can cause death or illness, depends on the levels of exposure), loss of lives of both humans and animals, and severe damage to the environment. Prevention of such consequences is of a global importance and it has led to the definition of safety & design guidelines, and regulations by various authorities such as IAEA, U.S. NRC, etc. The guidelines define general requirements for the integrity of a NPP’s physical barriers (such as protective walls) when challenged by external events, for example human induced explosion. A more specific relation to the design of a NPP is that its structures and equipment (reactor building, fuel building, safeguards building, diesel-generator building, pumping station, nuclear auxiliaries building, and effluent treatment building) must function properly: shutdown the reactor, removal of decayed heat, storage of spent fuel, and treatment and containment of radioactive effluents) under external explosion. It requires that the NPP’s structures and equipment resistance to external explosion should be analyzed and verified. The air blast loading created by external explosion, as well as its effects & consequences on different kinds of structures are described in the literature. Structural elements response to the air blast can be analyzed in general by a Single Degree of Freedom (SDOF) system that converts a distributed mass, loads, and resistance to concentrated mass, force, and stiffness located at a representative point of the structure's element where the displacements are the highest one. Proper shielding should be designed if the explosion blast effects are greater than the resistance capacity.External explosion effects should be considered within the Screening Distance Value (SDV) of the NPP

  7. Seismic calculations for underground reactor buildings

    International Nuclear Information System (INIS)

    Altes, J.; Koschmieder, D.

    1977-08-01

    Embedding the buildings in soil changes their seismic response behaviour as compared to surface buildings, i.e. higher stiffness and increased radiation damping is attained. Finite element models are best suited for determinig the effects of embedment and of layered subsoil. The code used was the LUSH2-programme, which is applicable to 2-dimensional problems and provides an approximate treatment for non-linear dynamic soil behaviour. For embedded buildings there is a good agreement between 2- and 3-dimensional models of the response for points below the soil surface. It is therefore permissible to use the less costly 2-dimensional programmes. To simulate earthquake, three different acceleration-time histories, derived from actual measurements and from artificial synthesis, with differing response spectra were fed in. The soil characteristics assumed are applicable to a representative site in Germany. Three different types of models were examined, using analytical models with only a few elements for parametric studies and with up to 716 elements for more precise calculations. A comparison was made between the semi-embedment, the total embedment, and installation of the reactor building above-ground. (orig.) [de

  8. Dynamic analysis of the reactor building for soft (Kozloduy) and hard (Temelin) soil conditions and different seismic loading

    International Nuclear Information System (INIS)

    Krutzik, N.

    1995-01-01

    Analyses were conducted for the reactor building to determine the dynamic responses of the coupled system, soil and structure and the forces in the characteristic structural members. This report summarizes the results of structural dynamic analyses derived for soft and hard soil conditions by the modal time history method using substructure models as well as (for soft soil conditions) in the frequency domain using complex (coupled) models of the soil and the structure. The mathematical model of the reactor building is represented as a lumped mass beam model. The capabilities of the soil were represented by means of global frequency independent springs and dampers (substructure models) or by an appropriate final element model. The results of the above-mentioned analysis presented in this report comprise in particular the maximum values of accelerations, displacements and internal forces as well as the acceleration response spectra for the relevant building regions. The time domain (modal time history) calculations were performed for real soil conditions which corresponds to the site Kozloduy (soft) and Temelin (hard). As seismic input data the corresponding free-field data here been used. The dynamic response obtained for the soft-soil conditions using both type of (substructure and complex) models were compared and demonstrated in one plot. Similar comparison were performed for the results obtained for soft and hard soil conditions

  9. Nuclear reactors. Introduction

    International Nuclear Information System (INIS)

    Boiron, P.

    1997-01-01

    This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

  10. Investigation of hydrogen-burn damage in the Three Mile Island Unit 2 reactor building

    International Nuclear Information System (INIS)

    Alvares, N.J.; Beason, D.G.; Eidem, G.R.

    1982-06-01

    About 10 hours after the March 28, 1979 Loss-of-Coolant Accident began at Three Mile Island Unit 2, a hydrogen deflagration of undetermined extent occurred inside the reactor building. Examinations of photographic evidence, available from the first fifteen entries into the reactor building, yielded preliminary data on the possible extent and range of hydrogen burn damage. These data, although sparse, contributed to development of a possible damage path and to an estimate of the extent of damage to susceptible reactor building items. Further information gathered from analysis of additional photographs and samples can provide the means for estimating hydrogen source and production rate data crucial to developing a complete understanding of the TMI-2 hydrogen deflagration. 34 figures

  11. Evaluation of the Three Mile Island Unit 2 reactor building decontamination process

    Energy Technology Data Exchange (ETDEWEB)

    Dougherty, D.; Adams, J. W.

    1983-08-01

    Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams. Solid wastes being disposed of in commercial shallow land burial include trash and rubbish, ion-exchange resins (Epicor-II) and strippable coatings. The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of 10 CFR Part 61. It appears that much of the Epicor-II ion-exchange resin being disposed of in commerical land burial will be Class B and require stabilization if current radionuclide loading practices continue to be followed. Some of the trash and rubbish from the cleanup of the reactor building so far would be Class B. Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, stability under immersion and biodegradability. Actual coating samples from reactor building decontamination testing were evaluated for radionuclide leaching and biodegradation.

  12. Evaluation of the Three Mile Island Unit 2 reactor building decontamination process

    International Nuclear Information System (INIS)

    Dougherty, D.; Adams, J.W.

    1983-08-01

    Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams. Solid wastes being disposed of in commercial shallow land burial include trash and rubbish, ion-exchange resins (Epicor-II) and strippable coatings. The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of 10 CFR Part 61. It appears that much of the Epicor-II ion-exchange resin being disposed of in commerical land burial will be Class B and require stabilization if current radionuclide loading practices continue to be followed. Some of the trash and rubbish from the cleanup of the reactor building so far would be Class B. Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, stability under immersion and biodegradability. Actual coating samples from reactor building decontamination testing were evaluated for radionuclide leaching and biodegradation

  13. Modern frame structure buildings

    Directory of Open Access Journals (Sweden)

    В. М. Першаков

    2013-07-01

    Full Text Available The article deals with the design, construction and implementation of reinforced concrete frame structures with span 18, 21 m for agricultural production buildings, hall-premises of public buildings and buildings of agricultural aviation. Structures are prefabricated frame buildings and have such advantages as large space inside the structure and lower cost compared with other facilities with same purpose

  14. Method for temporary shielding of reactor vessel internals

    International Nuclear Information System (INIS)

    Grimm, N.P.; Sejvar, J.

    1991-01-01

    This patent describes a method for shielding stored internals for reactor vessel annealing. It comprises removing nuclear fuel from the reactor vessel containment building; removing and storing upper and lower core internals under water in a refueling canal storage area; assembling a support structure in the refueling canal between the reactor vessel and the stored internals; introducing vertical shielding tanks individually through a hatch in the containment building and positioning each into the support structure; introducing horizontal shielding tanks individually through a hatch in the containment building and positioning each above the stored internals and vertical tanks; draining water from the refueling canal to the level of a flange of the reactor vessel; placing an annealing apparatus in the reactor vessel; pumping the remaining water from the reactor vessel; and annealing the reactor vessel

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

    International Nuclear Information System (INIS)

    2015-01-01

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

  16. On detonation dynamics in hydrogen-air-steam mixtures: Theory and application to Olkiluoto reactor building

    International Nuclear Information System (INIS)

    Silde, A.; Lindholm, I.

    2000-02-01

    This report consists of the literature study of detonation dynamics in hydrogen-air-steam mixtures, and the assessment of shock pressure loads in Olkiluoto 1 and 2 reactor building under detonation conditions using the computer program DETO developed during this work at VTT. The program uses a simple 1-D approach based on the strong explosion theory, and accounts for the effects of both the primary or incident shock and the first (oblique or normal) reflected shock from a wall structure. The code results are also assessed against a Balloon experiment performed at Germany, and the classical Chapman-Jouguet detonation theory. The whole work was carried out as a part of Nordic SOS-2.3 project, dealing with severe accident analysis. The initial conditions and gas distribution of the detonation calculations are based on previous severe accident analyses by MELCOR and FLUENT codes. According to DETO calculations, the maximum peak pressure in a structure of Olkiluoto reactor building room B60-80 after normal shock reflection was about 38.7 MPa if a total of 3.15 kg hydrogen was assumed to burned in a distance of 2.0 m from the wall structure. The corresponding pressure impulse was about 9.4 kPa-s. The results were sensitive to the distance used. Comparison of the results to classical C-J theory and the Balloon experiments suggested that DETO code represented a conservative estimation for the first pressure spike under the shock reflection from a wall in Olkiluoto reactor building. Complicated 3-D phenomena of shock wave reflections and focusing, nor the propagation of combustion front behind the shock wave under detonation conditions are not modeled in the DETO code. More detailed 3-D analyses with a specific detonation code are, therefore, recommended. In spite of the code simplifications, DETO was found to be a beneficial tool for simple first-order assessments of the structure pressure loads under the first reflection of detonation shock waves. The work on assessment

  17. Experimental study on joint construction method for aseismatic walls of reactor buildings, (1)

    International Nuclear Information System (INIS)

    Sugita, Kazunao; Mogami, Tatsuo; Ezaki, Tetsuro

    1987-01-01

    On the aseismatic walls of a reactor auxiliary building, many temporary openings are provided at the time of the construction for carrying equipment in later, due to the demand of shortening the construction period. Thus on the aseismatic walls, in most cases there are the joints due to the concrete placed later. As equipment tends to be unitized and become large, the quipment is placed close to the wall having an opening, consequently, the workability is poor, and the standardization of construction method is urgently demanded. The conventional method of closing temporary openings has the problems of safety and connecting reinforcing bars, therefore, the new construction method was proposed. In reactor buildings, the joints of walls are unavoidable, and since those are large scale structures, the joints are numerous. Therefore, at the joint parts, it abandoned and buried frames are used, it is advantageous in the time and cost of joint construction. In both cases, the mechanical properties were confirmed by the fundamental performance test partially modeling the joints and the verifying test modeling the whole walls. In this paper, the test of applying only shearing force to joint models is reported. (Kako, I.)

  18. Ultimate shearing strength of aseismatic walls with many small holes for reactor buildings

    International Nuclear Information System (INIS)

    Yoshizaki, Seiji; Ezaki, Tetsuro; Korenaga, Takeyoshi; Sotomura, Kentaro.

    1984-01-01

    The aseismatic walls for reactor buildings have complicated forms, and are characterized by large wall thickness and high reinforcement ratio as compared with ordinary aseismatic walls. The forms are mainly box, cylinder or irregular polygonal prism and their combination. The design of the walls with many small holes has been performed on the basis of the reinforced concrete structure calculation standard of the Architectural Institute of Japan, following the case with large opening. When there are many small holes, the arrangement of reinforcement for the openings becomes complex, and the construction is difficult. It is necessary to rationalize the design and to simplify the reinforcement work. Under the background like this, the experiment to examine the shearing property in bending of the aseismatic walls with many small holes for reactor buildings was carried out, and horizontal loading test was performed on 43 specimens. The method of calculating the ultimate shearing strength of a wall without opening was proposed, and the method of applying it to a wall with many small holes is shown. The experimental method and the results, the examination of the experimental results, and the ultimate shearing strength of the aseismatic walls are reported. (Kako, I.)

  19. Experimental and theoretical investigations of soil-structure interaction effect at HDR-reactor-building

    International Nuclear Information System (INIS)

    Wassermann, K.

    1983-01-01

    Full-scale dynamic testing on intermediate and high levels was performed at the Heissdampfreaktor (HDR) in 1979. Various types of dynamic forces were applied and response of the reactor containment structure and internal components was measured. Precalculations of dynamic behaviour and response of the structure were made through different mathematical models for the structure and the underlying soil. Soil-Structure Interaction effects are investigated and different theoretical models are compared with experimental results. In each model, the soil is represented by springs attached to the structural model. Stiffnesses of springs are calculated by different finite-element idealizations and half-space approximations. Eigenfrequencies and damping values related to interaction effects (translation, rocking, torsion) are identified from test results. The comparisons of dynamic characteristic of the soil-structure system between precalculations and test results show good agreement in general. (orig.)

  20. Remote tritium-in-air sampling in reactor building at NAPS

    International Nuclear Information System (INIS)

    Mitra, S.R.; Lal Chand

    2000-01-01

    Tritium-in-air activity is an important parameter in PHW reactors from the point of view of internal exposure and heavy water escape from the system. The sampling technique in vogue in PHWRs, for measurement of tritium-in-air activity, requires collection of on the spot sample from different areas using a portable sampler. This sampler uses the bubbler method of sampling. As the areas of sampling are numerous, this technique is time consuming, laborious and can lead to significant internal exposure in areas where tritium-in-air activity is high. This technique is also error prone due to the heavy workload involved. A new scheme, in which the sampling of all the areas of reactor building is done through a sampling station, has been introduced for the first time in NAPS. This sampling station facilitates collection of samples from all the areas of reactor building, remotely and simultaneously at one place thereby reducing time, labour, exposure and error. This paper gives the details of the sampling system installed at NAPS. (author)

  1. Task 24: Dynamic analysis of Kozloduy NPP unit 5 structures: Reactor building

    International Nuclear Information System (INIS)

    Zola, M.

    1999-01-01

    This report refers to the activities of a sub-contract to the Project RER/9/046, awarded to ISMES by the International Atomic Energy Agency (IAEA) of Vienna, to compare the results obtained from the experimental activities performed under previous contract by ISMES with those coming from analytical studies performed in the framework of the Coordinated Research Programme (CRP) on 'Benchmark Study for the Seismic Analysis and Testing of WWER-type Nuclear Power Plants' by other Institutions, relevant to Kozloduy Unit 5 reactor building. After a brief introduction to the problem in Chapter 1, the identification of the comparison positions and reference directions is given in Chapter 3. A very quick description of the performed experimental tests is given in Chapter 4, whereas the characteristics of both experimental and analytical data are presented in Chapter 5. The data processing procedures are reported in Chapter 6 and some simple remarks are given in Chapter 7. (author)

  2. Safety classification of systems, structures, and components for pool-type research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Ryong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2016-08-15

    Structures, systems, and components (SSCs) important to safety of nuclear facilities shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions. Although SSC classification guidelines for nuclear power plants have been well established and applied, those for research reactors have been only recently established by the International Atomic Energy Agency (IAEA). Korea has operated a pool-type research reactor (the High Flux Advanced Neutron Application Reactor) and has recently exported another pool-type reactor (Jordan Research and Training Reactor), which is being built in Jordan. Korea also has a plan to build one more pool-type reactor, the Kijang Research Reactor, in Kijang, Busan. The safety classification of SSCs for pool-type research reactors is proposed in this paper based on the IAEA methodology. The proposal recommends that the SSCs of pool-type research reactors be categorized and classified on basis of their safety functions and safety significance. Because the SSCs in pool-type research reactors are not the pressure-retaining components, codes and standards for design of the SSCs following the safety classification can be selected in a graded approach.

  3. Evaluation of tritiated water retention capacity of fusion reactor concrete building

    International Nuclear Information System (INIS)

    Numata, S.; Fujii, Y.; Okamoto, M.

    1992-01-01

    In this paper the diffusion of tritiated water vapor into concrete walls is studied to evaluate tritiated water retention capacity of a fusion reactor concrete building. Using a model of the tritiated water diffusion determined form experimental results, depth profiles of tritiated water in concrete are calculated in the case of being exposed to air containing tritiated water vapor during the normal operational condition of a fusion reactor. A 0.5-m-thick concrete is sufficient for reactor hall walls from a viewpoint of the tritium containment

  4. Experimental and analytical studies of a deeply embedded reactor building model considering soil-building interaction. Pt. 1

    International Nuclear Information System (INIS)

    Tanaka, H.; Ohta, T.; Uchiyama, S.

    1979-01-01

    The purpose of this paper is to describe the dynamic characteristics of a deeply embedded reactor building model derived from experimental and analytical studies which considers soil-building interaction behaviour. The model building is made of reinforced concrete. It has two stories above ground level and a basement, resting on sandy gravel layer at a depth of 3 meters. The backfill around the building was made to ground level. The model building is simplified and reduced to about one-fifteenth (1/15) of the prototype. It has bearing wall system for the basement and the first story, and frame system for the second. (orig.)

  5. Air leakage analysis of research reactor HANARO building in typhoon condition for the nuclear emergency preparedness

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Goany Up; Lee, Hae Cho; Kim, Bong Seok; Kim, Jong Soo; Choi, Pyung Kyu [Dept. of Emergency Preparedness, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    To find out the leak characteristic of research reactor 'HANARO' building in a typhoon condition MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. It was found that the leak rate is 0.1%·day{sup -1} of air, 0.004%·day{sup -1} of noble gas and 3.7×10{sup -5}%·day{sup -1} of aerosol during typhoon passing. The air leak rate of 0.1%·day can be converted into 1.36 m{sup 3}·hr{sup -1} , but the design leak rate in HANARO safety analysis report was considered as 600 m3·hr{sup -1} under the condition of 20 m·sec{sup -1} wind speed outside of the building by typhoon. Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor.

  6. Air leakage analysis of research reactor HANARO building in typhoon condition for the nuclear emergency preparedness

    International Nuclear Information System (INIS)

    Lee, Goany Up; Lee, Hae Cho; Kim, Bong Seok; Kim, Jong Soo; Choi, Pyung Kyu

    2016-01-01

    To find out the leak characteristic of research reactor 'HANARO' building in a typhoon condition MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. It was found that the leak rate is 0.1%·day -1 of air, 0.004%·day -1 of noble gas and 3.7×10 -5 %·day -1 of aerosol during typhoon passing. The air leak rate of 0.1%·day can be converted into 1.36 m 3 ·hr -1 , but the design leak rate in HANARO safety analysis report was considered as 600 m3·hr -1 under the condition of 20 m·sec -1 wind speed outside of the building by typhoon. Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor

  7. Reactor building design of nuclear power plant ATUCHA II, Argentina

    International Nuclear Information System (INIS)

    Rufino, R.E.; Hermann, E.R.; Richter, E.

    1984-01-01

    It is presented the civil engineering project carried out by the joint venture Hochtief - Techint-Bignoli (HTB) for the reactor building at the Atucha II power plant (PHWR of 745 MWe) in Buenos Aires. All the other civil projects at Atucha II are also being carried out by HTB. This building has the same general characteristics of the PWR plants developed by KWU in Germany, known for the spherical steel containment 56m in diameter. Nevertheless, it differs from those principally in the equipment lay-out and the remarkable foundation depth. From the basic engineering provided by ENACE, the joint venture has had to face the challenge of designing a tridimensional structure of large size. This has necessitated using simplified models which had to be superimposed, since the use of only one spatial mode would be highly inadequate, lacking the flexibility necessary to absorb the numerous modifications that this type of project undergoes during construction. In addition, this procedure has eliminated resorting to numerous and costly computer processings. (Author) [pt

  8. The 3D-FEM modeling of the LAES unit 1 reactor building for extreme external effects

    International Nuclear Information System (INIS)

    1999-01-01

    In order to study the extreme external effects, three dimensional model was applied to study the effects of aircraft crash and gas explosion on the reactor building of Leningrad-1 NPP which is modelled by finite element method. The crash loads taken into account were from Cessna civil airplane crash with impact velocity of 360 km/h and maximum impact force of 7 MN and the Phantom military airplane crash with impact velocity of 215 km/h and maximum impact force of 110 MN. The gas explosion load was assumed to affect the reactor building from one side parallel to one of the global coordinate axes of the model. The conclusion drawn from the obtained results is as follows: the intersections stiffen the structure considerably. In lower part, where many intersections exist, displacements were significantly smaller. Thus, the lower parts can resist the investigated loads such as high speed military aircraft crash loads much better than the upper part

  9. Seismic response of a nonsymmetric nuclear reactor building with a flexible stepped foundation

    International Nuclear Information System (INIS)

    Okano, H.; Sakai, A.; Takita, H.; Fukunishi, S.; Nakatogawa, T.; Kabayama, K.

    1993-01-01

    The effect of the non symmetry of a nuclear reactor building on its seismic response was studied. The nonsymmetric natures we considered, Included the eccentricity of the superstructure and the non symmetry of the cross section of the foundation. A three-dimensional analysis which employed Green's function was applied to study the interaction between the soil and the non symmetrically sectioned foundation. The effect of a flexible foundation on its seismic response was also studied by applying the sub structuring method, which combines the finite element method and Green's function method. (author)

  10. Decontamination and concrete core sampling by teleoperated robot at Fukushima Daiichi reactor buildings

    International Nuclear Information System (INIS)

    Watanabe, Masaru; Onitsuka, Hironori; Shimonabe, Noriaki; Fujita, Jun; Matsumura, Takumi; Okumura, Atsushi

    2015-01-01

    For decommissioning of Fukushima daiichi nuclear power station, reduction of the dose equivalent rates inside the reactor buildings is an important issue. Concrete core sampling from the buildings to investigate the contamination is necessary for study about effective decontamination. However, dose rate inside the reactor buildings is very high. For example, dose rate of 1st floor on the Unit 1 is 1.2 - 1820 [mSv / h], the Unit 2 is 2.5 - 220 [mSv / h] and Unit 3 is 2.2 - 4780 [mSv / h]. So it is difficult for workers to work long hours. Therefore, a teleoperated robot, named 'MHI-MEISTeR (Mitsubishi Heavy Industries - Maintenance Equipment Integrated System of Telecontrol Robot)', has been developed to conduct operations like concrete core samples from the reactor buildings. Actually, some concrete core samples from Fukushima daiichi were taken by MHI-MEISTeR. In addition, MHI-MEISTeR is designed as a versatile robot, and so it can conduct suction / blast decontamination works as well as concrete core sampling. The above operations were performed by MHI-MEISTeR in Fukushima daiichi nuclear power station. (author)

  11. Build-up of actinides in irradiated fuel rods of the ET-RR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Adib, M.; Naguib, K.; Morcos, H.N

    2001-09-01

    The content concentrations of actinides are calculated as a function of operating reactor regime and cooling time at different percentage of fuel burn-up. The build-up transmutation equations of actinides content in an irradiated fuel are solved numerically .A computer code BAC was written to operate on a PC computer to provide the required calculations. The fuel element of 10% {sup 235}U enrichment of ET-RR-1 reactor was taken as an example for calculations using the BAC code. The results are compared with other calculations for the ET-RR-1 fuel rod. An estimation of fissile build-up content of a proposed new fuel of 20% {sup 235}U enrichment for ET-RR-1 reactor is given. The sensitivity coefficients of build-up plutonium concentrations as a function of cross-section data uncertainties are also calculated.

  12. Structural analysis of the P reactor at the Savannah River Site

    International Nuclear Information System (INIS)

    Zaslawsky, M.; Maryak, M.

    1991-01-01

    A seismic analysis of the P-reactor buildings that were built in the early 1950's has been performed using current criteria and analysis techniques. The seismic input is based on the RG 1.60 free field response spectra anchored at 0.2g ZPA. The SSI analysis applied deconvolution techniques to establish soil parameters based on strain dependent damping and shear modulus relationships. The analysis used 2-dimensional soil structure interaction techniques to generate floor response spectra. The spectra were adjusted to account for torsional amplifications resulting from differences between the locations of the center of mass of the floors and the center of rigidity of the connecting vertical column elements. The resulting floor response spectra were smoothed and broadened in accordance with NRC criteria. In addition to developing floor response spectra, building shears and moments were obtained and an assessment of the structural capacity of the buildings to withstand the seismic loads was made

  13. The construction of a PWR power station reactor building liner

    International Nuclear Information System (INIS)

    Skirving, N.; Goulding, J.S.; Gibson, J.A.

    1991-01-01

    Cleveland Bridge and Engineering Co Ltd (CBE) are constructing the Reactor Building Liner Plate containment of the Sizewell 'B' Power Station for Nuclear Electric Ltd. This has entailed extensive offsite prefabrication of components and their subsequent erection at Sizewell. It has been necessary to engineer temporary supporting mechanisms to enable manufacture and erection to proceed, yet also to withstand wet concrete forces during the progressive construction. The Reactor Building Liner Plate is a safety related system and as such, in addition to strict compliance with the ASME code, the Quality Assurance (QA) requirements of BS 5882 are applicable. A dedicated Project Team was established by CBE to control and direct the work. Equally important as satisfying the rigorous Q.A. requirements has been the need to meet programme and budget. This paper details CBE execution of the Project. (author)

  14. Labour input in construction of composite structures of the Balakovo NPP reactor compartment

    International Nuclear Information System (INIS)

    Alasyuk, G.Ya.

    1988-01-01

    Technical-economical results achieved when constructing the Balakovo NPP second unit reactor compartment structures are presented. The obtained data analysis shows that in the case of building the walls of non-sealed reactor compartment section in the form of composite structures the major part of labour input requirements (54-59%) falls at works on production and mounting of these structures, performed at auxiliary plants. Labour input for works performed the construction (unit-cell and space frame mounting, preparation of units for concreting, joint sealing, concrete placement) make up 41-46%, and labour input for enlarged unit-cell mounting make up 8%. Labour input per 1 m 3 of the wall structure with 0.6 and 0.9 m thicness in the monolith option are respectively by 19 an 23% higher than the same indices for composite

  15. Reactor building indoor wireless network channel quality estimation using RSSI measurement of wireless sensor network

    International Nuclear Information System (INIS)

    Merat, S.

    2008-01-01

    Expanding wireless communication network reception inside reactor buildings (RB) and service wings (SW) has always been a technical challenge for operations service team. This is driven by the volume of metal equipment inside the Reactor Buildings (RB) that blocks and somehow shields the signal throughout the link. In this study, to improve wireless reception inside the Reactor Building (RB), an experimental model using indoor localization mesh based on IEEE 802.15 is developed to implement a wireless sensor network. This experimental model estimates the distance between different nodes by measuring the RSSI (Received Signal Strength Indicator). Then by using triangulation and RSSI measurement, the validity of the estimation techniques is verified to simulate the physical environmental obstacles, which block the signal transmission. (author)

  16. Reactor building indoor wireless network channel quality estimation using RSSI measurement of wireless sensor network

    Energy Technology Data Exchange (ETDEWEB)

    Merat, S. [Wardrop Engineering Inc., Toronto, Ontario (Canada)

    2008-07-01

    Expanding wireless communication network reception inside reactor buildings (RB) and service wings (SW) has always been a technical challenge for operations service team. This is driven by the volume of metal equipment inside the Reactor Buildings (RB) that blocks and somehow shields the signal throughout the link. In this study, to improve wireless reception inside the Reactor Building (RB), an experimental model using indoor localization mesh based on IEEE 802.15 is developed to implement a wireless sensor network. This experimental model estimates the distance between different nodes by measuring the RSSI (Received Signal Strength Indicator). Then by using triangulation and RSSI measurement, the validity of the estimation techniques is verified to simulate the physical environmental obstacles, which block the signal transmission. (author)

  17. Study on vertical seismic response characteristics of deeply embedded reactor building

    International Nuclear Information System (INIS)

    Morishita, H.; Nakamura, N.; Uchiyama, S.; Fukuoka, A.; Ishizaki, M.

    1993-01-01

    This paper describes vertical response characteristics, especially effects of embedment, and analytical methods for seismic design of a deeply embedded reactor building. The influence of embedment on vertical response was found to be minimal by evaluating results of forced vibration tests of a reactor building model and performing simplified analyses. Subsequently, simulation analyses of the forced vibration test and actual earthquake induced response were performed using both the axisymmetric FEM model and the simplified mass and spring model. It was concluded that the analytical models taking the embedment into the consideration closely simulated the observation records, and the omission of embedment in the analyses tended to increase the predicted response which was conservative in respect an actual design consideration. (author)

  18. Seismic evaluation and upgrading design of overhead roads between reactor buildings of WWER-1000 MW type NPP

    International Nuclear Information System (INIS)

    Jordanov, M.J.; Stoyanov, G.S.; Geshanov, I.H.; Kirilov, K.P.; Schuetz, W.

    2003-01-01

    This paper presents results obtained during the study of overhead roads between Reactor Building (RB) of WWER-1000 MW NPP and possible measures for their seismic upgrade. The main objective of this project is to evaluate the behavior of overhead roads under site-specific seismic loading and to determine whether this structure satisfies current international safety regulations, followed by development of upgrading concepts. Overhead roads are pre-cast RC structure, which can be divided to separate substructures. They comprise of pedestrian gallery and pipeline box, connecting reactor buildings with auxiliary building. They are mounted at approximately 10 m above ground level. The overhead roads are evaluated for Review Level Earthquake (RLE) as seismic category II structures. As seismic input motion is RLE, free field response spectra anchored to 0.2 g PGA are used with 0.5 scaling factor. Soil-Structure Interaction effects are taken into account through equivalent soil springs with frequency adjusted stiffness. In order to meet the objective of the project a technical design specification is developed for conformance with International, US and Bulgarian standards and codes, taking into account site specific conditions. The general approach is consistent with up-to-date practice for evaluation and upgrade of nuclear power plant facilities. The separate steps comprising the overall fulfillment of project's major objectives may be summarized as follows: study of all available data for initial design and as built conditions, creation of 3-D detailed finite element models for as-built structure, determination of dynamic characteristics, evaluation of adequacy of initial design under new seismic loading (calculation of D/C ratios for structural members and connections, evaluation of embedment lengths for embedded parts and rebars, deformation evaluation, stability checks), development of upgrading concepts for enhancement, verification of capability of upgraded structure

  19. Decontamination and decommissioning of the SPERT-I Reactor Building at the Idaho National Engineering Laboratory. Final report

    International Nuclear Information System (INIS)

    Dolenc, M.R.

    1986-02-01

    This final report documents the decontamination and decommissioning of the SPERT-I Reactor Building. This 20- by 40-ft galvanized steel building was dismantled; and the resultant contaminated sludge, liquid, and carbon steel were disposed of at the Radioactive Waste Management Complex of the Idaho National Engineering Laboratory. This report presents the results of the characterization, decision analysis, planning, and decommissioning of the facility. The total cost of these activities was $139,500. Of this total, $103,500 was required for decommissioning operations. (This latter figure represents a 20% savings over the estimated costs generated during the planning effort.) The objectives of decommissioning this facility were to stabilize the seepage pit area and remove the reactor building. The D and D work was divided into two parts; the seepage pit was decommissioned in 1984, and the reactor building in 1985. The entire area was backfilled with radiologically clean soil, graded, and seeded. Two markers were installed to identify the locations of the pit and reactor building. The only isotopes found in either decommissioning operation were cesium-137 and uranium-235 in very low concentrations. Decommissioning operations of the reactor building were carried out during August 1985. The project generate 297 ft 3 of radioactive waste. No personnel radiation exposure above background was received by D and D workers

  20. Vibration tests and analyses of the reactor building model on a small scale

    International Nuclear Information System (INIS)

    Tsuchiya, Hideo; Tanaka, Mitsuru; Ogihara, Yukio; Moriyama, Ken-ichi; Nakayama, Masaaki

    1985-01-01

    The purpose of this paper is to describe the vibration tests and the simulation analyses of the reactor building model on a small scale. The model vibration tests were performed to investigate the vibrational characteristics of the combined super-structure and to verify the computor code based on Dr. H. Tajimi's Thin Layered Element Theory, using the uniaxial shaking table (60 cm x 60 cm). The specimens consist of ground model, three structural model (prestressed concrete containment vessel, inner concrete structure, and enclosure building), a combined structural model and a combined structure-soil interaction model. These models are made of silicon-rubber, and they have a scale of 1:600. Harmonic step by step excitation of 40 gals was performed to investigate the vibrational characteristics for each structural model. The responses of the specimen to harmonic excitation were measured by optical displacement meters, and analyzed by a real time spectrum analyzer. The resonance and phase lag curves of the specimens to the shaking table were obtained respectively. As for the tests of a combined structure-soil interaction model, three predominant frequencies were observed in the resonance curves. These values were in good agreement with the analytical transfer function curves on the computer code. From the vibration tests and the simulation analyses, the silicon-rubber model test is useful for the fundamental study of structural problems. The computer code based on the Thin Element Theory can simulate well the test results. (Kobozono, M.)

  1. FEM analysis of foundation raft for 500 MWe pressurized heavy water reactor building

    International Nuclear Information System (INIS)

    Kulkarni, N.N.; Goray, J.S.; Joshi, M.H.; Paramasivam, V.

    1989-01-01

    Foundation raft supports the containment structure and internals for 500 MWe PHW reactor building. It also serves as bottom envelop of the containment structure. In view of this, the design of foundation raft assumes great importance. The foundation raft is subjected to various load, most significant of them are dead load of structure, equipment loads transferred through a system of floors, walls and structural steel columns, pressure load during accident conditions, seismic loads, earth pressure, uplift due to buoyancy loads, foundation reaction etc. In order to achieve optimum design, the detailed structural analysis is required to be performed methodically and in most realistic manner. Finite element methods which have come in vogue with the developments in digital computers can be successfully applied in this area. The paper describes the above methods in detail for the analysis of foundation raft for the various load combinations required to be considered for safe and optimum design

  2. The Structure of Affine Buildings

    CERN Document Server

    Weiss, Richard M

    2009-01-01

    In The Structure of Affine Buildings, Richard Weiss gives a detailed presentation of the complete proof of the classification of Bruhat-Tits buildings first completed by Jacques Tits in 1986. The book includes numerous results about automorphisms, completions, and residues of these buildings. It also includes tables correlating the results in the locally finite case with the results of Tits's classification of absolutely simple algebraic groups defined over a local field. A companion to Weiss's The Structure of Spherical Buildings, The Structure of Affine Buildings is organized around the clas

  3. Nonlinear seismic response analysis of embedded reactor buildings based on the substructure approach in time domain

    International Nuclear Information System (INIS)

    Hasegawa, M.; Nakai, S.; Watanabe, T.

    1985-01-01

    A practical method for elasto-plastic seismic response analysis is described under considerations of nonlinear material law of a structure and dynamic soil-structure interaction. The method is essentially based on the substructure approach of time domain analysis. Verification of the present method is carried out for typical BWR-MARK II type reactor building which is embedded in a soil, and the results are compared with those of the frequency response analysis which gives good accuracy for linear system. As a result, the present method exhibits sufficient accuracy. Furthermore, elasto-plastic analyses considering the soil-structure interaction are made as an application of the present method, and nonlinear behaviors of the structure and embedment effects are discussed. (orig.)

  4. Experimental and analytical studies of a deeply embedded reactor building model considering soil-building interaction. Pt. 3

    International Nuclear Information System (INIS)

    Tanaka, H.

    1983-01-01

    The paper describes the dynamic charachteristics of a deeply embedded reactor building model obtained from the forced vibration tests, earthquake observations and simulation analysis. The earthquake records of the structure and the surrounding soil were examined by using soil-building interaction model as used in the analyses of the forced vibration tests. It is considered that the response of the structure will be influenced by the seismic behaviour of the soil layer as the seismic wave is input to the bedrock of the soil-structure interaction model in the earthquake response analysis. Therefore, dynamic properties of the soil layer during earthquakes were investigated in detail, and applied to the seismic simulation analysis using soil-structure interaction model. Many earthquake records have been obtained since June, 1976 when the earthquake observation system was first established. From these, eight of them which had comparatively large acceleration values were used to investigate the transfer properties of soil layer. Besides, transfer functions computed using in-situ measurement shearing wave velocity showed good agreement with those of the earthquake records. The records of the Miyagiken-oki earthquake of February 20, 1978 (magnitude 6.7) was selected as an example for performing simulation analysis. The simulation analysis are as follows: (1) In the seismic simulation analysis using soil-structure interaction modal, computed results will be in good agreement with the observed ones, when the transfer function of soil layer is properly estimated. (2) Judging from the transfer function of soil layer with the characteristics that the modal damping value decreases gradually at a higher modal frequency, it is found that ddamping of soil-layer can be simulated more adequately by introducing external damping system together with structural damping. (orig./HP)

  5. A GASFLOW analysis of a steam explosion accident in a typical light-water reactor confinement building

    International Nuclear Information System (INIS)

    Travis, J.R.; Wilson, T.L.; Spore, J.W.; Lam, K.L.; Rao, D.V.

    1994-01-01

    Steam over-pressurization resulting from ex-vessel steam explosion (fuel-coolant interaction) may pose a serious challenge to the integrity of a typical light-water reactor confinement building. If the steam generation rate exceeds the removal capacity of the Airborne Activity Confinement System, confinement overpressurization occurs. Thus, there is a large potential for an uncontrolled and unfiltered release of fission products from the confinement atmosphere to the environment at the time of the steam explosion. The GASFLOW computer code was used to analyze the effects of a hypothetical steam explosion and the transport of steam and hydrogen throughout a typical light-water reactor confinement building. The effects of rapid pressurization and the resulting forces on the internal structures and the heat exchanger service bay hatch covers were calculated. Pressurization of the ventilation system and the potential damage to the ventilation fans and high-efficiency particulate air filters were assessed. Because of buoyancy forces and the calculated confinement velocity field, the hydrogen diffuses and mixes in the confinement atmosphere but tends to be transported to its upper region. (author). 2 refs., 14 figs

  6. A GASFLOW analysis of a steam explosion accident in a typical light-water reactor confinement building

    International Nuclear Information System (INIS)

    Travis, J.R.; Wilson, T.L.; Spore, J.W.; Lam, K.L.; Rao, D.V.

    1994-01-01

    Steam over-pressurization resulting from ex-vessel steam explosion (fuel-coolant interaction) may pose a serious challenge to the integrity of a typical light-water reactor confinement building. If the steam generation rate exceeds the removal capacity of the Airborne Activity Confinement System, confinement over pressurization occurs. Thus, there is a large potential for an uncontrolled and unfiltered release of fission products from the confinement atmosphere to the environment at the time of the steam explosion. The GASFLOW computer code was used to analyze the effects of a hypothetical steam explosion and the transport of steam and hydrogen throughout a typical light-water reactor confinement building. The effects of rapid pressurization and the resulting forces on the internal structures and the heat exchanger service bay hatch covers were calculated. Pressurization of the ventilation system and the potential damage to the ventilation fans and high-efficiency particulate air filters were assessed. Because of buoyancy forces and the calculated confinement velocity field, the hydrogen diffuses and mixes in the confinement atmosphere but tends to be transported to its upper region

  7. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  8. Environmental assessment for decontamination of the Three Mile Island Unit 2 reactor building atmosphere. Addendum 2. Draft NRC staff report for public comment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1980-04-01

    The reactor building purge system is an existing system originally installed for purging the reactor building atmosphere during normal operation or maintenance conditions. Use of the reactor building purge system in conjunction with the hydrogen control subsystem evaluated in Section 6.1 represents a variation in the purging alternative for decontaminating the Unit 2 reactor building atmosphere. This variation in the purging alternative would function only under meteorological conditions favorable to atmospheric dispersion. The reactor building purge system is capable of purging the building at flow rates of 5,000-50,000 cfm. Actual purge rates authorized during any time interval would be dependent on meteorological conditions and reactor building concentrations. Like the hydrogen control subsystem, this system would remove reactor building atmosphere through a filter system and discharge it through the 160-ft plant vent stack to the environment. The advantage of using the reactor building purge system in conjunction with the hydrogen control system is that it could decontaminate the reactor building atmosphere in a total elapsed purge time as short as approximately 5 days, as compared with the 60 days that would be required if the hydrogen purge subsystem were used alone. Use of this variation in the purge alternative would result in the release of radioactive materials to the environment. However, calculations based on actual meteorological and release-rate data would be used to monitor radioactive releases so that they do not exceed the requirements of 10 CFR Part 20, the design objectives of 10 CFR Part 50, Appendix 1 and the applicable requirements of 40 CFR 190.10.

  9. A study for structural safety of ISER reactor building under impact load

    International Nuclear Information System (INIS)

    Takeuchi, Yoichiro; Hasegawa, Toshiyasu; Mutoh, Atsushi; Wakabayashi, Hiroaki.

    1991-01-01

    ISER (Inherently Safe and Economical Reactor) proposed in Japan by an academic circle and industries is expected to be used world-wide particularly in developing countries where an energy crunch is feared in the 21-st century. A certain level of hardened structures for plant safety seems to be effective and may be required by the regulatory body, since the ISER is claimed to be inherently safe even against a kind of external load. This paper concerns impact resistant design of ISER. A brief state-of-the-art review on related works, impact resistant design flow and results of some preliminary analysis of a proposed ISER model is also presented. (author)

  10. Neutron activation of building materials used in the reactor shield

    International Nuclear Information System (INIS)

    Hernandez, A.T.; Perez, G.; D'Alessandro, K.

    1993-01-01

    Cuban concretes and their main components (mineral aggregates and cement) were investigated through long-lived activation products induced by neutrons from a reactor. The multielemental content in the materials studied was obtained by neutron activation analysis in an IBR-2 reactor and gamma activation analysis in an MT-25 microtron from Join Institute of Nuclear Research of Dubna. After irradiation of building materials for 30 years by a neutron flow of unitary density, induced radioactivity was calculated according to experimental data. The comparative evaluation of different concretes aggregates and two types of cement related to the activation properties is discussed

  11. On-line reactor building integrity testing at Gentilly-2 (summary of results 1987-1994)

    International Nuclear Information System (INIS)

    Collins, N.; Lafreniere, P.

    1994-01-01

    In 1987, Hydro-0uebec embarked on an ambitious development program to provide the Gentilly-2 Nuclear Power Station with an effective and practical Reactor Building Containment integrity Test (CIT). In October 1992, the inaugural low pressure (3 kPa(g) nominal) CIT at 100% F.P was performed. The test was conclusive and the CIT was declared In-Service for containment integrity verification on-line. Five subsequent CITs performed in 1993 and 1994 have demonstrated the expected leak rate results and good reliability. The outstanding feature of the CITs is the demonstrated accurary of better than 5% of the measured leak rate. The CIT was developed with the primary goal of demonstrating 'overall' containment availability. Specifically it was designed to detect a 25 mm. diameter leak or hole in the Reactor Building. However, the remarkable CIT accuracy allows reliable detection of a 2 mm. hole. The Gentilly-2 CIT is an innovative approach based on the Temperature Compensation Method (TCM) which uses a reference volume composed of an extensive tubular network of several different diameters. This eliminates the need to track numerous temperature points. A second independent tubular network includes numerous humidity sampling points, thereby enabling the mearurernent of minute pressure variations inside the Reactor Building, independant of the spatial and temporal humidity behaviour. This Gentilly-2 TOM System has been demonstrated to work at both high and low test pressures. The GentiIly-2 design allows the CIT to be performed at a nominal 3 kPa(g) test pressure during a 12-hour period (28 hours total with alignment time) with the reactor at full power. The traditional Reactor Building Pressure Test (RBPT) is typically performed at high pressure (124 kPa(g) in a 5-day critical path window (7 days total with alignment time) during an annual shutdown

  12. Main results of the analysis of internal flooding in the reactor building of Kozloduy NPP Unit 6

    International Nuclear Information System (INIS)

    Demireva, E.; Goranov, S.; Horstmann, R.

    2004-01-01

    For modernization of Units 5 and 6 of Kozloduy NPP, a comprehensive analysis of internal flooding scenarios has been carried out for the reactor building outside the containment and for the turbine hall by FRAMATOME ANP and ENPRO Consult. The objective of the presentation is to provide information on the main results obtained in the flooding analysis of the reactor building (outside containment). The flooding analysis is being performed under application of the 'Methodology and boundary conditions'. Flooding calculations are provided for all of the rooms in the reactor building outside the containment in which the fluid systems, having the capacity for flooding, are mounted. The performed functional analysis shows whether the consequences of a postulated initial event are within the NPP design or could lead to situations which are not taken into account in the design. The proposals for overcoming of identified unacceptable situations and the possible strategy of room draining are also given. Several cases of leaks inside the sealed rooms in the restricted area lead to the situation that the rooms will get totally flooded. Even if this should be acceptable from the point of view of loss of system function, the water pressure effect on the structural elements, as walls and doors, does not allow such complete filling-up. The second relevant identified effect was spreading of humidity and high temperatures to adjacent rooms. Long-lasting effects of this type have to be avoided, in order to prevent potential common cause effects on safety system equipment (authors)

  13. Nonlinear analysis of a reactor building for airplane impact loadings

    International Nuclear Information System (INIS)

    Zimmermann, T.; Rodriguez, C.; Rebora, B.

    1981-01-01

    The purpose is to analyze the influence of material nonlinear behavior on the response of a reinforced concrete reactor building and on equipment response for airplane impact loadings. Two analyses are performed: first, the impact of a slow-flying commercial airplane (Boeing 707), then the impact of a fast flying military airplane (Phantom). (orig./HP)

  14. Dynamic Analysis of AP1000 Shield Building Considering Fluid and Structure Interaction Effects

    Directory of Open Access Journals (Sweden)

    Qiang Xu

    2016-02-01

    Full Text Available The shield building of AP1000 was designed to protect the steel containment vessel of the nuclear reactor. Therefore, the safety and integrity must be ensured during the plant life in any conditions such as an earthquake. The aim of this paper is to study the effect of water in the water tank on the response of the AP1000 shield building when subjected to three-dimensional seismic ground acceleration. The smoothed particle hydrodynamics method (SPH and finite element method (FEM coupling method is used to numerically simulate the fluid and structure interaction (FSI between water in the water tank and the AP1000 shield building. Then the grid convergence of FEM and SPH for the AP1000 shield building is analyzed. Next the modal analysis of the AP1000 shield building with various water levels (WLs in the water tank is taken. Meanwhile, the pressure due to sloshing and oscillation of the water in the gravity drain water tank is studied. The influences of the height of water in the water tank on the time history of acceleration of the AP1000 shield building are discussed, as well as the distributions of amplification, acceleration, displacement, and stresses of the AP1000 shield building. Research on the relationship between the WLs in the water tank and the response spectrums of the structure are also taken. The results show that the high WL in the water tank can limit the vibration of the AP1000 shield building and can more efficiently dissipate the kinetic energy of the AP1000 shield building by fluid-structure interaction.

  15. Environmental assessment for decontamination of the Three Mile Island Unit 2 reactor building atmosphere. Draft NRC staff report for public comment

    International Nuclear Information System (INIS)

    1980-03-01

    The krypton-85 (Kr-85) released to the reactor building during the accident at TMI-2 must be removed from the reactor building in order to permit greater access to the building than is currently possible. The gases currently in the building emit sufficient radiation (1.2 rem/hr total body, 150 rad/hr skin dose) that occupation of the reactor building is severely limited even with protective clothing. Greater access is likely to be necessary to maintain instrumentation and equipment required to keep the reactor in a safe shutdown condition. In addition greater access would facilitate the gathering of data needed for planning the building decontamination program. An additional consideration is that prolonged enclosure of the Kr-85 within the building greatly increases the risk of its successive uncontrolled releases to the outside environment. The staff's evaluation of alternative methods for removing the krypton shows that each could be implemented with little risk to the health and safety of the public. The reactor building purge system, charcoal adsorption system, gas compression, selective absorption process system, and cryogenic processing system could each be operated to keep levels of airborne radioactive materials to unrestricted areas in compliance with the requirements of 10 CFR Part 20, and the design objectives of Appendix 1 to 10 CFR Part 50 of the Commission's regulations, and with the applicable requirements of 40 CFR Part 190.10

  16. Environmental assessment for decontamination of the Three Mile Island Unit 2 reactor building atmosphere. Draft NRC staff report for public comment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1980-03-01

    The krypton-85 (Kr-85) released to the reactor building during the accident at TMI-2 must be removed from the reactor building in order to permit greater access to the building than is currently possible. The gases currently in the building emit sufficient radiation (1.2 rem/hr total body, 150 rad/hr skin dose) that occupation of the reactor building is severely limited even with protective clothing. Greater access is likely to be necessary to maintain instrumentation and equipment required to keep the reactor in a safe shutdown condition. In addition greater access would facilitate the gathering of data needed for planning the building decontamination program. An additional consideration is that prolonged enclosure of the Kr-85 within the building greatly increases the risk of its successive uncontrolled releases to the outside environment. The staff's evaluation of alternative methods for removing the krypton shows that each could be implemented with little risk to the health and safety of the public. The reactor building purge system, charcoal adsorption system, gas compression, selective absorption process system, and cryogenic processing system could each be operated to keep levels of airborne radioactive materials to unrestricted areas in compliance with the requirements of 10 CFR Part 20, and the design objectives of Appendix 1 to 10 CFR Part 50 of the Commission's regulations, and with the applicable requirements of 40 CFR Part 190.10.

  17. Exposure mode study to xenon-133 in a reactor building

    International Nuclear Information System (INIS)

    Perier, Aurelien

    2014-01-01

    The work described in this thesis focuses on the external and internal dose assessment to xenon-133. During the nuclear reactor operation, fission products and radioactive inert gases, as 133 Xe, are generated and might be responsible for the exposure of workers in case of clad defect. Particle Monte Carlo transport code is adapted in radioprotection to quantify dosimetric quantities. The study of exposure to xenon-133 is conducted by using Monte-Carlo simulations based on GEANT4, an anthropomorphic phantom, a realistic geometry of the reactor building, and compartmental models. The external exposure inside a reactor building is conducted with a realistic and conservative exposure scenario. The effective dose rate and the eye lens equivalent dose rate are determined by Monte-Carlo simulations. Due to the particular emission spectrum of xenon-133, the equivalent dose rate to the lens of eyes is discussed in the light of expected new eye dose limits. The internal exposure occurs while xenon-133 is inhaled. The lungs are firstly exposed by inhalation, and their equivalent dose rate is obtained by Monte-Carlo simulations. A biokinetic model is used to evaluate the internal exposure to xenon-133. This thesis gives us a better understanding to the dosimetric quantities related to external and internal exposure to xenon-133. Moreover the impacts of the dosimetric changes are studied on the current and future dosimetric limits. The dosimetric quantities are lower than the current and future dosimetric limits. (author)

  18. Clinch River Breeder Reactor Plant: a building block in nuclear technology

    International Nuclear Information System (INIS)

    McCormack, M.

    1979-01-01

    Interest in breeder reactors dates from the Manhatten Project to the present effort to build the Clinch River Liquid Metal Fast Breeder Reactor (LMFBR) demonstration plant. Seven breeder-type reactors which were built during this time are described and their technological progress assessed. The Clinch River Breeder Reactor Project (CRBRP) has been designed to demonstrate that it can be licensed, can operate on a large power grid, and can provide industry with important experience. As the next logical step in LMFBR development, the project has suffered repeated cancellation efforts with only minor modifications to its schedule. Controversies have developed over the timing of a large-scale demonstration plant, the risks of proliferation, economics, and other problems. Among the innovative developments adopted for the CRBRP is a higher thermal efficiency potential, the type of development which Senator McCormack feels justifies continuing the project. He argues that the nuclear power program can and should be revitalized by continuing the CRBRP

  19. Clinch River Breeder Reactor Plant: a building block in nuclear technology

    Energy Technology Data Exchange (ETDEWEB)

    McCormack, M.

    1979-01-01

    Interest in breeder reactors dates from the Manhatten Project to the present effort to build the Clinch River Liquid Metal Fast Breeder Reactor (LMFBR) demonstration plant. Seven breeder-type reactors which were built during this time are described and their technological progress assessed. The Clinch River Breeder Reactor Project (CRBRP) has been designed to demonstrate that it can be licensed, can operate on a large power grid, and can provide industry with important experience. As the next logical step in LMFBR development, the project has suffered repeated cancellation efforts with only minor modifications to its schedule. Controversies have developed over the timing of a large-scale demonstration plant, the risks of proliferation, economics, and other problems. Among the innovative developments adopted for the CRBRP is a higher thermal efficiency potential, the type of development which Senator McCormack feels justifies continuing the project. He argues that the nuclear power program can and should be revitalized by continuing the CRBRP.

  20. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    International Nuclear Information System (INIS)

    Joo, Young Sang; Lee, Jae Han

    2007-02-01

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification

  1. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young Sang; Lee, Jae Han

    2007-02-15

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification.

  2. Aircraft Impact Assessment of APR1400 Reactor Containment Building

    International Nuclear Information System (INIS)

    Moon, Il Hwan; Kim, Do Yeon; Kim, Jae Hee; Kim, Sang Yun

    2011-01-01

    The implementation of a protection to withstand aircraft impact on safety-related structures and systems is basically based on a probabilistic evaluation for each site, if the licensing body doesn't require a deterministic approach. Existing nuclear power plants in Korea were designed based on the probabilistic approach, and the aircraft impact hazard remained less than a probability of 10 -7 . However, a man-made aircraft impact have been considered as a possible external accident for the nuclear power plant. New plant designs that are to be constructed in the U.S. after July 2009 must consider the effect of impact from a large commercial aircraft according to the requirements of 10 CFR 50.150. Especially, Reactor Containment Building (RCB) housing the safety-related equipment and fuels should be protected safely against aircraft crash without perforation and scabbing failure of external wall. APR1400 RCB is constructed as a prestressed concrete containment vessel (PCCV) which is surrounded by the auxiliary building housing additional safety-related equipment and other systems. In this study, the aircraft impact analyses for the RCB are carried out using Riera forcing function and aircraft model. Considered external wall thickness is 4 ft 6 in. for the cylindrical wall and 4 ft for the dome. Actual strengths of concrete and steel are considered as the material properties. For these analyses, the dynamic increment factor and concrete aging effect are considered in accordance with NEI 07-13(2011)

  3. Modelling of internal structure in seismic analysis of a PHWR building

    International Nuclear Information System (INIS)

    Reddy, G.R.; Vaze, K.K.; Kushawaha, H.S.; Ingle, R.K.; Subramanian, K.V.

    1991-01-01

    Seismic analysis of complex and large structures, consisting of thick shear walls, such as Reactor Building is very involved and time consuming. It is a standard practice to model the structure as a stick model to predict reasonably the dynamic behaviour of the structure. It is required to determine approximate equivalent sectional properties of Internal Structure for representation in the stick model. The restraint to warping can change the stress distribution thus affecting the centre of rigidity and torsional inertia, Hence, standard formulae does not hold good for determination of sectional properties of the Internal Structure. In this case the equivalent sectional properties for the Internal Structure are calculated using a Finite Element Model (FEM) of the Internal Structure and applying unit horizontal forces in each direction. A 3-D stick model is developed using the guidelines. Using the properties calculated by FEM and also by standard formulae, the responses of the 3-D stick model are compared. (J.P.N.)

  4. Positron annihilation studies on structural materials for nuclear reactors

    International Nuclear Information System (INIS)

    Rajaraman, R.; Amarendra, G.; Sundar, C.S.

    2012-01-01

    Structural steels for nuclear reactors have renewed interest owing to the future advanced fission reactor design with increased burn-up goals as well as for fusion reactor applications. While modified austenitic steels continue to be the main cladding materials for fast breeder reactors, Ferritic/martensitic steels and oxide dispersion strengthened ferritic steels are the candidate materials for future reactors applications in India. Sensitivity and selectivity of positron annihilation spectroscopy to open volume type defects and nano clusters have been extensively utilized in studying reactor materials. We have recently reviewed the application of positron techniques to reactor structural steels. In this talk, we will present successful application of positron annihilation spectroscopy to probe various structural materials such as D9, ferritic/martensitic, oxide dispersion strengthened (ODS) steels and related model alloys, highlighting our recent studies. (author)

  5. Structural Integrity Assessment of Reactor Containment Subjected to Aircraft Crash

    International Nuclear Information System (INIS)

    Kim, Junyong; Chang, Yoonsuk

    2013-01-01

    When an accident occurs at the NPP, containment building which acts as the last barrier should be assessed and analyzed structural integrity by internal loading or external loading. On many occasions that can occur in the containment internal such as LOCA(Loss Of Coolant Accident) are already reflected to design. Likewise, there are several kinds of accidents that may occur from the outside of containment such as earthquakes, hurricanes and strong wind. However, aircraft crash that at outside of containment is not reflected yet in domestic because NPP sites have been selected based on the probabilistic method. After intentional aircraft crash such as World Trade Center and Pentagon accident in US, social awareness for safety of infrastructure like NPP was raised world widely and it is time for assessment of aircraft crash in domestic. The object of this paper is assessment of reactor containment subjected to aircraft crash by FEM(Finite Element Method). In this paper, assessment of structural integrity of containment building subjected to certain aircraft crash was carried out. Verification of structure integrity of containment by intentional severe accident. Maximum stress 61.21MPa of horizontal shell crash does not penetrate containment. Research for more realistic results needed by steel reinforced concrete model

  6. Preliminary conceptual design and analysis on KALIMER reactor structures

    International Nuclear Information System (INIS)

    Kim, Jong Bum

    1996-10-01

    The objectives of this study are to perform preliminary conceptual design and structural analyses for KALIMER (Korea Advanced Liquid Metal Reactor) reactor structures to assess the design feasibility and to identify detailed analysis requirements. KALIMER thermal hydraulic system analysis results and neutronic analysis results are not available at present, only-limited preliminary structural analyses have been performed with the assumptions on the thermal loads. The responses of reactor vessel and reactor internal structures were based on the temperature difference of core inlet and outlet and on engineering judgments. Thermal stresses from the assumed temperatures were calculated using ANSYS code through parametric finite element heat transfer and elastic stress analyses. While, based on the results of preliminary conceptual design and structural analyses, the ASME Code limits for the reactor structures were satisfied for the pressure boundary, the needs for inelastic analyses were indicated for evaluation of design adequacy of the support barrel and the thermal liner. To reduce thermal striping effects in the bottom are of UIS due to up-flowing sodium form reactor core, installation of Inconel-718 liner to the bottom area was proposed, and to mitigate thermal shock loads, additional stainless steel liner was also suggested. The design feasibilities of these were validated through simplified preliminary analyses. In conceptual design phase, the implementation of these results will be made for the design of the reactor structures and the reactor internal structures in conjunction with the thermal hydraulic, neutronic, and seismic analyses results. 4 tabs., 24 figs., 4 refs. (Author)

  7. Seismic response analysis of reactor containment structures - axisymmetric model with modified ground motion

    International Nuclear Information System (INIS)

    Saha, S.; Dasgupta, A.; Basu, P.C.

    1993-01-01

    Seismic analysis of a Reactor Building is performed idealising the system as a beam model (BM) and also an Axi-symmetric model (ASM) and the results compared. In both the cases effect of Soil-Structure Interaction have been taken Into account. Since the lower boundary of the ASM was at a depth much lower than that of the BM, deconvolution of the specified Free-Field Motion (FFM) was necessary. The deconvolution has been performed using frequency domain approach. (author)

  8. ORGANIZATIONAL STRUCTURE FOR BUILDINGS RECONSTRUCTION OF HISTORICAL BUILDING OF ODESSA

    Directory of Open Access Journals (Sweden)

    POSTERNAK I. М.

    2016-12-01

    Full Text Available Formulation of the problem. As one of perspective forms of integration various complexes act in town-planning structure. In the course of formation of plans of social and economic development of large cities even more often there is a situation when for increase of efficiency of used resources concentration of efforts is necessary not simply, but also new progressive forms of the organization of building manufacture. Purpose. To offer the organizational structure using in practice the saved up scientific and technical potential for reconstruction of buildings of historical building of Odessa 1820 … 1920 years under standards power efficiency and to execute researches engineering architectonics residential buildings of historical building of a city of Odessa. Conclusion. It is offered to create in the city of Odessa "the Corporate scientific and technical complex town-planning power reconstruction "CSTC T-PPR", as innovative organizational structure which uses in practice the saved up scientific and technical potential for reconstruction of buildings of historical building of Odessa under standards power efficiency. It is considered engineering architectonics residential buildings of historical building of a city of Odessa, in particular, not looking on diverse buildings of inhabited appointment of Odessa, for them there are defining factors on which probably to make their grouping and at the same time to allocate the general lines inherent to a housing estate as a whole. It is resulted a general characteristic and classification of residential buildings of historical building of a city of Odessa ХІХ … beginnings ХХ centuries It is allocated and expanded classification of such buildings of inhabited appointment by duration of residing at them.

  9. ORGANIZATIONAL STRUCTURE FOR RECONSTRUCTION OF BUILDINGS HISTORICAL BUILDING OF ODESSA

    Directory of Open Access Journals (Sweden)

    POSTERNAK I. М.

    2017-05-01

    Full Text Available Summary. Raising of problem. As one of perspective forms of integration various complexes act in town- planning structure. In the course of formation of plans of social and economic development of large cities even more often there is a situation when for increase of efficiency of used resources concentration of efforts is necessary not simply, but also new progressive forms of the organization of building manufacture. Purpose. To offer the organizational structure using in practice the saved up scientific and technical potential for reconstruction of buildings of historical building of Odessa 1820 … 1920 years under standards power efficiency and to execute researches engineering architectonics residential buildings of historical building of a city of Odessa. Conclusion. It is offered to create in the city of Odessa "the Corporate scientific and technical complex town-planning power reconstruction "CSTC T-PPR", as innovative organizational structure which uses in practice the saved up scientific and technical potential for reconstruction of buildings of historical building of Odessa under standards power efficiency. It is considered engineering architectonics residential buildings of historical building of a city of Odessa, in particular, not looking on diverse buildings of inhabited appointment of Odessa, for them there are defining factors on which probably to make their grouping and at the same time to allocate the general lines inherent to a housing estate as a whole. It is resulted a general characteristic and classification of residential buildings of historical building of a city of Odessa ХІХ beginnings ХХ centuries It is allocated and expanded classification of such buildings of inhabited appointment by duration of residing at them.

  10. Design study of plant system for the fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    Iida, Hiromasa; Kuroda, Hideo; Yamada, Masao; Suzuki, Tatsushi; Honda, Tsutomu; Ohmura, Hiroshi; Itoh, Shinichi.

    1986-11-01

    This report describes design study results of the FER plant system. The purpose of this study is to have an image of the FER plant system as a whole by designing major auxiliary systems, reactor building and maintenance and radwaste desposal systems. The major auxiliary systems include tritium, cooling, evacuation and fueling systems. For these each systems, flowdiagrams are studied and designs of devices and pipings are conducted. In the reactor building design, layout of the above auxiliary systems in the building is studied with careful zoning concept by the radiation level. Structural integrity of the reactor building is also studied including seismic analysis. In the design of the maintenance and radwaste system flowdiagram of failed reactor components is developed and transfer vehicles and buildings are designed. Finally assuming JAERI Naka site as the reactor site layout of the whole FER plant system is developed. (author)

  11. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Decreton, M.

    2002-01-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed

  12. Nuclear reactor structural material forming less radioactive corrosion product

    International Nuclear Information System (INIS)

    Nakazawa, Hiroshi.

    1988-01-01

    Purpose: To provide nuclear reactor structural materials forming less radioactive corrosion products. Constitution: Ni-based alloys such as inconel alloy 718, 600 or inconel alloy 750 and 690 having excellent corrosion resistance and mechanical property even in coolants at high temperature and high pressure have generally been used as nuclear reactor structural materials. However, even such materials yield corrosion products being attacked by coolants circulating in the nuclear reactor, which produce by neutron irradiation radioactive corrosion products, that are deposited in primary circuit pipeways to constitute exposure sources. The present invention dissolves dissolves this problems by providing less activating nuclear reactor structural materials. That is, taking notice on the fact that Ni-58 contained generally by 68 % in Ni changes into Co-58 under irradiation of neutron thereby causing activation, the surface of nuclear reactor structural materials is applied with Ni plating by using Ni with a reduced content of Ni-58 isotopes. Accordingly, increase in the radiation level of the nuclear reactor structural materials can be inhibited. (K.M.)

  13. Management of research reactor; dynamic characteristics analysis for reactor structures related with vibration of HANARO fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Chang Kee; Shim, Joo Sup [Shinwa Technology Information, Seoul (Korea)

    2001-04-01

    The objective of this study is to deduce the dynamic correlation between the fuel assembly and the reactor structure. Dynamic characteristics analyses for reactor structure related with vibration of HANARO fuel assembly have been performed For the dynamic characteristic analysis, the in-air models of the round and hexagonal flow tubes, 18-element and 36-element fuel assemblies, and reactor structure were developed. By calculating the hydrodynamic mass and distributing it on the in-air models, the in-water models of the flow tubes, the fuel assemblies, and the reactor structure were developed. Then, modal analyses for developed in-air and in-water models have been performed. Especially, two 18-element fuel assemblies and three 36-element fuel assemblies were included in the in-water reactor models. For the verification of the modal analysis results, the natural frequencies and the mode shapes of the fuel assembly were compared with those obtained from the experiment. Finally the analysis results of the reactor structure were compared with them performed by AECL Based on the reactor model without PCS piping, the in-water reactor model including the fuel assemblies was developed, and its modal analysis was performed. The analysis results demonstrate that there are no resonance between the fuel assembly and the reactor structures. 26 refs., 419 figs., 85 tabs. (Author)

  14. 7 CFR 51.56 - Buildings and structures.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 2 2010-01-01 2010-01-01 false Buildings and structures. 51.56 Section 51.56... § 51.56 Buildings and structures. The packing plant buildings shall be properly constructed and... be sufficient light consistent with the use to which the particular portion of the building is...

  15. Assessment of extent and degree of thermal damage to polymeric materials in the Three Mile Island Unit 2 reactor building

    International Nuclear Information System (INIS)

    Alvares, N.J.

    1986-01-01

    This paper describes assumptions and procedures used to perform thermal damage analysis caused by post loss-of-coolant-accident (LOCA) hydrogen deflagration at Three Mile Island Unit 2 Reactor. Examination of available photographic evidence yields data on the extent and range of thermal and burn damage. Thermal damage to susceptible material in accessible regions of the reactor building was distributed in non-uniform patterns. No clear explanation for non-uniformity was found in examined evidence, e.g., burned materials were adjacent to materials that appear similar but were not burned. Because these items were in proximity to vertical openings that extend the height of the reactor building, the authors assume the unburned materials preferentially absorbed water vapor during periods of high, local steam concentration. A control pendant from the polar crane located in the top of the reactor building sustained asymmetric burn damage of decreasing degree from top to bottom. Evidence suggests the polar-crane pendant side that experienced heaviest damage was exposed to intense radiant energy from a transient fire plume in the reactor containment volume. Simple hydrogen-fire-exposure tests and heat transfer calculations approximate the degree of damage found on inspected materials from the containment building and support for an estimated 8% pre-fire hydrogen

  16. Assessment of extent and degree of thermal damage to polymeric materials in the Three Mile Island Unit 2 Reactor building

    International Nuclear Information System (INIS)

    Alvares, N.J.

    1985-06-01

    This paper describes assumptions and procedures used to perform thermal damage analysis caused by post loss-of-coolant-accident (LOCA) hydrogen deflagration at Three Mile Island Unit 2 Reactor. Examination of available photographic evidence yields data on the extent and range of thermal and burn damage. Thermal damage to susceptible material in accessible regions of the reactor building was distributed in non-uniform patterns. No clear explanation for non-uniformity was found in examined evidence, e.g., burned materials were adjacent to materials that appear similar but were not burned. Because these items were in proximity to vertical openings that extend the height of the reactor building, we assume the unburned materials preferentially absorbed water vapor during periods of high, local steam concentration. A control pendant from the polar crane located in the top of the reactor building sustained asymmetric burn damage of decreasing degree from top to bottom. Evidence suggests the polar-crane pendant side that experienced heaviest damage was exposed to intense radiant energy from a transient fire plume in the reactor containment volume. Simple hydrogen-fire-exposure tests and heat transfer calculations approximate the degree of damage found on inspected materials from the containment building and support for an estimated 8% pre-fire hydrogen

  17. Estimation of radioactivity in structural materials of ETRR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Imam, M [National Center for Nuclear Safety and Radiation Control Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    Precise knowledge of the thermal neutron flux in the different structural materials of a reactor is necessary to estimate the radioactive inventory in these materials that are needed in any decommissioning study of the reactor. ETRR-1 is a research reactor that went critical on 2/1691. In spite of this long age of the reactor, the effective operation time of this reactor is very short since the reactor was shutdown for long periods. Because of this long age one may think of reactor decommissioning. For this purpose, the radioactivity of the reactor structural materials was estimated. Apart from the reactor core, the important structural materials in the ETRR-1 are the reactor tank, shielding concrete, and the graphite thermal column. The thermal neutron flux was determined by the monte Carlo method in these materials and the isotope inventory and the radioactivity were calculated by the international code ORIGEN-JR. 1 fig.

  18. Fiber reinforced concrete as a material for nuclear reactor containment buildings

    International Nuclear Information System (INIS)

    Mallikarjuna; Banthia, N.; Mindess, S.

    1991-01-01

    The fiber reinforced concrete as a constructional material for nuclear reactor containment buildings calls for an examination of its individual characteristics and potentialities due to its inherent superiority over normal plain and reinforced concrete. In the present investigation, first, to study the static behavior of straight, hooked-end and crimped fibers, recently developed nonlinear three-dimensional interface (contact) element has been used in conjunction with the eight nodded hexahedron and two nodded bar elements for concrete and steel fiber respectively. Then impact tests were carried out on fiber reinforced concrete beams with an instrumented drop weight impact machine. Two different concrete mixes were tested: normal strength and high strength concrete specimens. Fibers in the concrete mix found to significantly increase the ductility and the impact resistance of the composite. Deformed fibers increase peak pull-out load and pull-out distance, and perform better in the steel fiber reinforced concrete (SFRC) structures. (author)

  19. Reactor building pressure proof test (PPT) and leak rate test (LRT) of Qinshan phase III (CANDU) project

    International Nuclear Information System (INIS)

    Gu Jun; Shi Jinqi; Fan Fuping

    2004-12-01

    As the first reactor building (R/B) without stainless steel liner in china, TQNPC studied the containment characteristics, such as strong concrete absorb/release air effect, poor containment penetration. etc. And carefully prepared test scheme and emergency response, creatively introduced the instrument air self-supply system in reactor building, developed the special measurement and analysis system for PPT and LRT, organized work under high-pressure on large-scale in the test. Finally got the containment leak rate result and the test-cost-time value is the best in all same type tests. (authors)

  20. Savannah River Site production reactor safety analysis report. K production reactor

    International Nuclear Information System (INIS)

    1996-01-01

    This section provides the structural criteria for the K-Area buildings that are common to Seismic Category I structures. Exceptions to this criteria for specific buildings are given in Sections 3.8.1 through 3.8.4. The original SRS buildings and structures were designed and constructed before current nuclear codes or standards were developed. However, to withstand a bomb attack, a blast-resistant classification and loading were imposed in the structural criteria. The blast-resistant construction of Buildings 105 and 108 provides significant resistance to earthquake and tornado conditions. In fact, this premise was the basis for qualifying the structures classified as Seismic Category I when the buildings were designed and constructed in the 1950s. A classification of buildings and structures according to the blast and seismic resistance criterion is presented in Table 3.8-1. Knowledge gained in seismic technology since the 1950s has resulted in methods and techniques to design and qualify structures and to develop seismic loading for equipment and piping within those structures. Many seismic analyses of SRS buildings and structures have applied the emerging seismic technologies at various levels of sophistication. A review of seismic analyses and qualifications of the K-reactor building, stack building, cooling water reservoir structures, and process effluent sump structure is required to confirm that these structures can resist the postulated design basis earthquake (DBE). If the capacity of a structure to withstand the DBE cannot be confirmed from the review, additional analysis or design modifications are required. Such actions are governed by the Interim Seismic Program. The seismic design basis for the new evaluations is provided in Section 3.7.2

  1. Study of ex-vessel steam explosion risk of Reactor Pit Flooding System and structural response of containment for CPR1000"+ Unit

    International Nuclear Information System (INIS)

    Zhang Juanhua; Chen Peng

    2015-01-01

    Reactor Pit Flooding System is one of the special mitigation measures for severe accident for CPR1000"+ Unit. If the In-Vessel Relocation function of Reactor Pit Flooding System is failed, there is the steam explosion risk in reactor cavity. This paper firstly adopts MC3D code to build steam explosion model in order to calculate the pressure load and impulses of steam explosion that are as the input data of containment structural response analysis. The next step is to model the containment structure and analyze the structural response by ABAQUS code. The analysis results show that the integral damage induced by steam explosion to the external containment wall is shallow, and the containment structural integrity can be maintained. The risk and damage to the containment integrity reduced by steam explosion of RPF is small, and it does not influence the design and implementation of RPF. (author)

  2. Mechanics of structures and maintenance of pressurized water reactors

    International Nuclear Information System (INIS)

    Hutin, J.P.

    1992-01-01

    Electricite de France nowadays has in operation 34 units of 900 MW and 17 units of 1300 MW of PWR. Since the first unit was run, this means that more than 350 reactor-years have been performed, to which should be added the experience already gained on fossil fuel or natural uranium plants. This enabled EDF to build its own philosophy and a strategy for maintenance that are best suited for the specific requirements of the hardware with which the actual nuclear boilers are made-up. This philosophy and strategy rest upon an analysis which calls widely for the mechanics of structures, to such an extent that major decisions concerning maintenance depend on the ability that one has for resolving problems within the scope of that discipline

  3. 3-dimensional earthquake response analysis of embedded reactor building using hybrid model of boundary elements and finite elements

    International Nuclear Information System (INIS)

    Muto, K.; Motosaka, M.; Kamata, M.; Masuda, K.; Urao, K.; Mameda, T.

    1985-01-01

    In order to investigate the 3-dimensional earthquake response characteristics of an embedded structure with consideration for soil-structure interaction, the authors have developed an analytical method using 3-dimensional hybrid model of boundary elements (BEM) and finite elements (FEM) and have conducted a dynamic analysis of an actual nuclear reactor building. This paper describes a comparative study between two different embedment depths in soil as elastic half-space. As the results, it was found that the earthquake response intensity decreases with the increase of the embedment depth and that this method was confirmed to be effective for investigating the 3-D response characteristics of embedded structures such as deflection pattern of each floor level, floor response spectra in high frequency range. (orig.)

  4. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2002-04-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed.

  5. Pre-service proof pressure and leak rate tests for the Qinshan CANDU project reactor buildings

    International Nuclear Information System (INIS)

    Petrunik, K.J.; Khan, A.; Ricciuti, R.; Ivanov, A.; Chen, S.

    2003-01-01

    The Qinshan CANDU Project Reactor Buildings (Units 1 and 2) have been successfully tested for the Pre-Service Proof Pressure and Integrated Leak Rate Tests. The Unit 1 tests took place from May 3 to May 9, 2002 and from May 22 to May 25, 2002, and the Unit 2 tests took place from January 21 to January 27, 2003. This paper discusses the significant steps taken at minimum cost on the Qinshan CANDU Project, which has resulted in a) very good leak rate (0.21%) for Unit 1 and excellent leak rate (0.130%) for Unit 2; b) continuous monitoring of the structural behaviour during the Proof Pressure Test, thus eliminating any repeat of the structural test due to lack of data; and c) significant schedule reduction achieved for these tests in Unit 2. (author)

  6. Reactor-building-basement radionuclide and source distribution studies. Volume 3

    International Nuclear Information System (INIS)

    Cox, T.E.; Horan, J.T.; Worku, G.

    1983-06-01

    The Three Mile Island Unit 2 (TMI-2) Reactor Building basement has been sampled several times since August 1979. This report compiles the analytical results and sample history for the liquid and solid samples obtained to date. In addition, basement radiation levels were also obtained using thermoluminescent dosimeters (TLDs). The data obtained will provide information to support ongoing mass balance and source term studies and will aid in characterizing the 282-ft elevation for decontamination planning and dose reduction

  7. HELB Analysis for ESBWR Reactor Building and Main Steam Tunnel

    Energy Technology Data Exchange (ETDEWEB)

    Noguera Oliva, O.

    2011-07-01

    The Reactor Building compartments and tbe Main Steam Tunnel are modeled using GOTHIC 7.2a. These models are based on Control Volumes (Rooms/Compartments/Regions), Flow Paths (junctions such as vent path or any opening) and Boundary Conditions (Mass and energy releases and outside conditions). Due to the different break locations, four models are built to analyze the short-term pressurization response. Are shown the cases analyzed, the results obtained and the models used for this purpose.

  8. Precautions against axial fan stall in reactor building to Tianwan NPP

    International Nuclear Information System (INIS)

    Liu Chunlong; Pei Junmin

    2011-01-01

    The paper introduces the mechanism and harm of rotating stall of axial fans, analyzes the necessity for prevention against axial fan stall in reactor building of Tianwan NPP, introduces the precautions, and then makes an assessment on anti-stall effect of flow separators. It can provide reference for model-selection or reconstruction of similar fans in power stations, and for operation and maintenance of axial fans. (authors)

  9. Effect of the flexibility of the base mat on seismic response of a PWR-reactor building

    International Nuclear Information System (INIS)

    Waas, G.; Riggs, H.R.

    1983-01-01

    The flexibility of the base mat influences the stiffness and the radiation damping of foundations, In this paper its effect on the seismic response of an axisymmetric PWR-reactor building is investigated. The base mat of the building is stiffened by cylindrical concrete walls and by a rigid block in the center. Soft and stiff soil conditions are considered. The structure and its foundation are modelled by axisymmetric shell and volume elements with Fourier expansions in the circumferential direction. The soil is treated as a horizontally layered viscoelastic medium. Soil and structure are coupled along nodal rings. The stiffness matrix of the soil is computed using an explicit semi-analytic solution for displacements caused by ring loads acting on the surface or within a layered medium. The analysis is performed in the frequency domain, and the response in the time domain is computed by the fast Fourier transformation. The earthquake response is computed with and without including the flexibility of the relatively stiff base mat. The comparison shows that including the flexibility of the mat has hardly any effect on the resonant frequencies and the damping of the fundamental rocking and vertical modes. This is the case for soft and stiff soil conditions. However, the flexibility of the mat strongly affects the first structural deformation mode, in which the external and internal structures deflect in opposite directions. (orig./HP)

  10. Ventilation system in the RA reactor building - design specifications; Sistem ventilacije u objektu 'RA' - Tehnicki opis

    Energy Technology Data Exchange (ETDEWEB)

    Badrljica, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1984-09-15

    Protective role of the ventilation system of nuclear facilities involve construction of ventilation barriers which prevent release of radioactive particulates or gases, elimination od radioactive particulates and gases from the air which is released from contaminated zones into the reactor environment. Ventilation barriers are created by dividing the building into a number of ventilation zones with different sub pressure compared to the atmospheric pressure. The RA reactor building is divided into four ventilation zones. First zone is the zone of highest risk. It includes reactor core with horizontal experimental channels, underground rooms of the primary coolant system (D{sub 2}O), helium system, hot cells and the space above the the reactor core. Second zone is the reactor hall and the room for irradiated fuel storage. The third zone includes corridors in the basement, ground floor and first floor where the probability of contamination is small. The fourth zone includes the annex where the contamination risk is low. There is no have natural air circulation in the reactor building. Ventilators for air input and outlet maintain the sub pressure in the building (pressure lower than the atmospheric pressure). This prevents release of radioactivity into the atmosphere. Zastitne uloge ventilacionog sistema kod nuklearnih postrojenja obuhvataju formiranje ventilacionih barijera koje onemugucavaju sirenje radioaktivnih cestica ili gasova putem cirkulacije vazduha; eliminaciju radioaktivnih cestica i gasova iz vazduha koji se evakuise iz kontaminiranih prostora u okolinu reaktorskog postrojenja. Formiranje zastitnih ventilacionih barijera ostvaruje se obicno podelom unutrasnjosti objekta na vise ventilacionih zona razlicitih podpritisaka u odnosu na spoljni atmosferski pritisak. Celi prostor zgrade reaktora RA podeljen je u cetiri ventilacione zone. Prva zona je zona najveceg rizika, u koju spadaju reaktorsko jezgro sa horizontalnim eksperimentalnim kanalima, tehnoloske

  11. Development of technology for next generation reactor - Development of next generation reactor in Korea -

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyun; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); and others

    1993-09-01

    The project, development of next generation reactor, aims overall related technology development and obtainment of related license in 2001. The development direction is to determine the reactor type and to build up the design concept in 1994. For development trend analysis of foreign next generation reactor, level-1 PSA, fuel cycle analysis and computer code development are performed on System 80+ and AP 600. Especially for design characteristics analysis and volume upgrade of AP 600, nuclear fuel and reactor core design analysis, coolant circuit design analysis, mechanical structure design analysis and safety analysis etc. are performed. (Author).

  12. An analysis of the estimated capital cost of a fusion reactor

    International Nuclear Information System (INIS)

    Hollis, A.A.

    1981-06-01

    The cost of building a fusion reactor similar to the Culham Conceptual Tokamak reactor Mark IIB is assessed and compared with other published capital costs of fusion and fission reactors. It is concluded that capital-investment and structure-renewal costs for a typical fusion reactor as presently conceived are likely to be higher than for thermal-fission reactors. (author)

  13. An analysis of the estimated capital cost of a fusion reactor

    International Nuclear Information System (INIS)

    Hollis, A.A.; Evans, L.S.

    1981-01-01

    The cost of building a fusion reactor similar to the Culham Conceptual Tokamak reactor Mark IIB is assessed and compared with other published capital costs of fusion and fission reactors. It is concluded that capital-investment and structure-renewal costs for a typical fusion reactor as presently conceived are likely to be higher than for thermal-fission reactors. (author)

  14. Structure and creep of Russian reactor steels with a BCC structure

    Science.gov (United States)

    Sagaradze, V. V.; Kochetkova, T. N.; Kataeva, N. V.; Kozlov, K. A.; Zavalishin, V. A.; Vil'danova, N. F.; Ageev, V. S.; Leont'eva-Smirnova, M. V.; Nikitina, A. A.

    2017-05-01

    The structural phase transformations have been revealed and the characteristics of the creep and long-term strength at 650, 670, and 700°C and 60-140 MPa have been determined in six Russian reactor steels with a bcc structure after quenching and high-temperature tempering. Creep tests were carried out using specially designed longitudinal and transverse microsamples, which were fabricated from the shells of the fuel elements used in the BN-600 fast neutron reactor. It has been found that the creep rate of the reactor bcc steels is determined by the stability of the lath martensitic and ferritic structures in relation to the diffusion processes of recovery and recrystallization. The highest-temperature oxide-free steel contains the maximum amount of the refractory elements and carbides. The steel strengthened by the thermally stable Y-Ti nanooxides has a record high-temperature strength. The creep rate at 700°C and 100 MPa in the samples of this steel is lower by an order of magnitude and the time to fracture is 100 times greater than that in the oxide-free reactor steels.

  15. Structural mechanics and reactor safety

    International Nuclear Information System (INIS)

    Brandes, K.

    1983-01-01

    Operational safety and reliability of nuclear power plants widely depend on the mechanical behaviour of their structural components and their resistance to the various and complex influences. Durability and consistency of structural components are determined by the kind of strain - during the life - and by environmental conditions. The Conferences on Structural Mechanics in Reactor Technology (SMiRT) are dedicated to the discussion of such questions. The 7th of these Conferences taking place in 2-year increments was held in Chicago in August 1983. The number of contributions again increased, the number of participants slightly decreased. There are some trends in this field worth mentioning, in particular the fact that experience from design and operation of nuclear power plants now available is more and more made use of, and that more and more attention is given the problems of fusion reactors. (orig./HP) [de

  16. Using half-cell potential measurement to access the severity of corrosion in reinforced concrete structures in Gentilly-2 reactor building

    International Nuclear Information System (INIS)

    Picard, S.; Kadoum, N.; Poirier, F.

    2009-01-01

    The half-cell potential technique has been used to assess the corrosion in the reactor's building ring beam of the Gentilly-2 nuclear power plant. It is a non-destructive technique based on the ASTM C 876 Standard. Corrosion is the result of a difference of potential between anodic and cathodic zones within the re-bars network and these potential differences are measured in the half-cell potential technique. Time exposure is the leading factor and we recommend the installation of permanent electrodes of reference in strategic areas. The results show a low corrosion activity level on 98% of the investigated surface and no severe corrosion potential reading has been registered. Furthermore the exercise shows that the repair technique has no influence on the corrosion activity of the steel network. Since most of the readings are located in the low corrosion activity level (from 0 to -100 mV), it illustrates that there is heterogeneity of the corrosion activity within the ring beam. We recommend a system to monitor the evolution of the corrosion phenomena in real time. The installation of reference electrodes positioned in some ring beam strategic areas is a simple and accurate way of monitoring the corrosion activity of the steel in the structure. In the case where an evolution in higher level is noted in the corrosion activity, it would be possible to act and prevent any further degradation of the structure

  17. Seismic analysis, evaluation and upgrade design for a nuclear facility exhaust stack building

    International Nuclear Information System (INIS)

    Malik, L.E.; Kabir, A.F.

    1991-01-01

    This paper reports on an exhaust stack building of a nuclear reactor facility with complex structural configuration that has been analyzed and evaluated for seismic forces. This building was built in the 1950's and had not been designed to resist seismic forces. A very rigorous analysis and evaluation program was implemented to minimize the costly retrofits required to upgrade the building to resist high seismic forces. The seismic evaluations were performed for the building in its as-is configuration, and as modified for several upgrade schemes. Soil-structure-interaction, base mat flexibility and the influence of the nearby reactor building have been considered in the seismic analyses. The rigorous analyses and evaluation enabled limited upgrades to qualify the stack building for the seismic forces

  18. Construction engineering and planning of buildings related to the EPR

    International Nuclear Information System (INIS)

    Kaercher, H.

    1995-01-01

    Among the site-independent unit buildings are the reactor building with annulus; building for safety systems with main control room; new fuel storage pit, emergency power unit, reactor auxiliaries, access building; conventional switchgear building, and the turbine building. All buildings housing safety-related systems are protected against external and internal influences. Among the design-determining external influences are earthquakes, explosion high-pressure wave and aircraft crash. Internal incidents are caused by failure of components and pipes. The most discussed incident in the connection is the failure of the reactor pressure vessel involving core melt release which is safely retained by special devices. Earlier 3D plant models made of plastic have been replaced by 3D CAE computer models. Thus the graphic data of CAD systems have been added to the immense amount of logistic programmes/process data chains. This leads to new planning tools which are able to safely process such amounts of data and at the same time notably reduce planning time and expense. The whole data processing concept is characterized by simple, consistent data structures according to a uniform data model. It enables continual treatment throughout all planning stages and data exchange through simple, uniformly structured interfaces. (orig./HP) [de

  19. Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Bunz, H.; Dunbar, I.; Gauvain, J.; Ricchena, R.

    1986-01-01

    The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSIM-M, UK; AEROSOLS/B1, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR. Topics considered in this paper include aerosols, containment buildings, reactor safety, fission product release, reactor cores, meltdown, and monitoring

  20. Simquake 3: Seismic interactions between building structures and rock-socketed foundations: Final report

    International Nuclear Information System (INIS)

    Howard, G.E.; Chitty, D.E.; Oleck, R.F.

    1988-04-01

    It has long been recognized that soil-structure interaction can significantly influence the earthquake response of massive structures such as nuclear power plant reactor buildings. The linear analysis methods that are widely used to model interaction phenomena can result in often unrecognized safety margins in design for earthquake excitation. Use of improved interaction models which capture nonlinear characteristics of interaction---such as energy dissipation and significant changes in stiffness---can provide realistic predictions of the earthquake loads imposed on nuclear power plant structures and equipment, supplying an improved basis for seismic design review. This report documents the results of a research effort investigating the soil-structure (or structure-media) interaction of reinforced concrete structures founded in backfilled rock sockets. The objectives of the research, which included field testing with semi-scale structural models, were: to examine the influence of the backfilled socket on structural dynamic response; and to develop an experimental data base for the benchmarking of computer simulation procedures

  1. Challenges and achievements - Prototype Fast Breeder Reactor construction

    International Nuclear Information System (INIS)

    Subramani, V.A.; Dhere, S.S.; Manoharan, V.; Subbaraman, P.

    2010-01-01

    Prototype fast breeder reactor presently under construction poses several challenges in materials, design and construction. The civil structure and equipment are of very large size and complex in nature. This paper presents the features of the design and construction of the PFBR excavation, raft, civil structure of the nuclear island connected buildings and reactor vault. This paper also brings out the details of the large size equipment of special stainless steel and handling structure for their lifting and placement inside the reactor vault. The paper is divided into three parts viz. introduction, challenges and achievements during construction of civil structures and erection of large size components. (author)

  2. Structure optimization of CFB reactor for moderate temperature FGD

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yuan; Zhang, Jie; Zheng, Kai; You, Changfu [Tsinghua Univ., Beijing (China). Dept. of Thermal Engineering; Ministry of Education, Beijing (China). Key Lab. for Thermal Science and Power Engineering

    2013-07-01

    The gas velocity distribution, sorbent particle concentration distribution and particle residence time in circulating fluidized bed (CFB) reactors for moderate temperature flue gas desulfurization (FGD) have significant influence on the desulfurization efficiency and the sorbent calcium conversion ratio for sulfur reaction. Experimental and numerical methods were used to investigate the influence of the key reactor structures, including the reactor outlet structure, internal structure, feed port and circulating port, on the gas velocity distribution, sorbent particle concentration distribution and particle residence time. Experimental results showed that the desulfurization efficiency increased 5-10% when the internal structure was added in the CFB reactor. Numerical analysis results showed that the particle residence time of the feed particles with the average diameter of 89 and 9 {mu}m increased 40% and 17% respectively, and the particle residence time of the circulating particles with the average diameter of 116 {mu}m increased 28% after reactor structure optimization. The particle concentration distribution also improved significantly, which was good for improving the contact efficiency between the sorbent particles and SO{sub 2}. In addition, the optimization guidelines were proposed to further increase the desulfurization efficiency and the sorbent calcium conversion ratio.

  3. Determination of n, γ radiation field around the building of the swimming-pool reactor

    International Nuclear Information System (INIS)

    Jiang Jinling; Wen Youqin; Chen Changmao

    1986-01-01

    This work has measured the dose distribution of n, gamma radiation field around the building of the swimming-pool reactor by use of the highly sensitive neutron Rem counter and PTB-H 7907 exposure ratemeter. The measured datum show that the maximum value of n, gamma dose are 3-4 times greater than the background on certain distance from the building. Generally, the neutron doses are 2-3 times larger than gamma doses on most points

  4. Super-structure and building performance

    CSIR Research Space (South Africa)

    Van Wyk, Llewellyn V

    2010-11-01

    Full Text Available The super-structure consists predominantly of the load- and no-load-bearing walls-including all doors and windows and suspended floor slabs. The building envelope plays a significant role in the performance of a building, especially with regard...

  5. Experimental Breeder Reactor I Preservation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Julie Braun

    2006-10-01

    Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

  6. Piercing of the containment shell of a reactor building in case of airplane crash

    International Nuclear Information System (INIS)

    Herzog, M.

    1978-01-01

    The author presents a simple calculation model for a realistic check of the piercing safety of containments of reactor buildings in case of airplane crash. Its application is illustrated by a numerical example (Starfighter crash on the Unterweser nuclear power plant). (orig.) [de

  7. Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2 reactor building decontamination. Summary status report. Volume 2

    International Nuclear Information System (INIS)

    Doerge, D.H.; Miller, R.L.; Scotti, K.S.

    1986-05-01

    This document summarizes information relating to decontamination of the Three Mile Island Unit 2 (TMI-2) reactor building. The report covers activities for the period of June 1, 1979 through March 29, 1985. The data collected from activity reports, reactor containment entry records, and other sources were entered into a computerized data system which permits extraction/manipulation of specific information which can be used in planning for recovery from an accident similar to that experienced at TMI-2 on March 28, 1979. This report contains summaries of man-hours, manpower, and radiation exposures incurred during decontamination of the reactor building. Support activities conducted outside of radiation areas are excluded from the scope of this report. Computerized reports included in this document are: a chronological summary listing work performed relating to reactor building decontamination for the period specified; and summary reports for each major task during the period. Each task summary is listed in chronological order for zone entry and subtotaled for the number of personnel entries, exposures, and man-hours. Manually-assembled table summaries are included for: labor and exposures by department and labor and exposures by major activity

  8. Aging Management Plan for a Typical Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ebrahimi, Mahsa; Nazififard, Mohammad; Suh, Kune Y. [Seoul National University, Seoul (Korea, Republic of)

    2012-05-15

    Development of an aging management plan (AMP) is a crucial contributor to maintaining the reactor safety and controlling the risk of degradation of the concrete reactor building of a nuclear power plant. The design, operation and utilization of a research reactor (RR) fundamentally differ from those of power reactors. The AMP should nonetheless be present on account of radioactive materials and radiation risks involved. This is mainly because the RR is deemed to be used as an experiment itself or to conduct separate experiments during its operation. The AMP aims to determine the requisites for specific structural concrete components of the reactor building that entail regular inspections and maintenance to ensure safe and reliable operation of the plant. The safety of a RR necessitates the provision which is made in its design to facilitate aging management. Aging management of RR's structures is one of the vital factors to safety, to ensure continued adequacy of the safety level, reliable operation of the reactor, and compliance with the operational limits and conditions.Moreover, engineering systems should be qualified to meet the functional requirements for which they were designed with aging and environmental conditions for all situations and at all times taken into account. This study aims to present an integrated methodology for the application of an AMP for the concrete of the reactor building of a typical RR. For the purpose of safety analysis, geometry and ambient conditions were taken from a 5 MW pool-type, light-water moderated, heterogeneous, solid fuel RR in which the water is also used for cooling and shielding (Fig. 1). The reactor core is immersed in either section of a two-section concrete pool filled with water. This paper makes available background information regarding the document and the strategy developed to manage potential degradation of the reactor building concrete as well as specific programs and preventive and corrective

  9. MANHATTAN PROJECT B REACTOR HANFORD WASHINGTON [HANFORD'S HISTORIC B REACTOR (12-PAGE BOOKLET)

    Energy Technology Data Exchange (ETDEWEB)

    GERBER MS

    2009-04-28

    The Hanford Site began as part of the United States Manhattan Project to research, test and build atomic weapons during World War II. The original 670-square mile Hanford Site, then known as the Hanford Engineer Works, was the last of three top-secret sites constructed in order to produce enriched uranium and plutonium for the world's first nuclear weapons. B Reactor, located about 45 miles northwest of Richland, Washington, is the world's first full-scale nuclear reactor. Not only was B Reactor a first-of-a-kind engineering structure, it was built and fully functional in just 11 months. Eventually, the shoreline of the Columbia River in southeastern Washington State held nine nuclear reactors at the height of Hanford's nuclear defense production during the Cold War era. The B Reactor was shut down in 1968. During the 1980's, the U.S. Department of Energy began removing B Reactor's support facilities. The reactor building, the river pumphouse and the reactor stack are the only facilities that remain. Today, the U.S. Department of Energy (DOE) Richland Operations Office offers escorted public access to B Reactor along a designated tour route. The National Park Service (NPS) is studying preservation and interpretation options for sites associated with the Manhattan Project. A draft is expected in summer 2009. A final report will recommend whether the B Reactor, along with other Manhattan Project facilities, should be preserved, and if so, what roles the DOE, the NPS and community partners will play in preservation and public education. In August 2008, the DOE announced plans to open B Reactor for additional public tours. Potential hazards still exist within the building. However, the approved tour route is safe for visitors and workers. DOE may open additional areas once it can assure public safety by mitigating hazards.

  10. Structural building screening and evaluation

    Science.gov (United States)

    Kurniawandy, Alex; Nakazawa, Shoji; Hendry, Andy; Ridwan, Firdaus, Rahmatul

    2017-10-01

    An earthquake is a disaster that can be harmful to the community, such as financial loss and also dead injuries. Pekanbaru is a city that located in the middle of Sumatera Island. Even though the city of Pekanbaru is a city that rarely occurs earthquake, but Pekanbaru has ever felt the impact of the big earthquake that occurred in West Sumatera on September 2009. As we know, Indonesia located between Eurasia plate, Pacific plate, and Indo-Australian plate. Particularly the Sumatera Island, It has the Semangko fault or the great Sumatra fault along the island from north to south due to the shift of Eurasia and Indo-Australian Plates. An earthquake is not killing people but the building around the people that could be killing them. The failure of the building can be early prevented by doing an evaluation. In this research, the methods of evaluation have used a guideline for the Federal Emergency Management Agency (FEMA) P-154 and Applied Technology Council (ATC) 40. FEMA P-154 is a rapid visual screening of buildings for potential seismic hazards and ATC-40 is seismic evaluation and retrofit of Concrete Buildings. ATC-40 is a more complex evaluation rather than FEMA P-154. The samples to be evaluated are taken in the surroundings of Universitas Riau facility in Pekanbaru. There are four buildings as case study such as the rent student building, the building of mathematics and natural science faculty, the building teacher training and education faculty and the buildings in the faculty of Social political sciences. Vulnerability for every building facing an earthquake is different, this is depending on structural and non-structural components of the building. Among all of the samples, only the building of mathematics and the natural science faculty is in critical condition according to the FEMA P-154 evaluation. Furthermore, the results of evaluation using ATC-40 for the teacher training building are in damage control conditions, despite the other three buildings are

  11. Seismic strengthening of the ILL High Flux Reactor building

    International Nuclear Information System (INIS)

    Germane, Lionel; Plewinski, Francois; Thiry, Jean-Michel

    2006-01-01

    The Institut Max von Laue - Paul Langevin is an international research organisation and world leader in neutron science and technology. Since 1971 it has been operating the ILL HFR (High-Flux Reactor), the most intense continuous neutron source in the world. The ILL is governed by an international cooperation agreement between France, Germany and the United Kingdom; the fourth ten-year extension to the agreement was signed at the end of 2002, thus ensuring that the Institute will continue to operate until at least the end of 2013. In 2002 the facility underwent a general safety review, including an assessment of the impact of a safe shutdown earthquake. A broader programme for upgrading the installations and improving safety levels is now under way. As this has been treated in another paper, we will focus here on the seismic study carried out on the reactor building. The paper has the following contents: 1. Context; 1.1. Presentation of the ILL; 1.2. Description of the installations; 1.3. Safety objectives in the event of an earthquake; 1.4. Safety functions to be guaranteed in the event of an earthquake; 1.5. Safety functions required of the building; 2. Description of the building; 3. Organisation of the project; 3.1. Background; 3.2. Organisation; 4. General Methodology of the studies; 5. Progress of the studies; 5.1. Definition of the strengthening measures; 5.2. Validation of the strengthening option; 6. Seismic strengthening of the building; 6.1. Description of the strengthening measures; 6.2. Implementation of the strengthening measures; 6.2.1. Pilot operation; 6.2.2. Main operation; 7. Conclusion. To summarize, the presence of specialists in the ILL team, and the fact that the initial studies were performed by the project team itself, improved our general understanding of the issues and facilitated dialogue and exchange between all those involved (operators, technicians, outside experts, technical contractors and the French safety authorities). Everyone was

  12. Catalyst support structure, catalyst including the structure, reactor including a catalyst, and methods of forming same

    Science.gov (United States)

    Van Norman, Staci A.; Aston, Victoria J.; Weimer, Alan W.

    2017-05-09

    Structures, catalysts, and reactors suitable for use for a variety of applications, including gas-to-liquid and coal-to-liquid processes and methods of forming the structures, catalysts, and reactors are disclosed. The catalyst material can be deposited onto an inner wall of a microtubular reactor and/or onto porous tungsten support structures using atomic layer deposition techniques.

  13. Iterative model-building, structure refinement, and density modification with the PHENIX AutoBuild Wizard

    Energy Technology Data Exchange (ETDEWEB)

    Los Alamos National Laboratory, Mailstop M888, Los Alamos, NM 87545, USA; Lawrence Berkeley National Laboratory, One Cyclotron Road, Building 64R0121, Berkeley, CA 94720, USA; Department of Haematology, University of Cambridge, Cambridge CB2 0XY, England; Terwilliger, Thomas; Terwilliger, T.C.; Grosse-Kunstleve, Ralf Wilhelm; Afonine, P.V.; Moriarty, N.W.; Zwart, P.H.; Hung, L.-W.; Read, R.J.; Adams, P.D.

    2007-04-29

    The PHENIX AutoBuild Wizard is a highly automated tool for iterative model-building, structure refinement and density modification using RESOLVE or TEXTAL model-building, RESOLVE statistical density modification, and phenix.refine structure refinement. Recent advances in the AutoBuild Wizard and phenix.refine include automated detection and application of NCS from models as they are built, extensive model completion algorithms, and automated solvent molecule picking. Model completion algorithms in the AutoBuild Wizard include loop-building, crossovers between chains in different models of a structure, and side-chain optimization. The AutoBuild Wizard has been applied to a set of 48 structures at resolutions ranging from 1.1 {angstrom} to 3.2 {angstrom}, resulting in a mean R-factor of 0.24 and a mean free R factor of 0.29. The R-factor of the final model is dependent on the quality of the starting electron density, and relatively independent of resolution.

  14. Iterative model building, structure refinement and density modification with the PHENIX AutoBuild wizard

    International Nuclear Information System (INIS)

    Terwilliger, Thomas C.; Grosse-Kunstleve, Ralf W.; Afonine, Pavel V.; Moriarty, Nigel W.; Zwart, Peter H.; Hung, Li-Wei; Read, Randy J.; Adams, Paul D.

    2008-01-01

    The highly automated PHENIX AutoBuild wizard is described. The procedure can be applied equally well to phases derived from isomorphous/anomalous and molecular-replacement methods. The PHENIX AutoBuild wizard is a highly automated tool for iterative model building, structure refinement and density modification using RESOLVE model building, RESOLVE statistical density modification and phenix.refine structure refinement. Recent advances in the AutoBuild wizard and phenix.refine include automated detection and application of NCS from models as they are built, extensive model-completion algorithms and automated solvent-molecule picking. Model-completion algorithms in the AutoBuild wizard include loop building, crossovers between chains in different models of a structure and side-chain optimization. The AutoBuild wizard has been applied to a set of 48 structures at resolutions ranging from 1.1 to 3.2 Å, resulting in a mean R factor of 0.24 and a mean free R factor of 0.29. The R factor of the final model is dependent on the quality of the starting electron density and is relatively independent of resolution

  15. Assessment of thermal damage to polymeric materials by hydrogen deflagration in the Three Mile Island Unit 2 Reactor Building

    International Nuclear Information System (INIS)

    Alvares, N.J.

    1985-05-01

    Thermal damage to susceptible material in accessible regions of the reactor building was distributed in non-uniform patterns. No clear explanation for non-uniformity was found in examined evidence, e.g., burned materials were adjacent to materials that appear similar but were not burned. Because these items were in proximity to vertical openings that extend the height of the reactor building, we assume the unburned materials preferentially absorbed water vapor during periods of high, local steam concentration. Simple hydrogen-fire-exposure tests and heat transfer calculations duplicate the degree of damage found on inspected materials from the containment building. These data support estimated 8% pre-fire hydrogen concentration predictions based on various hydrogen production mechanisms

  16. STRUCTURAL VULNERABILITY ASSESSMENT OF MASONRY BUILDINGS IN TURKEY

    OpenAIRE

    KORKMAZ, Kasım Armagan; CARHOGLU, Asuman Isıl

    2011-01-01

    Turkey is located in an active seismic zone. Mid to high rise R/C building and low rise masonry buildings are very common construction type in Turkey. In recent earthquakes, lots of existing buildings got damage including masonry buildings. Masonry building history in Turkey goes long years back. For sure, it is an important structure type for Turkey. Therefore, earthquake behavior and structural vulnerability of masonry buildings are crucial issues for Turkey as a earthquake prone country. I...

  17. Structural materials issues for the next generation fission reactors

    Science.gov (United States)

    Chant, I.; Murty, K. L.

    2010-09-01

    Generation-IV reactor design concepts envisioned thus far cater to a common goal of providing safer, longer lasting, proliferation-resistant, and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-W reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses, and extremely corrosive environments, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This article addresses the material requirements for these advanced fission reactor types, specifically addressing structural materials issues depending on the specific application areas.

  18. Earthquake risk assessment of building structures

    International Nuclear Information System (INIS)

    Ellingwood, Bruce R.

    2001-01-01

    During the past two decades, probabilistic risk analysis tools have been applied to assess the performance of new and existing building structural systems. Structural design and evaluation of buildings and other facilities with regard to their ability to withstand the effects of earthquakes requires special considerations that are not normally a part of such evaluations for other occupancy, service and environmental loads. This paper reviews some of these special considerations, specifically as they pertain to probability-based codified design and reliability-based condition assessment of existing buildings. Difficulties experienced in implementing probability-based limit states design criteria for earthquake are summarized. Comparisons of predicted and observed building damage highlight the limitations of using current deterministic approaches for post-earthquake building condition assessment. The importance of inherent randomness and modeling uncertainty in forecasting building performance is examined through a building fragility assessment of a steel frame with welded connections that was damaged during the Northridge Earthquake of 1994. The prospects for future improvements in earthquake-resistant design procedures based on a more rational probability-based treatment of uncertainty are examined

  19. Vibration test on KMRR reactor structure and primary cooling system piping

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

  20. Vibration test on KMRR reactor structure and primary cooling system piping

    International Nuclear Information System (INIS)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author)

  1. Conceptual design of nuclear fusion power reactor DREAM. Reactor structures and remote maintenance

    International Nuclear Information System (INIS)

    Nishio, Satoshi; Seki, Yasushi; Ueda, Shuzo; Kurihara, Ryoichi; Adachi, Junichi; Yamazaki, Seiichiro; Hashimoto, Toshiyuki.

    1997-01-01

    Nuclear fusion reactors are required to be able to compete another energy sources in economy, reliability, safety and environmental integrity for commercial use. In the DREAM (DRastically EAsy Maintenance) reactor, a very low activated material of SiC/SiC composite has been introduced for the structural material, a reactor configuration for very easy maintenance and the helium gas of a high temperature for the cooling system, and hence DREAM has been proven to be very attractively as the commercial power reactor due to the high availability and efficiency of the plant and minimization of radioactive wastes. (author)

  2. STARFIRE remote maintenance and reactor facility concept

    International Nuclear Information System (INIS)

    Graumann, D.W.; Field, R.E.; Lutz, G.R.; Trachsel, C.A.

    1981-01-01

    A total remote maintenance facility has been designed for all equipment located within the reactor building and hot cell, although operational flexibility has been provided by design of the reactor shielding such that personnel access into the reactor building within 24 hours after reactor shutdown is possible. The reactor design permits removal and replacement of all components if necessary, however, the vacuum pumps, isolation valves and blanket require scheduled, routine maintenance. Reactor scheduled maintenance does not dominate annual plant downtime, therefore, several scheduled operations can be added without affecting reactor availability. The maintenance facilities consist of the reactor building, the hot cell, the reactor service area and the remote maintenance control room. The reactor building contains the reactor, selected support system modules, and required maintenance equipment. The reactor and the support systems are maintained with (1) equipment that is mounted on a monorail system; (2) overhead cranes; and (3) bridge-mounted electromechanical manipulators. The hot cell is located outside of the reactor building to localize contamination products and permit independent operation. An equipment air lock connects the reactor building to the hot cell

  3. Structure of thermonuclear reactor wall

    International Nuclear Information System (INIS)

    Yamazaki, Seiichiro.

    1991-01-01

    In a thermonuclear reactor wall, there has been a worry that the brazing material is melted by high temperature heat and particle load, to peel off the joined portion and the protecting material is destroyed by temperature elevation, to expose the heat sink material. Then, in the reactor core structures of a thermonuclear reactor, such as a divertor plate comprising a protecting material made of carbon material and the heat sink material joined by brazing, a plate material made of a so-called refractory metal having a high atomic number such as tungsten, molybdenum or the alloy thereof is embedded or attached to an accurate position of the protecting material. This can prevent the brazing portion from destruction by escaping electrons generated upon occurrence of abnormality in the thermonuclear reactor, and peeling or destroy of the protecting material and the heat sink material. Sufficient characteristics of plasmas can always be maintained by disposing a material having a small atomic number, for example, carbon material, to the position facing to the plasmas. (N.H.)

  4. Space Fission Reactor Structural Materials: Choices Past, Present and Future

    International Nuclear Information System (INIS)

    Busby, Jeremy T.; Leonard, Keith J.

    2007-01-01

    Nuclear powered spacecraft will enable missions well beyond the capabilities of current chemical, radioisotope thermal generator and solar technologies. The use of fission reactors for space applications has been considered for over 50 years, although, structural material performance has often limited the potential performance of space reactors. Space fission reactors are an extremely harsh environment for structural materials with high temperatures, high neutron fields, potential contact with liquid metals, and the need for up to 15-20 year reliability with no inspection or preventative maintenance. Many different materials have been proposed as structural materials. While all materials meet many of the requirements for space reactor service, none satisfy all of them. However, continued development and testing may resolve these issues and provide qualified materials for space fission reactors.

  5. Reactor container structure

    International Nuclear Information System (INIS)

    Sato, Yoshimi; Fukuda, Yoshio.

    1993-01-01

    A main container of an FBR type reactor using liquid sodium as coolants is attached to a roof slug. The main container contains, as coolants, lower temperature sodium, and high temperature sodium above a reactor core and a partitioning plate. The main container has a structure comprising only longitudinal welded joints in parallel with axial direction in the vicinity of the liquid surface of high temperature sodium where a temperature gradient is steep and great thermal stresses are caused without disposing lateral welded joints in perpendicular to axial direction. Only the longitudinal welded joints having a great fatigue strength are thus disposed in the vicinity of the liquid surface of the high temperature sodium where axial thermal stresses are caused. This can improve reliability of strength at the welded portions of the main container against repeating thermal stresses caused in vicinity of the liquid surface of the main container from a view point of welding method. (I.N.)

  6. Pressure loadings of Soviet-designed VVER [Water-Cooled, Water-Moderated Energy Reactor] reactor release mitigation structures from large-break LOCAs

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Horak, W.C.

    1989-01-01

    Analyses have been carried out of the pressurization of the accident release mitigation structures of Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactor) pressurized water reactors following large-break loss-of-coolant accidents. Specific VVER systems for which calculations were performed are the VVER-440 model V230, VVER-440 model V213, and VVER-1000 model V320. Descriptions of the designs of these and other VVER models are contained in the report DOE/NE-0084. The principal objective of the current analyses is to calculate the time dependent pressure loadings inside the accident localization or containment structures immediately following the double-ended guillotine rupture of a primary coolant pipe. In addition, the pressures are compared with the results of calculations of the response of the structures to overpressure. Primary coolant system thermal hydraulic conditions and the fluid conditions at the break location were calculated with the RETRAN-02 Mod2 computer code (Agee, 1984). Pressures and temperatures inside the building accident release mitigation structures were obtained from the PACER (Pressurization Accompanying Coolant Escape from Ruptures) multicompartment containment analysis code developed at Argonne National Laboratory. The analyses were carried out using best estimate models and conditions rather than conservative, bounding-type assumptions. In particular, condensation upon structure and equipment was calculated using correlations based upon analyses of the HDR, Marviken, and Battelle Frankfurt containment loading experiments. The intercompartment flow rates incorporate an effective discharge coefficient and liquid droplet carryover fraction given by expressions of Schwan determined from analyses of the Battelle Frankfurt and Marviken tests. 5 refs., 4 figs

  7. Reactor structure and superconducting magnet system of ITER

    International Nuclear Information System (INIS)

    Tada, Eisuke; Yoshida, Kiyoshi; Shibanuma, Kiyoshi; Okuno, Kiyoshi; Tsuji, Hiroshi; Shimamoto, Susumu

    1993-01-01

    Fusion Experimental Reactors are one of the major steps toward realization of the fusion energy and the key objective are to demonstrate the scientific and technological feasibility prior to the Demo Fusion Reactor. ITER (International Thermonuclear Experimental Reactor) is one of experimental reactors and the conceptual design has been completed by the united efforts of USA, USSR, EC and Japan. In parallel with the conceptual design, key technology development in various areas has being conducted. This paper describes the overall design concepts and the latest technological achievements of the ITER reactor structure and superconducting magnet system. (author)

  8. Intelligent seismic risk mitigation system on structure building

    Science.gov (United States)

    Suryanita, R.; Maizir, H.; Yuniorto, E.; Jingga, H.

    2018-01-01

    Indonesia located on the Pacific Ring of Fire, is one of the highest-risk seismic zone in the world. The strong ground motion might cause catastrophic collapse of the building which leads to casualties and property damages. Therefore, it is imperative to properly design the structural response of building against seismic hazard. Seismic-resistant building design process requires structural analysis to be performed to obtain the necessary building responses. However, the structural analysis could be very difficult and time consuming. This study aims to predict the structural response includes displacement, velocity, and acceleration of multi-storey building with the fixed floor plan using Artificial Neural Network (ANN) method based on the 2010 Indonesian seismic hazard map. By varying the building height, soil condition, and seismic location in 47 cities in Indonesia, 6345 data sets were obtained and fed into the ANN model for the learning process. The trained ANN can predict the displacement, velocity, and acceleration responses with up to 96% of predicted rate. The trained ANN architecture and weight factors were later used to build a simple tool in Visual Basic program which possesses the features for prediction of structural response as mentioned previously.

  9. Seismic isolation structure for pool-type LMFBR - isolation building with vertically isolated floor for NSSS

    International Nuclear Information System (INIS)

    Sakurai, A.; Shiojiri, H.; Aoyagi, S.; Matsuda, T.; Fujimoto, S.; Sasaki, Y.; Hirayama, H.

    1987-01-01

    The NSSS isolation floor vibration characteristics were made clear. Especially, the side support bearing (rubber bearing) is effective for horizontal floor motion restraint and rocking motion control. Seismic isolation effects for responses of the reactor components can be sufficiently expected, using the vertical seismic isolation floor. From the analytical and experimental studies, the following has been concluded: (1) Seismic isolation structure, which is suitable for large pool-type LMFBR, were proposed. (2) Seismic response characteristics of the seismic isolation structure were investigated. It was made clear that the proposed seismic isolation (Combination of the isolated building and the isolated NSSS floor) was effective. (orig./HP)

  10. Structural integrity aspects of reactor safety

    Indian Academy of Sciences (India)

    A large experimental programme supported the structural integrity demonstration. ... Categories in which the structures, systems and components (SSC) are .... One of the ways in which the decision to live with the defect can be aided is the .... The Advanced Heavy Water Reactor (AHWR) (figure 18) being designed by BARC ...

  11. The inner containment of an EPR trademark pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ostermann, Dirk; Krumb, Christian; Wienand, Burkhard [AREVA GmbH, Offenbach (Germany)

    2014-08-15

    On February 12, 2014 the containment pressure and subsequent leak tightness tests on the containment of the Finnish Olkiluoto 3 EPR trademark reactor building were completed successfully. The containment of an EPR trademark pressurized water reactor consists of an outer containment to protect the reactor building against external hazards (such as airplane crash) and of an inner containment that is subjected to internal overpressure and high temperature in case of internal accidents. The current paper gives an overview of the containment structure, the design criteria, the validation by analyses and experiments and the containment pressure test.

  12. Structural Integrity Evaluation of the KALIMER-600 Reactor Core Support Structure

    International Nuclear Information System (INIS)

    Park, Chang Gyu; Kim, Jong Bum; Lee, Jae Han

    2005-01-01

    KALIMER-600(Korea Advanced LIquid MEtal Reactor, 600MWe) is a pool type sodium-cooled liquid metal reactor. Since the normal operating temperature of KALIMER-600 is 545 .deg. C, the reactor structures in the hot pool region are designed and evaluated according to the elevated temperature design rules such as the ASME Boiler and Pressure Vessel Code Section III, Subsection NH. Since the core support structure of KALIMER-600 is in the cold pool region under 400 .deg. C, a high temperature inelastic behavior is not expected. Thus the stress and fatigue limits are the main concerns to assure the structural design integrity following the ASME Subsection NG. In this paper, the evaluations of the stress and fatigue damage for the core support structure of KALIMER-600 are carrried out in the case of a normal operation condition using the rules of ASME Subsection NG. To obtain the stress values, a heat transfer analysis and a stress analysis under a combined loading condition are performed. From the stress distribution results, the critical sections are selected and the stress and fatigue limits are evaluated for the selected regions

  13. Measurement methods of building structures deflections

    Directory of Open Access Journals (Sweden)

    Wróblewska Magdalena

    2018-01-01

    Full Text Available Underground mining exploitation is leading to the occurrence of deformations manifested by, in particular, sloping terrain. The structures situated on the deforming subsoil are subject to uneven subsidence which is leading in consequence to their deflection. Before a building rectification process takes place by, e.g. uneven raising, the structure's deflection direction and value is determined so that the structure is restored to its vertical position as a result of the undertaken remedial measures. Deflection can be determined by applying classical as well as modern measurement techniques. The article presents examples of measurement methods used considering the measured elements of building structures’ constructions and field measurements. Moreover, for a given example of a mining area, the existing deflections of buildings were compared with mining terrain sloping.

  14. Geological and geotechnical aspects of the foundation pit of Kaiga atomic power plant reactor building 2, Kaiga, Uttara Kannada district, Karnataka

    International Nuclear Information System (INIS)

    Katti, Vinod J.; Shah, V.L.; Pande, A.K.

    2014-01-01

    In India Nuclear Power Plants are constructed as per the guidelines laid by IAEA and AERB. Before concrete is poured into reactor building pits, they are systematically mapped and Iithostructural maps are prepared for pit base and side walls. The constraints noticed are carefully attended with geotechnical solutions and remedies to make foundation safe for the entire period of reactor life. Similarly, pit of Kaiga Reactor Building II was systematically mapped for circular base and side walls. Geo-engineering solutions like scrapping out loose, foliated schistose patches, scooping out soft altered zones, filling with grouting, rock-bolting rock segments with major joints and fractures for stopping seepage points were suggested. (author)

  15. Protective guide structure for reactor control rod

    International Nuclear Information System (INIS)

    Ban, Minoru; Umeda, Kenji; Kubo, Noboru; Ito, Tomohiro.

    1996-01-01

    The present invention provides an improved protective guide structure for control rods, which does not cause swirling of coolants and resonance even though a slit is formed on a protective tube which surrounds a control rod element in a PWR type reactor. Namely, a reactor control rod is constituted with elongated control elements collectively bundled in the form of a cluster. The protective guide structure protectively guides the collected constituent at the upper portion of a reactor container. The protective structure comprises a plurality of protective tubes each having a C-shaped cross section disposed in parallel for receiving control rod elements individually in which the corners of the opening of the cross section of the protective tube are chamfered to an appropriate configuration. With such a constitution, even if coolant flows in a circumferential direction along the protective tubes surrounding the control rod elements, no shearing stream is caused to the coolants flow since the corners of the cross sectional opening (slit) of the tube are chamfered. Accordingly, occurrence of swirlings can be suppressed. (I.S.)

  16. Monitoring actual temperatures in Susquehanna SES reactor buildings

    International Nuclear Information System (INIS)

    Derkacs, A.P.

    1991-01-01

    PP and L has been monitoring temperatures in the Susquehanna SES reactor building with digital temperature recorders since 1986. In early 1990, data from four representative areas was analyzed to determine the temperature in each area which would produce the same rate of degradation as the distribution of actual temperatures recorded over about 40 months. From these effective average temperatures, qualified life multipliers were determined for activation energies in the range of 0.5 to 1.5 and those multipliers were used to estimate new qualified lives and the number of replacements which might be saved during the life of the plant. The results indicate that pursuing a program of determining EQ qualified lives from actual temperatures, rather than maximum design basis temperatures, will provide a substantial payback in reduced EQ driven maintenance

  17. Dynamic containment of gaseous effluents in the auxiliary buildings and reinjection of liquid effluents from these buildings back into the reactor building for 900 MWe PWRs under accident condition

    International Nuclear Information System (INIS)

    Demoulin, F.; Collinet, J.; Nguyen, C.

    1987-04-01

    Examination of the lessons to be learned from the accident of the Three Mile Island nuclear power plant on 20 March 1979 led the French Safety Authorities and EDF (Electricite de France) to adopt a series of measures intended to improve the performance of the containment of French PWRs, especially in the event of accident. Among the measures adopted, two of them contribute to the upgrading of the containment of nuclear island buildings, by reducing radioactivity constraints inside these buildings and by limiting radioactive releases into the environment. These are: (1) dynamic containment of auxiliary buildings likely to be contaminated following an accident, (2) reinjection back into the reactor building of liquid effluents arising in the auxiliary buildings. In this paper we shall discuss, for each measure, the approach to the problem and describe the arrangements made to arrive at a satisfactory solution [fr

  18. Mobile means for the monitoring of atmospheric contamination in a reactor building

    International Nuclear Information System (INIS)

    Marques, S.; Lestang, M.

    2009-01-01

    After having evoked the context and challenges of contamination monitoring when exploiting nuclear reactors, the authors discuss the representativeness of the atmospheric contamination measurement as it depends on the different physicochemical forms of radionuclides present in the circuits. They indicate the different gaseous or aerosol radioactive elements which are monitored within EDF installations. They discuss the incorporation of monitoring means at the installation design level, briefly present the use of beacons inside and outside the reactor building. They describe how monitoring is organized on the basis of alert threshold adjustments: an investigation threshold and an evacuation threshold. They discuss the beacon (or sensor) selection and indicate recommendations for their implementation for optimization purposes. They indicate where these beacons are installed and evoke the experimentation of networked mobile beacons with data remote transmission

  19. Experimental building with new types of building envelope structures. Part 1: Structures/systems. Building system: Brick walls; Forsoegshus med nye typer klimaskaermskonstruktioner. Del 1: Konstruktioner/systemer - Byggesystem: Fuldmuret

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-07-01

    The house described in this report is one of several experimental houses forming part of the project 'Experimental buildings with new types of building envelope structures'. One purpose of the project is to demonstrate that it is possible to build typical single-family houses with an energy consumption that meets expected increased building regulations. Furthermore, it is important that the houses can be made securely as regards construction technology and within reasonable financial limits. Thus, the purpose is also to contribute to strengthen the development of improved building envelope structures. Another purpose is to carry out detailed measurements of energy consumption in order to validate thermal performance of future building envelope structures. The report describes the constructive design and energy systems of the house plus heat loss calculations and expected energy consumption. (BA)

  20. Control technologies for quadruped walking robot to facilitate carrying operations in reactor buildings

    International Nuclear Information System (INIS)

    Suganuma, Naotaka; Uehara, Takuya; Nakamura, Norihito

    2014-01-01

    At the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Co., Inc., which was seriously damaged by the Great East Japan Earthquake of March 11, 2011, it has been difficult for workers to approach the reactor buildings due to the hazardous surrounding environment. The need has therefore arsen for remote-controlled robots to facilitate inspection and restoration work on behalf of workers in such a high-level radiation environment. Toshiba has developed a quadruped walking robot that can carry various tools for decommissioning work. This robot is capable of maintaining its balance while walking on uneven surfaces, slopes, and stairs due to the adoption of control technologies to not only autonomously determine the leg trajectories and center of gravity, but also to correct the leg landing positions and posture with operator intervention according to the walking situation. It also offers high mobility and workability through a manipulation function that allows it to unload tools carried on its back storage area by using two of its legs like arms. This quadruped walking robot was applied to the investigation of suspected water leakage areas in the reactor building of Fukushima Daiichi Nuclear Power Station Unit 2 in December 2012. (author)

  1. Environmental effect of structural solutions and building materials to a building

    International Nuclear Information System (INIS)

    Haapio, Appu; Viitaniemi, Pertti

    2008-01-01

    The field of building environmental assessment tools has become a popular research area over the past decade. However, how the service life of a building affects the results of the environmental assessment of a building has not been emphasised previously. The aim of this study is to analyse how different structural solutions and building materials affect the results of the environmental assessment of a whole building over the building's life cycle. Furthermore, how the length of the building's service life affects the results is analysed. The environmental assessments of 78 single-family houses were calculated for this study. The buildings have different wall insulations, claddings, window frames, and roof materials, and the length of the service life varies from 60 years up to 160 years. The current situation and the future of the environmental assessment of buildings are discussed. In addition, topics for further research are suggested; for example, how workmanship affects the service life and the environmental impact of a building should be studied

  2. TA-2 Water Boiler Reactor Decommissioning Project

    International Nuclear Information System (INIS)

    Durbin, M.E.; Montoya, G.M.

    1991-06-01

    This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m 3 of low-level solid radioactive waste and 35 m 3 of mixed waste. 15 refs., 25 figs., 3 tabs

  3. Structural rehabilitation of old buildings

    CERN Document Server

    Guedes, João; Varum, Humberto

    2014-01-01

    The present book describes the different construction systems and structural materials and solutions within the main old buildings typologies, and it analyses the particularities of each of them, including mechanical properties, structural behaviour, typical damage patterns and collapse mechanisms. Common or pioneering intervention measures to repair and/or strengthen some of these structural elements are also reviewed.

  4. European R and D co-ordinate programme on structural integrity of fast breeder reactors

    International Nuclear Information System (INIS)

    Hoffmann, A.; Combescure, A.; Acker, D.; Corsi, F.; Martelli, A.; Vinzens, K.; Angerbauer, A.

    1989-01-01

    After a period of development of medium size prototype plants, PFR in United Kingdom, Phenix in France and SNR 300 in West Germany, and the build up of the large size prototype plant Super Phenix in France by the European Consortium, NERSA, Belgium, France, Italy, United Kingdom and West Germany decided to join their efforts in order to pursue the development of LMFBR nuclear power plants. This paper presents the European Research and Development coordinated program in the field of structural integrity of fast breeder reactors with its organization, its objectives, its programs and the resources allocated for development. (author)

  5. Proceedings of 18th international conference on structural mechanics in reactor technology

    International Nuclear Information System (INIS)

    2005-07-01

    The 18th International Conference on Structural Mechanics in Reactor Technology was held on August 7-12, 2005 in Beijing, China, and Sponsored by International Association for Structural Mechanics in Reactor Technology, Chinese Nuclear Society, Chinese Society of Theoretical and Applied Mechanics, and Tsinghua University. 486 abstracts are Collected. The contents includes: opening, plenary and keynote presentations; computational mechanics; fuel and core structures; aging, life extension, and license renewal; design methods and rules for components; fracture mechanics; concrete material, containment and other structures; analysis and design for dynamic and extreme loads; seismic analysis, design and qualification; structural reliability and probabilistic safety assessment (PSA); operation, inspection and maintenance; severe accident management and structural evaluation; advanced reactors and generation IV reactors; decommissioning of nuclear facilities and waste management.

  6. Proposal for the use of new materials in the TOKAMAK building cover

    International Nuclear Information System (INIS)

    Chiva, L.

    2011-01-01

    It was considered relevant and innovative to apply new structural materials to the construction of the roof of the building that lodged the TOKAMAK reactor, with the aim of achieving a severe reduction of the weight of the roof structure that result in greater ease of mounting, minor charges on the walls and foundations of the building and a reduced impact on the distribution of masses of the building scheme.

  7. Continuous Monitoring of GAMMA Radiation Field in the Reactor RA Building

    International Nuclear Information System (INIS)

    Stalevski, T.

    2008-01-01

    This paper presents the system for continuos monitoring of gamma doze rate in the reactor RA building. Industrial (PC compatible) computer acquires analog signals from eight ionization chambers and eight analog signals from three BPH devices. Digital output interface is used for testing ionization chambers and BPH devices. Computer program for data analyzes and presentation is written in graphical programming language LabVIEW and enables monitoring of measured data in real time. Measured data can be monitored over local computer network, Internet and mobile devices using standard web browsers. (author)

  8. ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

    International Nuclear Information System (INIS)

    Blanford, E.; Keldrauk, E.; Laufer, M.; Mieler, M.; Wei, J.; Stojadinovic, B.; Peterson, P.F.

    2010-01-01

    Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement, and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using

  9. ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    E. Blanford; E. Keldrauk; M. Laufer; M. Mieler; J. Wei; B. Stojadinovic; P.F. Peterson

    2010-09-20

    Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement, and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using

  10. A study on the functional assessment of the prestressed system and main structural elements in life extended containment building

    Energy Technology Data Exchange (ETDEWEB)

    Cheong, C. H.; Kim, S. W.; Choi, J. G. [DAEWOO E and C Institute of Costruction Technology, Suwon (Korea, Republic of)] (and others)

    2001-10-15

    The design life of KNGR (Korean Next Generation Reactor) containment buildings is extended from 40 years to 60 years. However, nuclear reactor buildings are passive structures that are impossible to be exchanged in the case of degradation by the deterioration and so on when extending the design life of structures. Therefore, it is necessary to consider the long-term safety endurance in the design and construction of KNGR. Also, ti is judged that choice of the material and various test methods should be prescribed clearly. In this study, the reduction schemes of deterioration and the safety-ensuring schemes are drawn for the expected performance to be maintained from the beginning of the service to the required period together with ensuring the safety and serviceability of KNGR which will be constructed with the design life of 60 years, taking into account the dimensions, selection of material and construction methods in the design and construction stages. Also, the validity is to be examined for the estimation method of long-term losses of stress introduced to KNGR whose design life is increased to 60 years. The durability enhancement scheme on the design and construction for the design life extension of nuclear containment buildings is to be drawn through these studies. These results are utilized as the basic data for the safety inspection and examination guides of KNGR and finally the additional investigations are proposed for the items which require long-term studies.

  11. N Reactor Deactivation Program Plan

    International Nuclear Information System (INIS)

    Walsh, J.L.

    1993-12-01

    This N Reactor Deactivation Program Plan is structured to provide the basic methodology required to place N Reactor and supporting facilities · in a radiologically and environmentally safe condition such that they can be decommissioned at a later date. Deactivation will be in accordance with facility transfer criteria specified in Department of Energy (DOE) and Westinghouse Hanford Company (WHC) guidance. Transition activities primarily involve shutdown and isolation of operational systems and buildings, radiological/hazardous waste cleanup, N Fuel Basin stabilization and environmental stabilization of the facilities. The N Reactor Deactivation Program covers the period FY 1992 through FY 1997. The directive to cease N Reactor preservation and prepare for decommissioning was issued by DOE to WHC on September 20, 1991. The work year and budget data supporting the Work Breakdown Structure in this document are found in the Activity Data Sheets (ADS) and the Environmental Restoration Program Baseline, that are prepared annually

  12. Effects of soil stiffness and embedment on reactor building response

    International Nuclear Information System (INIS)

    Michalopoulos, A.P.; Vardanega, C.; Cornaggia, L.

    1981-01-01

    A parametric study was made to assess the influence of soil conditions and foundation embedment depth on the floor response spectra for a reactor building. The analyses incorporated soft, medium and hard soils, and three different embedment depths, in a seismic environment described by a 0.36 g peak ground acceleration. The shear wave velocity profiles for the soft, medium and hard soil conditions, were assumed to increase in proportion to the square root of depth from their ground surface values of 300, 600 and 900 meters per second, respectively. Foundation embedment depths of zero, eight and fourteen meters were analyzed using elastic half-space theory, accounting for kinematic interaction. The variation of shear modulus with depth under earthquake excitation was determined using a deconvolution process. Horizontal and vertical synthetic time histories, matching the USNRC Regulatory Guide 1.60 design ground response spectra, were applied at the ground surface and then deconvolved to the foundation level to obtain the input for the soil-structure model. The mathematical model of the superstructure consisted of four lumped-mass close-coupled systems, representing containment shells and components, while the foundation mat was modeled as rigid. Lumped soil compliances (springs and dashpots) were used to represent the horizontal, vertical and rotational modes of vibration. The dynamic analyses were performed utilizing the computer code DAPSYS, and consisted of mode frequency analyses and modal superposition. Modal damping was computed as a weighted average of structural and soil (radiation and material) damping, using the strain energy stored in the respective components as the weighting factor and distinguishing the hysteric nature of the structural and soil material damping, and the viscous nature of the soil radiation damping. (orig./RW)

  13. Multicriteria Analysis of Assembling Buildings from Steel Frame Structures

    Science.gov (United States)

    Miniotaite, Ruta

    2017-10-01

    Steel frame structures are often used in the construction of public and industrial buildings. They are used for: all types of slope roofs; walls of newly-built public and industrial buildings; load bearing structures; roofs of renovated buildings. The process of assembling buildings from steel frame structures should be analysed as an integrated process influenced by such factors as construction materials and machinery used, the qualification level of construction workers, complexity of work, available finance. It is necessary to find a rational technological design solution for assembling buildings from steel frame structures by conducting a multiple criteria analysis. The analysis provides a possibility to evaluate the engineering considerations and find unequivocal solutions. The rational alternative of a complex process of assembling buildings from steel frame structures was found through multiple criteria analysis and multiple criteria evaluation. In multiple criteria evaluation of technological solutions for assembling buildings from steel frame structures by pairwise comparison method the criteria by significance are distributed as follows: durability is the most important criterion in the evaluation of alternatives; the price (EUR/unit of measurement) of a part of assembly process; construction workers’ qualification level (category); mechanization level of a part of assembling process (%), and complexity of assembling work (in points) are less important criteria.

  14. Residual stress improving method for reactor structural component and residual stress improving device therefor

    Energy Technology Data Exchange (ETDEWEB)

    Enomoto, Kunio; Otaka, Masahiro; Kurosawa, Koichi; Saito, Hideyo; Tsujimura, Hiroshi; Tamai, Yasukata; Urashiro, Keiichi; Mochizuki, Masato

    1996-09-03

    The present invention is applied to a BWR type reactor, in which a high speed jetting flow incorporating cavities is collided against the surface of reactor structural components to form residual compression stresses on the surface layer of the reactor structural components thereby improving the stresses on the surface. Namely, a water jetting means is inserted into the reactor container filled with reactor water. Purified water is pressurized by a pump and introduced to the water jetting means. The purified water jetted from the water jetting means and entraining cavities is abutted against the surface of the reactor structural components. With such procedures, since the purified water is introduced to the water jetting means by the pump, the pump is free from contamination of radioactive materials. As a result, maintenance and inspection for the pump can be facilitated. Further, since the purified water injection flow entraining cavities is abutted against the surface of the reactor structural components being in contact with reactor water, residual compression stresses are exerted on the surface of the reactor structural components. As a result, occurrence of stress corrosion crackings of reactor structural components is suppressed. (I.S.)

  15. Residual stress improving method for reactor structural component and residual stress improving device therefor

    International Nuclear Information System (INIS)

    Enomoto, Kunio; Otaka, Masahiro; Kurosawa, Koichi; Saito, Hideyo; Tsujimura, Hiroshi; Tamai, Yasukata; Urashiro, Keiichi; Mochizuki, Masato.

    1996-01-01

    The present invention is applied to a BWR type reactor, in which a high speed jetting flow incorporating cavities is collided against the surface of reactor structural components to form residual compression stresses on the surface layer of the reactor structural components thereby improving the stresses on the surface. Namely, a water jetting means is inserted into the reactor container filled with reactor water. Purified water is pressurized by a pump and introduced to the water jetting means. The purified water jetted from the water jetting means and entraining cavities is abutted against the surface of the reactor structural components. With such procedures, since the purified water is introduced to the water jetting means by the pump, the pump is free from contamination of radioactive materials. As a result, maintenance and inspection for the pump can be facilitated. Further, since the purified water injection flow entraining cavities is abutted against the surface of the reactor structural components being in contact with reactor water, residual compression stresses are exerted on the surface of the reactor structural components. As a result, occurrence of stress corrosion crackings of reactor structural components is suppressed. (I.S.)

  16. Structural integrity evaluation of PWR nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Cruz, Julio R.B.; Mattar Neto, Miguel

    1999-01-01

    The reactor pressure vessel (RPV) is the most important structural component of a PWR nuclear power plant. It contains the reactor core and is the main component of the primary system pressure boundary, the system responsible for removing the heat generated by the nuclear reactions. It is considered not replaceable and, therefore, its lifetime is a key element to define the plant life as a whole. Three critical issues related to the reliability of the RPV structural integrity come out by reason of the radiation damage imposed to the vessel material during operation. These issues concern the definition of pressure versus temperature limits for reactor heatup and cooldown, pressurized thermal shock evaluation and assessment of reactor vessels with low upper shelf Charpy impact energy levels. This work aims to present the major aspects related to these topics. The requirements for preventing fracture of the RPV are reviewed as well as the available technology for assessing the safety margins. For each mentioned problem, the several steps for structural integrity evaluation are described and the analysis methods are discussed. (author)

  17. Improving the Earthquake Resilience of Buildings The worst case approach

    CERN Document Server

    Takewaki, Izuru; Fujita, Kohei

    2013-01-01

    Engineers are always interested in the worst-case scenario. One of the most important and challenging missions of structural engineers may be to narrow the range of unexpected incidents in building structural design. Redundancy, robustness and resilience play an important role in such circumstances. Improving the Earthquake Resilience of Buildings: The worst case approach discusses the importance of worst-scenario approach for improved earthquake resilience of buildings and nuclear reactor facilities. Improving the Earthquake Resilience of Buildings: The worst case approach consists of two parts. The first part deals with the characterization and modeling of worst or critical ground motions on inelastic structures and the related worst-case scenario in the structural design of ordinary simple building structures. The second part of the book focuses on investigating the worst-case scenario for passively controlled and base-isolated buildings. This allows for detailed consideration of a range of topics includin...

  18. Investigation of structural materials of reactors using high-energy heavy-ion irradiations

    International Nuclear Information System (INIS)

    Wang Zhiguang

    2007-01-01

    Radiation damage in structural materials of fission/fusion reactors is mainly attributed to the evolution of intensive atom displacement damage induced by energetic particles (n, α and/or fission fragments) and high-rate helium doping by direct α particle bombardments and/or (n, α) reactions. It can cause severe degradation of reactor structural materials such as surface blistering, bulk void swelling, deformation, fatigue, embrittlement, stress erosion corrosion and so on that will significantly affect the operation safety of reactors. However, up to now, behavior of structural materials at the end of their service can hardly be fully tested in a real reactor. In the present work, damage process in reactor structural materials is briefly introduced, then the advantages of energetic ion implantation/irradiation especially high-energy heavy ion irradiation are discussed, and several typical examples on simulation of radiation effects in reactor candidate structural materials using high-energy heavy ion irradiations are pronounced. Experimental results and theoretical analysis suggested that irradiation with energetic particles especially high-energy heavy ions is very useful technique for simulating the evolution of microstructures and macro-properties of reactor structural materials. Furthermore, an on-going plan of material irradiation experiments using high energy H- and He-ions based on the Heavy Ion Research Facilities in Lanzhou (HIRFL) is also briefly interpreted. (authors)

  19. Floor response spectra of buildings with uncertain structural properties

    International Nuclear Information System (INIS)

    Chen, P.C.

    1975-01-01

    All Category I equipment, such as reactors, vessels, and major piping systems of nuclear power plants, is required to withstand earthquake loadings in order to minimize risk of seismic damage. The equipment is designed by using response spectra of the floor on which the equipment is mounted. The floor response spectra are constructed usually from the floor response time histories which are obtained through a deterministic dynamic analysis. This analysis assumes that all structural parameters, such as mass, stiffness, and damping have been calculated precisely, and that the earthquakes are known. However, structural parameters are usually difficult to determine precisely if the structures are massive and/or irregular, such as nuclear containments and its internal structures with foundation soil incorporated into the analysis. Faced with these uncertainties, it has been the practice to broaden the floor response spectra peaks by +-10 percent of the peak frequencies on the basis of conservatism. This approach is based on engineering judgement and does not have an analytical basis to provide a sufficient level of confidence in using these spectra for equipment design. To insure reliable design, it is necessary to know structural response variations due to variations in structural properties. This consideration leads to the treatment of structural properties as random variables and the use of probabilistic methods to predict structural response more accurately. New results on floor response spectra of buildings with uncertain structural properties obtained by determining the probabilistic dynamic response from the deterministic dynamic response and its standard deviation are presented. The resulting probabilistic floor response spectra are compared with those obtained deterministically, and are shown to provide a more reliable method for determining seismic forces

  20. Common floor system vertical earthquake-proof structure for reactor equipment

    International Nuclear Information System (INIS)

    Morishita, Masaki.

    1996-01-01

    In an LMFBR type reactor, a reactor container, a recycling pump and a heat exchanger are disposed on a common floor. Vertical earthquake-proof devices which can be stretched only in vertical direction formed by laminating large-sized bellevilles are disposed on a concrete wall at the circumference of each of reactor equipments. A common floor is placed on all of the vertical earthquake-proof devices to support the entire earthquake-proof structure simultaneously. If each of reactor equipments is loaded on the common floor and the common floor is entirely supported against earthquakes altogether, since the movement of each of the reactor equipments loaded on the common floor is identical, relative dislocation is not exerted on the main pipelines which connect the equipments. In addition, since the entire earthquake structure has a flat common floor and each of the reactor equipments is suspended to minimize the distance between a gravitational center and a support point, locking vibration is less caused to the horizontal earthquake. (N.H.)

  1. A Review of Influence of Various Types of Structural Bracing to the Structural Performance of Buildings

    Science.gov (United States)

    Razak, S. M.; Kong, T. C.; Zainol, N. Z.; Adnan, A.; Azimi, M.

    2018-03-01

    Excessive lateral drift can contribute significantly towards crack formation, leading to structural damage. The structural damage will in turn reduce the capacity of the structure and weaken it from the intended design capacity. Generally, lateral drift is more pronounced in higher and longer structure, such as high rise buildings and bridges. A typical method employed to control lateral drift is structural bracing, which works by increasing stiffness and stability of structure. This paper reviews the influence of various types of structural bracing to structural performance of buildings. The history of structural bracing is visited and the differences between numerous structural bracing in term of suitability to different types of buildings and loading, mechanisms, technical details, advantages and limitations, and the overall effect on the structural behaviour and performance are dissected. Proper and efficient structural bracing is pertinent for each high rise building as this will lead towards safer, sustainable and more economical buildings, which are cheaper to maintain throughout the life of the buildings in the future.

  2. Prospect of Ti-Ni shape memory alloy applied in reactor structures

    International Nuclear Information System (INIS)

    Duan Yuangang

    1995-01-01

    Shape memory effect mechanism, physical property, composition, manufacturing process and application in mechanical structure of Ti-Ni shape memory alloy are introduced. Applications of Ti-Ni shape memory alloy in reactor structure are prospected and some necessary technical conditions of shape memory alloy applied in the reactor structure are put forward initially

  3. Jules Horowitz reactor (RJH): its design

    International Nuclear Information System (INIS)

    Dupuy, J.P.

    2002-01-01

    This article presents the design of the new irradiation facility (Jules Horowitz reactor) that is planned to be built on the Cadarache site of Cea. 2 principles have been followed. The first one is based on a physical separation between the systems and activities related to the reactor and the experiments from one hand and the other systems and means dedicated to the treatment of the experimental devices before and after irradiation on the other hand. This first principle implies to build 2 buildings: the reactor building and the nuclear auxiliaries building. Inside the reactor building activities from the reactor itself are separated from those dedicated to experimentation. In order to maximize the efficiency of such a reactor, an important number of simultaneous experiments is expected, which will generate an endless flux of incoming and out-going experiments and as a consequence an important handling work between the different work posts. The second principle aims at easing any handling work without breaking the rules of confinement. The different storing pools, the water pits that lead to the 5 hot cells and the reactor tank will communicate through a water-filled canal that will link the 2 buildings. (A.C.)

  4. An Evaluation Report on the High Temperature Design of the KALIMER-600 Reactor Structures

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Gyu; Lee, Jae Han

    2007-03-15

    This report is on the validity evaluation of high temperature structural design for the reactor structures and piping of the pool-type Liquid Metal Reactor, KALIMER-600 subjected to the high temperature thermal load condition. The structural concept of the Upper Internal Structure located above the core is analyzed and the adequate UIS conceptual design for KALIMER-600 is proposed. Also, the high temperature structural integrity of the thermal liner which is to protect the UIS bottom plate from the high frequency thermal fatigue damage was evaluated by the thermal stripping analysis. The high temperature structural design of the reactor internal structure by considering the reactor startup-shutdown cycle was carried out and the structural integrity of it for a normal operating condition as well as the transient condition of the primary pump trip accident was confirmed. Additionally the structure design of the reactor internal structural was changed to prevent the non-uniform deformation of the primary pump which is induced by the thermal expansion difference between the reactor head and the baffle plate. The arrangement of the IHTS piping system which is a part of the reactor system is carried out and the structural integrity and the accumulated deformation by considering the reactor startup-shutdown cycle of a normal operating condition were evaluated. The structural integrity and the accumulated deformation of the PDRC hot leg piping by considering the PDRC operating condition were evaluated. The validity of KALIMER-600 high temperature structural design is confirmed through this study, and it is clearly found that the methodology research to evaluate the structural integrity considering the reactor life time of 60 years ensured is necessary.

  5. Shield structure for a nuclear reactor

    International Nuclear Information System (INIS)

    Rouse, C.A.; Simnad, M.T.

    1979-01-01

    An improved nuclear reactor shield structure is described for use where there are significant amounts of fast neutron flux above an energy level of approximately 70 keV. The shield includes structural supports and neutron moderator and absorber systems. A portion at least of the neutron moderator material is magnesium oxide either alone or in combination with other moderator materials such as graphite and iron. (U.K.)

  6. Structural and piping issues in the design certification of advanced reactors

    International Nuclear Information System (INIS)

    Ali, S.A.; Terao, D.; Bagchi, G.

    1996-01-01

    The purpose of this paper is to discuss the design certification of structures and piping for evolutionary and passive advanced light water reactors. Advanced reactor designs are based on a set of assumed site-related parameters that are selected to envelop a majority of potential nuclear power plant sites. Multiple time histories are used as the seismic design basis in order to cover the majority of potential sites in the US. Additionally, design are established to ensure that surface motions at a particular site will not exceed the enveloped standard design surface motions. State-of-the-art soil-structure interaction (SSI) analyses have been performed for the advanced reactors, which include structure-to-structure interaction for all seismic Category 1 structures. Advanced technology has been utilized to exclude the dynamic effects of pipe rupture from structural design by demonstrating that the probability of pipe rupture is extremely low. For piping design, the advanced reactor vendors have developed design acceptance criteria (DAC) which provides the piping design analysis methods, design procedures, and acceptance criteria. In SECY-93-087 the NRC staff recommended that the Commission approve the approach to eliminate the OBE from the design of structures and piping in advanced reactors and provided guidance which identifies the necessary changes to existing seismic design criteria. The supplemental criteria address fatigue, seismic anchor motion, and piping stress limits when the OBE is eliminated

  7. Automated detection of repeated structures in building facades

    Directory of Open Access Journals (Sweden)

    M. Previtali

    2013-10-01

    Full Text Available Automatic identification of high-level repeated structures in 3D point clouds of building façades is crucial for applications like digitalization and building modelling. Indeed, in many architectural styles building façades are governed by arrangements of objects into repeated patterns. In particular, façades are generally designed as the repetition of some few basic objects organized into interlaced and\\or concatenated grid structures. Starting from this key observation, this paper presents an algorithm for Repeated Structure Detection (RSD in 3D point clouds of building façades. The presented methodology consists of three main phases. First, in the point cloud segmentation stage (i the building façade is decomposed into planar patches which are classified by means of some weak prior knowledge of urban buildings formulated in a classification tree. Secondly (ii, in the element clustering phase detected patches are grouped together by means of a similarity function and pairwise transformations between patches are computed. Eventually (iii, in the structure regularity estimation step the parameters of repeated grid patterns are calculated by using a Least- Squares optimization. Workability of the presented approach is tested using some real data from urban scenes.

  8. 30 CFR 57.4530 - Exits for surface buildings and structures.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Exits for surface buildings and structures. 57... Fire Prevention and Control Installation/construction/maintenance § 57.4530 Exits for surface buildings and structures. Surface buildings or structures in which persons work shall have a sufficient number...

  9. Progress of decommissioning of Rikkyo reactor in FY2014

    International Nuclear Information System (INIS)

    Suzuki, M.; Kato, M.; Tanzawa, T.; Kawaguchi, K.; Terasawa, T.; Yamada, Shigeru; Nakai, Masaru

    2015-01-01

    Institute for Atomic Energy, Rikkyo University, applied in 2012 for changes in the decommissioning plan toward the abolition of the reactor facilities, and received approval. It promoted the decommissioning work of the research reactors in a plan for two years from 2012, conducted the removal of the structure installed in the reactor tank and storage management measures, and implemented the function stop of the disposal facility of liquid waste and the removal of part of them. These procedures achieved the safe storage condition of core internal structure / equipment with relatively high radioactivity due to neutron irradiation. In addition, the maintenance management of partial facilities and equipment that had been maintained in operational conditions had come to be unnecessary. Based on these results, the implementation plan for decommissioning scheduled for 2015-2016 was prepared. The contents of main works are as follows: (1) dismantling and removal of disposal facilities for liquid waste and storage management of subsequently generated radioactive waste in the reactor building control area, (2) storage management of radioactive solid waste of solid waste storage facilities in the reactor building control area, (3) dismantling and removal of solid waste storage facilities that become unnecessary, and (4) release of part of the controlled area associated with the above actions. (A.O.)

  10. Seismic analysis of the pile foundation of the reactor building of the NPP ANGRA 2

    International Nuclear Information System (INIS)

    Wolf, J.P.; Arx, G.A. von; Barros, F.C.P. de; Kakubo, M.

    1981-01-01

    A pile foundation subjected to dynamic loads interacts with the surrounding soil. Frequency-dependent stiffness and radiation damping must be properly taken into account in pile-soil-pile interaction. Assuming that the soil consists of horizontal layers of elastic material with hysteretic damping, the dynamic stiffness of a group of (even battered) piles can be determined, accounting rigorously for the cavities where the soil is subsequently replaced by the piles. By way of illustration, this substructure procedure, which works in the frequency domain, is applied to the final design of the pile foundation of the Reactor Building of Angra 2 in Brazil. Below the basemat, a strongly horizontally-layered compressive soil of 36 m thickness rests on bedrock. The reactor building is founded on 202 endbearing piles and 88 floating piles of 15 m length. Every pile is modelled. Along each pile, compatibility between the pile and the soil in all three directions is formulated in seven nodes. The basemat is assumed to be rigid. On the level of bedrock a broad-banded response spectrum specifies the excitation (outcropping). (orig./WL)

  11. Structural analysis of the Upper Internals Structure for the Clinch River Breeder Reactor Plant

    International Nuclear Information System (INIS)

    Houtman, J.L.

    1979-01-01

    The Upper Internals Structure (UIS) of the Clinch River Breeder Reactor Plant (CRBRP) provides control of core outlet flow to prevent severe thermal transients from occuring at the reactor vessel and primary heat transport outlet piping, provides instrumentation to monitor core performance, provides support for the control rod drivelines, and provides secondary holddown of the core. All of the structural analysis aspects of assuring the UIS is structurally adequate are presented including simplified and rigorous inelastic analysis methods, elevated temperature criteria, environmental effects on material properties, design techniques, and manufacturing constraints

  12. Fuel, structural material and coolant for an advanced fast micro-reactor

    International Nuclear Information System (INIS)

    Nascimento, Jamil A. do; Guimaraes, Lamartine N.F.; Ono, Shizuca

    2011-01-01

    The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials. (author)

  13. Verification Survey of the Building 315 Zero Power Reactor-6 Facility, Argonne National Laboratory-East, Argonne, Illinois

    International Nuclear Information System (INIS)

    W. C. Adams

    2007-01-01

    Oak Ridge Institute for Science and Education (ORISE) conducted independent verification radiological survey activities at Argonne National Laboratory's Building 315, Zero Power Reactor-6 facility in Argonne, Illinois. Independent verification survey activities included document and data reviews, alpha plus beta and gamma surface scans, alpha and beta surface activity measurements, and instrumentation comparisons. An interim letter report and a draft report, documenting the verification survey findings, were submitted to the DOE on November 8, 2006 and February 22, 2007, respectively (ORISE 2006b and 2007). Argonne National Laboratory-East (ANL-E) is owned by the U.S. Department of Energy (DOE) and is operated under a contract with the University of Chicago. Fundamental and applied research in the physical, biomedical, and environmental sciences are conducted at ANL-E and the laboratory serves as a major center of energy research and development. Building 315, which was completed in 1962, contained two cells, Cells 5 and 4, for holding Zero Power Reactor (ZPR)-6 and ZPR-9, respectively. These reactors were built to increase the knowledge and understanding of fast reactor technology. ZPR-6 was also referred to as the Fast Critical Facility and focused on fast reactor studies for civilian power production. ZPR-9 was used for nuclear rocket and fast reactor studies. In 1967, the reactors were converted for plutonium use. The reactors operated from the mid-1960's until 1982 when they were both shut down. Low levels of radioactivity were expected to be present due to the operating power levels of the ZPR's being restricted to well below 1,000 watts. To evaluate the presence of radiological contamination, DOE characterized the ZPRs in 2001. Currently, the Melt Attack and Coolability Experiments (MACE) and Melt Coolability and Concrete Interaction (MCCI) Experiments are being conducted in Cell 4 where the ZPR-9 is located (ANL 2002 and 2006). ANL has performed final

  14. Clearance of buildings for demolition: ways to clearance on the standing structure for covered surfaces and inaccessible areas

    International Nuclear Information System (INIS)

    Thraenert, S.; Riemann, T.

    2014-01-01

    structure for about half of this area, and until 2014 most of the remaining area will be treated accordingly. However, due to the building design of Wuergassen NPP, not all surfaces are directly accessible to full-area characterization on the standing structure and not all areas can be completely decontaminated in a cost-effective manner. The most prominent structure is the open gap between the turbine foundations and the surrounding turbine building, but similar gaps also exist between the turbine building and, e.g., the reactor building. In addition, the lower basement floors of the RCA buildings consist of several horizontal layers of concrete which, during construction, did not completely bond. As a result, contaminated liquids propagated along the existing boundary surfaces and hot spots up to 100 kBq are now covered by about 0.5 m of low-activity material. For this variety of challenging examples Wuergassen NPP proposed the release process to the competent authority, performed measurements and provided interpretation of the data. The release process is aimed at verifying compliance with the clearance levels on the standing structure. TUV NORD Nuclear is involved in the release process on behalf of the competent authority to expertise the procedure, to supervise the measurements and to review the data evaluation. Regarding the boundary surfaces between two concrete layers it was possible to gain access to the covered surface by fragmenting the upper concrete layer into blocks. The decontaminated surfaces comply with the clearance levels and the decontaminated concrete blocks may remain on site. An existing gap between the turbine building and the reactor building was characterized by using special equipment. It could be shown, that the residual contamination within the gap complies with the clearance levels. If it is not possible to do a complete decontamination according to static reasons, the residual radioactivity can be determined for further use in the dose

  15. Development of the APR+ Auxiliary Building General Arrangement (GA)

    International Nuclear Information System (INIS)

    Moon, Hyung Keun; Park, Young Sheop; Kang, Yong Chul

    2011-01-01

    The general arrangement (GA) drawing of a nuclear power plant is the most basic drawing which contains all of the plant equipment, systems, and rooms. Therefore, it should be issued at an early design stage to provide the contours of the overall plant structure. This type of drawing is typically used widely throughout the design stages. The development project of APR+ (Advanced Power Reactor+), as a succeeding model of the APR1400 (Advanced Power Reactor 1400) design, has its own GA that encompasses all of its power buildings. This was developed starting in October of 2009. Among several of the buildings in this design, the Auxiliary Building (AB) is one of the most important buildings to produce electricity, and to protect against undesirable radiation emissions. This paper focuses on the design characteristics of the general arrangement of the AB

  16. A contribution to the static and dynamic calculation of research reactor structures

    International Nuclear Information System (INIS)

    Goncalves Filho, O.J.A.; Brito Aghina, L.O. de; Gomes, P.A.

    1978-01-01

    Some results in the analysis of a research reactor, using the finite element method are presented. The distribution of internal forces is discussed for the conditions of a Borax accident. An special computer automatic program for the static and dynamic analysis of this Kind of reactor buildings was developed. The program may use either plane triangular elements or double-curvature shell elements and allows the analysis of laminated shells, as it the case of concrete containment vessels with steel liners. (Author)

  17. Parametric study of the Ignalina reactor building capability as barrier against accidental releases of radioactivity

    International Nuclear Information System (INIS)

    Blomquist, R.; Johansson, Kjell; Nilsson, Lars.

    1993-01-01

    The results of a parametric study are offered to the Ignalina plant management staff and to the Lithuanian and Swedish nuclear inspectorates as a basis for a decision whether there is mutual interest in a project for the purpose of strengthening the Ignalina reactor buildings inherent capabilities to provide a barrier against accidental releases of radioactivity. Practical measures to consider are: * establish natural convection of warm air from the steam drums to the tall stack of 150 m height. * reduce the resulting draught of air through the reactor hall floor between the fuel channel shield blocks into the steam drum compartments. * apply filtration to the stack air flow. 18 refs

  18. Structured building model reduction toward parallel simulation

    Energy Technology Data Exchange (ETDEWEB)

    Dobbs, Justin R. [Cornell University; Hencey, Brondon M. [Cornell University

    2013-08-26

    Building energy model reduction exchanges accuracy for improved simulation speed by reducing the number of dynamical equations. Parallel computing aims to improve simulation times without loss of accuracy but is poorly utilized by contemporary simulators and is inherently limited by inter-processor communication. This paper bridges these disparate techniques to implement efficient parallel building thermal simulation. We begin with a survey of three structured reduction approaches that compares their performance to a leading unstructured method. We then use structured model reduction to find thermal clusters in the building energy model and allocate processing resources. Experimental results demonstrate faster simulation and low error without any interprocessor communication.

  19. Integrated leak rate testing of the fast flux test facility reactor containment building

    International Nuclear Information System (INIS)

    James, E.B.; Farabee, O.A.; Bliss, R.J.

    1978-01-01

    The initial Integrated Leak Rate Test (ILRT) of the Fast Flux Test Facility containment building was performed from May 27 to June 2, 1978. The test was conducted in air with systems vented and with the containment recirculating coolers in operation. 10 psig and 5 psig tests were run using the absolute pressure test method. The measured leakage rates were .033% Vol/24 hr. and -.0015% Vol/24 hrs. respectively. Subsequent verification tests at both 10 psig and 5 psig proved that the test equipment was operating properly and it was sensitive enough to detect leaks at low pressures. This ILRT was performed at a lower pressure than any previous ILRT on a reactor containment structure in the United States. While the initial design requirements for ice condenser containments called for a part pressure test at 6 psig, the tests were waived due to the apparent statistical problems of data analysis and the repeatability of the data itself at such low pressure. In contrast to this belief, both the 5 and 10 psig ILRT's were performed in a successful manner at FFTF

  20. 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

  1. Nuclear reactor having an inflatable vessel closure seal structure

    International Nuclear Information System (INIS)

    1980-01-01

    An improved type of closure head seal for the rotatable plugs of the reactor vessel of a liquid metal fast breeder reactor is described. The seal prevents the release of radioactive particles while allowing the plug to be rotated without major manipulation of the seal structure. (UK)

  2. Assessment of structural reliability of precast concrete buildings

    Directory of Open Access Journals (Sweden)

    Koyankin Alexandr

    2018-01-01

    Full Text Available Precast housing construction is currently being under rapid development, however, reliability of building structures made from precast reinforced concrete cannot be assessed rationally due to insufficient research data on that subject. In this regard, experimental and numerical studies were conducted to assess structural reliability of precast buildings as described in the given paper. Experimental studies of full-scale and model samples were conducted; numerical studies were held based on finite element models using “Lira” software. The objects under study included fragment of flooring of a building under construction, full-size fragment of flooring, full-scale models of precast cross-beams-to-columns joints and joints between hollow-core floor slabs and precast and cast-in-place cross-beams. Conducted research enabled to perform an objective assessment of structural reliability of precast buildings.

  3. Reactor costs and maintenance, with reference to the Culham Mark II conceptual tokamak reactor design

    International Nuclear Information System (INIS)

    Hancox, R.; Mitchell, J.T.D.

    1977-01-01

    Published designs of tokamak reactors have proposed conceptual solutions for most of the technological problems encountered. Two areas which remain uncertain, however, are the capital cost of the reactor and the practicability of reactor maintenance. A cost estimate for the Culham Conceptual Tokamak Reactor (Mk I) is presented. The capital cost of a power station incorporating this reactor would be significantly higher than that of an equivalent fast breeder fission power station, mainly because of the low power density of the fusion reactor which affects both the reactor and building costs. To reduce the fusion station capital costs a new conceptual design is proposed (Mk II) which incorporates a shaped plasma cross-section to give a higher plasma pressure ratio, βsub(t) approximately 0.1. Since the higher power density implies more severe radiation damage of the blanket structure, the question of reactor maintenance assumes greater importance. With the proposed scheme for regular replacement of the blanket, a fusion power station availability around 0.9 should be achievable. (author)

  4. Reactor costs and maintenance, with reference to the Culham Mark II conceptual Tokamak reactor design

    International Nuclear Information System (INIS)

    Hancox, R.; Mitchell, J.T.D.

    1976-01-01

    Published designs of tokamak reactors have proposed conceptual solutions for most of the technological problems encountered. Two areas which remain uncertain, however, are capital cost of the reactor and the practicability of reactor maintenance. A cost estimate for the Culham Conceptual Tokamak Reactor (Mk I) is presented. The capital cost of a power station incorporating this reactor would be significantly higher than that of an equivalent fast breeder fission power station, due mainly to the low power density of the fusion reactor which affects both the reactor and building costs. In order to reduce the fusion station capital costs a new conceptual design is proposed (Mk II) which incorporates a shaped plasma cross-section to give a higher plasma pressure ratio, βsub(t) approximately 0.1. Since the higher power density implies more severe radiation damage of the blanket structure, the question of reactor maintenance assumes greater importance. With the proposed scheme for regular replacement of the blanket, a fusion power station availability around 0.9 should be achievable. (orig.) [de

  5. Lining up device for the internal structures of a nuclear reactor vessel

    International Nuclear Information System (INIS)

    Silverblatt, B.L.

    1977-01-01

    The invention concerns a nuclear reactor of the type with a vessel, a vessel head carried at the top of this vessel by a core cylinder comprising a flange internally supported by the vessel, and an upper support structure supported between the core cylinder flange and the vessel head to align laterally the head, vessel, flange and support structure. A bottom key device is provided for lining up the flange, support structure and vessel, and an upper key device for laterally lining up support structure and the vessel head and for maintaining this alignment when they are removed simultaneously from the core cylinder and vessel. When re-assembling the reactor, the top support structure and the vessel head are lowered simultaneously so that an opening in the top alignment structure engages in the upper extension of the bottom alignment structure. A plurality of alignment stuctures may be utilised round the circumference of the reactor vessel. The disposition of the invention also facilitates the removal of the core cylinder from the reactor vessel. In this way, the alignment on re-assembly is ensured by the re-entry of the bottom extension under the flange of the core cylinder with the groove or keyway of the reactor vessel [fr

  6. Elements for computing and forecasting the leakage rate of the inner containment of nuclear reactor buildings

    International Nuclear Information System (INIS)

    Asali, M.; Capra, B.; Mazars, J.; Colliat, J.B.

    2015-01-01

    This study aims at introducing a methodology based on a macro-element discretization to compute and forecast the air leak rate of double-wall reactor buildings during air pressure tests. Assumptions at the basis of a weakly coupled strategy are checked in the case of a typical porous concrete section of an inner containment modeled during a 33 year period including four decennial regulatory pressure tests. However, air leakage due to porosity is only part of in situ measurements. Leakage due to cracking is another part and should be taken into account. A first macro-element is then presented, that superimposes Darcy flow within a porous matrix together with Poiseuille flow within a crack. Those elements are then used in a 3D hydraulic model to compute more accurately the total air leakage rate of considered structures. (authors)

  7. Assessment of Extent and Degree of Thermal Damage to Polymeric Materials in the Three Mile Island Unit 2 Reactor Building

    International Nuclear Information System (INIS)

    Alvares, N. J.

    1984-02-01

    Thermal damage to susceptible materials in accessible regions of the TMI-2 reactor building shows damage-distribution patterns that indicate non-uniform intensity of exposure. No clear explanation for non-uniformity is found in existing evidence; e.g., in some regions a lack of thermally susceptible materials frustrates analysis. Elsewhere, burned materials are present next to materials that seem similar but appear unscathed-leading to conjecture that the latter materials preferentially absorb water vapor during periods of high local steam concentration. Most of the polar crane pendant shows heavy burns on one half of its circumferential surface. This evidence suggests that the polar crane pendant side that experienced heaviest burn damage was exposed to intense radiant energy from a transient fire plume in the reactor containment volume. Tests and simple heat-transfer calculations based on pressure and temperature records from the accident show that the atmosphere inside the reactor building was probably 8% hydrogen in air, a value not inconsistent with the extent of burn damage. Burn-pattern geography indicates uniform thermal exposure in the dome volume to the 406-ft level (about 6 ft below the polar crane girder), partial thermal exposure in the volume between the 406- and 347-ft levels as indicated by the polar crane cable, and lack of damage to most thermally susceptible materials in the west quadrant of the reactor building; some evidence of thermal exposure Is seen in the free volume between the 305- and 347-ft levels. (author)

  8. Procedures of ASME code case N-201 for KALIMER. Reactor internal structures

    International Nuclear Information System (INIS)

    Koo, Gyeong Hoi; Yoo, B.

    2001-02-01

    The main objective of this report is to describe the design procedure of ASME Boiler and Pressure Vessel Code, Code Case N-201-4, which is an elevated temperature structural design code of the Nuclear reactor internal structures, checking the criteria of stress limit, accumulated inelastic strain and deformation, creep-fatigue damage, and buckling limit. As one of examples, the creep-fatigue damage evaluations are carried out for the KALIMER reactor internal structures of baffle annulus. This report is expected to be very useful in evaluating the structural integrity of the liquid metal reactor operating under an elevated temperature

  9. High-temperature-structural design and research and development for reactor system components

    International Nuclear Information System (INIS)

    Matsumura, Makoto; Hada, Mikio

    1985-01-01

    The design of reactor system components requires high-temperature-structural design guide with the consideration of the creep effect of materials related to research and development on structural design. The high-temperature-structural design guideline for the fast prototype reactor MONJU has been developed under the active leadership by Power Reactor and Nuclear Fuel Development Corporation and Toshiba has actively participated to this work with responsibility on in-vessel components, performing research and development programs. This paper reports the current status of high-temperature-structural-design-oriented research and development programs and development of analytical system including stress-evaluation program. (author)

  10. FEM Updating of the Heritage Court Building Structure

    DEFF Research Database (Denmark)

    Ventura, C. E.; Brincker, Rune; Dascotte, E.

    2001-01-01

    . The starting model of the structure was developed from the information provided in the design documentation of the building. Different parameters of the model were then modified using an automated procedure to improve the correlation between measured and calculated modal parameters. Careful attention......This paper describes results of a model updating study conducted on a 15-storey reinforced concrete shear core building. The output-only modal identification results obtained from ambient vibration measurements of the building were used to update a finite element model of the structure...

  11. FRF-based structural damage detection of controlled buildings with podium structures: Experimental investigation

    Science.gov (United States)

    Xu, Y. L.; Huang, Q.; Zhan, S.; Su, Z. Q.; Liu, H. J.

    2014-06-01

    How to use control devices to enhance system identification and damage detection in relation to a structure that requires both vibration control and structural health monitoring is an interesting yet practical topic. In this study, the possibility of using the added stiffness provided by control devices and frequency response functions (FRFs) to detect damage in a building complex was explored experimentally. Scale models of a 12-storey main building and a 3-storey podium structure were built to represent a building complex. Given that the connection between the main building and the podium structure is most susceptible to damage, damage to the building complex was experimentally simulated by changing the connection stiffness. To simulate the added stiffness provided by a semi-active friction damper, a steel circular ring was designed and used to add the related stiffness to the building complex. By varying the connection stiffness using an eccentric wheel excitation system and by adding or not adding the circular ring, eight cases were investigated and eight sets of FRFs were measured. The experimental results were used to detect damage (changes in connection stiffness) using a recently proposed FRF-based damage detection method. The experimental results showed that the FRF-based damage detection method could satisfactorily locate and quantify damage.

  12. Performance tests of the reactor containment structures of HTTR

    International Nuclear Information System (INIS)

    Sakaba, Nariaki; Iigaki, Kazuhiko; Kawaji, Satoshi; Iyoku, Tatsuo

    1998-03-01

    The containment structures of the HTTR consist of the reactor containment vessel (CV), service area (SA) and emergency air purification system, which minimize the release of FPs in the postulated accidents with FP release from the reactor facilities. The CV is designed to withstand the temperature and pressure transients and to be leak-tight within the specified leakage limit even in the case of a rupture of the primary concentric hot gas duct. The pressure of inside of the SA should be maintained slightly lower than that of atmosphere by the emergency air purification system. The radioactive materials are released from the stack to environment via the emergency air purification system under the accident condition. Then the emergency air purification system should remove airborne radio-activities and should maintain proper pressure in the SA. We established the method to measure leak rate of the CV with closed reactor coolant pressure boundary although it is normally measured under opened reactor coolant pressure boundary as employed in LWRs. The CV leak rate test was carried out by the newly developed method and the expected performance was obtained. The SA and emergency air purification system were also confirmed by the performance test. We concluded that the reactor containment structures were fabricated to minimize the release of FPs in the postulated accidents with FP release from the reactor facilities. (author)

  13. Ageing evaluation model of nuclear reactors structural elements

    International Nuclear Information System (INIS)

    Ziliukas, A.; Jutas, A.; Leisis, V.

    2002-01-01

    In this article the estimation of non-failure probability by random faults on the structural elements of nuclear reactors is presented. Ageing is certainly a significant factor in determining the limits of nuclear plant lifetime or life extensions. Usually the non failure probability rates failure intensity, which is characteristic for structural elements ageing in nuclear reactors. In practice the reliability is increased incorrectly because not all failures are fixed and cumulated. Therefore, the methodology with using the fine parameter of the failures flow is described. The comparison of non failure probability and failures flow is carried out. The calculation of these parameters in the practical example is shown too. (author)

  14. Decommissioning of TRIGA Mark II type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dooseong; Jeong, Gyeonghwan; Moon, Jeikwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    The first research reactor in Korea, KRR 1, is a TRIGA Mark II type with open pool and fixed core. Its power was 100 kWth at its construction and it was upgraded to 250 kWth. Its construction was started in 1957. The first criticality was reached in 1962 and it had been operated for 36,000 hours. The second reactor, KRR 2, is a TRIGA Mark III type with open pool and movable core. These reactors were shut down in 1995, and the decision was made to decommission both reactors. The aim of the decommissioning activities is to decommission the KRR 2 reactor and decontaminate the residual building structures and site, and to release them as unrestricted areas. The KRR 1 reactor was decided to be preserve as a historical monument. A project was launched for the decommissioning of these reactors in 1997, and approved by the regulatory body in 2000. A total budget for the project was 20.0 million US dollars. It was anticipated that this project would be completed and the site turned over to KEPCO by 2010. However, it was discovered that the pool water of the KRR 1 reactor was leaked into the environment in 2009. As a result, preservation of the KRR 1 reactor as a monument had to be reviewed, and it was decided to fully decommission the KRR 1 reactor. Dismantling of the KRR 1 reactor takes place from 2011 to 2014 with a budget of 3.25 million US dollars. The scope of the work includes licensing of the decommissioning plan change, removal of pool internals including the reactor core, removal of the thermal and thermalizing columns, removal of beam port tubes and the aluminum liner in the reactor tank, removal of the radioactive concrete (the entire concrete structure will not be demolished), sorting the radioactive waste (concrete and soil) and conditioning the radioactive waste for final disposal, and final statuses of the survey and free release of the site and building, and turning over the site to KEPCO. In this paper, the current status of the TRIGA Mark-II type reactor

  15. Structural acceptance criteria Remote Handling Building Tritium Extraction Facility

    Energy Technology Data Exchange (ETDEWEB)

    Mertz, G.

    1999-12-16

    This structural acceptance criteria contains the requirements for the structural analysis and design of the Remote Handling Building (RHB) in the Tritium Extraction Facility (TEF). The purpose of this acceptance criteria is to identify the specific criteria and methods that will ensure a structurally robust building that will safely perform its intended function and comply with the applicable Department of Energy (DOE) structural requirements.

  16. Structural acceptance criteria Remote Handling Building Tritium Extraction Facility

    International Nuclear Information System (INIS)

    Mertz, G.

    1999-01-01

    This structural acceptance criteria contains the requirements for the structural analysis and design of the Remote Handling Building (RHB) in the Tritium Extraction Facility (TEF). The purpose of this acceptance criteria is to identify the specific criteria and methods that will ensure a structurally robust building that will safely perform its intended function and comply with the applicable Department of Energy (DOE) structural requirements

  17. Code qualification of structural materials for AFCI advanced recycling reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Li, M.; Majumdar, S.; Nanstad, R.K.; Sham, T.-L. (Nuclear Engineering Division); (ORNL)

    2012-05-31

    This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP

  18. RA Reactor

    International Nuclear Information System (INIS)

    1978-02-01

    In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation [sr

  19. On elastic structural elements for nuclear reactors

    International Nuclear Information System (INIS)

    Povolo, F.

    1978-03-01

    The in-pile stress-relaxation behaviour of materials usually employed for the elastic structural elements, in nuclear reactors, is critically reviewed and the results are compared with those obtained in commercial zirconium alloys irradiated under similar conditions. Finally, it is shown that, under certain conditions, some zirconium alloys may be used as an alternative material for these structural elements. (orig.) [de

  20. Analysis of the structural design process of the adaptive reuse of building structures

    NARCIS (Netherlands)

    Pasterkamp, S.

    2014-01-01

    In the field of structural building engineering there is a market shift taking place as a result of the growing number of buildings that are listed as cultural heritage, secularization, the economic situation and the increasing office vacancy rate in Europe and the US. More and more structural

  1. Prediction of failure modes for concrete nuclear-containment buildings

    International Nuclear Information System (INIS)

    Butler, T.A.

    1982-01-01

    The failure modes and associated failure pressures for two common generic types of PWR containments are predicted. One building type is a lightly reinforced, posttensioned structure represented by the Zion nuclear reactor containment. The other is the normally reinforced Indian Point containment. Two-dimensional models of the buildings developed using the finite element method are used to predict the failure modes and failure pressures. Predicted failure modes for both containments involve loss of structural integrity at the intersection of the cylindrical sidewall with the base slab

  2. Seismic analysis, evaluation and upgrade design for a DOE exhaust stack building

    International Nuclear Information System (INIS)

    Malik, L.E.; Maryak, M.E.

    1991-01-01

    An exhaust stack building of a nuclear reactor facility with complex structural configuration has been analyzed and evaluated and retrofitted for seismic forces. The building was built in the 1950's and had not been designed to resist seismic forces. A rigorous analysis and evaluation program was implemented to minimize costly retrofits required to upgrade the building to resist high seismic forces. Seismic evaluations were performed for the building in its as-is configuration, and as modified for several upgrade schemes. Soil-structure-interaction, basemat flexibility and the influence of the nearby reactor building were considered in rigorous seismic analyses. These analyses and evaluations enabled limited upgrades to qualify the stack building for the seismic forces. Some of the major conclusions of this study are: (1) a phased approach of seismic analyses, utilizing simplified models to evaluate practicable upgrade schemes, and, then incorporating the most suitable scheme in a rigorous model to obtain design forces for upgrades, is an efficient and cost-effective approach for seismic qualification of nuclear facilities to higher seismic criteria; and, (2) finalizing the upgrade of a major nuclear facility is an iterative process, which continues throughout the construction of the upgrades

  3. Performance evaluation of existing building structure with pushover analysis

    Science.gov (United States)

    Handana, MAP; Karolina, R.; Steven

    2018-02-01

    In the management of the infrastructure of the building, during the period of buildings common building damage as a result of several reasons, earthquakes are common. The building is planned to work for a certain service life. But during the certain service life, the building vulnerable to damage due to various things. Any damage to cultivate can be detected as early as possible, because the damage could spread, triggering and exacerbating the latest. The newest concept to earthquake engineering is Performance Based Earthquake Engineering (PBEE). PBEE divided into two, namely Performance Based Seismic Design (PBSD) and Performance Based Seismic Evaluation (PBSE). Evaluation on PBSE one of which is the analysis of nonlinear pushover. Pushover analysis is a static analysis of nonlinear where the influence of the earthquake plan on building structure is considered as burdens static catch at the center of mass of each floor, which it was increased gradually until the loading causing the melting (plastic hinge) first within the building structure, then the load increases further changes the shapes of post-elastic large it reached the condition of elastic. Then followed melting (plastic hinge) in the location of the other structured.

  4. Answers to questions about removing krypton from the Three Mile Island, Unit 2 reactor building. Public information report

    International Nuclear Information System (INIS)

    1980-05-01

    This document presents answers to frequently asked questions about the probable effects of controlled releases of the krypton presently contained within the reactor building of Three Mile Island, Unit 2. Also answered are questions about alternative means for removing the krypton

  5. Elementary structural building blocks encountered in silicon surface reconstructions

    International Nuclear Information System (INIS)

    Battaglia, Corsin; Monney, Claude; Didiot, Clement; Schwier, Eike Fabian; Garnier, Michael Gunnar; Aebi, Philipp; Gaal-Nagy, Katalin; Onida, Giovanni

    2009-01-01

    Driven by the reduction of dangling bonds and the minimization of surface stress, reconstruction of silicon surfaces leads to a striking diversity of outcomes. Despite this variety even very elaborate structures are generally comprised of a small number of structural building blocks. We here identify important elementary building blocks and discuss their integration into the structural models as well as their impact on the electronic structure of the surface. (topical review)

  6. REACTOR GROUT THERMAL PROPERTIES

    Energy Technology Data Exchange (ETDEWEB)

    Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

    2011-01-28

    Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

  7. Natural vibration experimental analysis of Novovoronezhskaya NPP main building

    International Nuclear Information System (INIS)

    Zoubkov, D.; Isaikin, A.; Shablinsky, G.; Lopanchuk, A.; Nefedov, S.

    2005-01-01

    1. Natural vibration frequencies are main characteristics of buildings and structures which allow to give integral estimation of their in-service state. Even relatively small changes of these frequencies as compared to the initially registered values point to serious defects of building structures. In this paper we analyzed natural vibration frequencies and natural modes of the main building (MB) of Novovoronezhskaya NPP operating nuclear unit with WWER-440 type reactor. The MB consists of a reactor compartment (RC), a machine room (MR) and an electric device (ED) unit positioned in between. 2. Natural vibration frequencies and natural modes of the MB were determined experimentally by analyzing its microvibrations caused by operation of basic equipment (turbines, pumps, etc.). Microvibrations of the main building were measured at 12 points. At each point measurements were carried out along two or three mutually perpendicular vibration directions. Spectral analysis of vibration records has been conducted. Identification of natural vibration frequencies was carried out on the basis of the spectral peaks and plotted vibration modes (taking into account operating frequencies of the basic equipment of the power generating unit). On the basis of the measurement results three transverse modes and corresponding natural vibration frequencies of the MB, one longitudinal mode and corresponding natural vibration frequency of the MB and two natural frequencies of vertical vibrations of RC and MR floor trusses (1st and 2nd symmetric forms) were determined. Dynamic characteristics of the main building of NV NPP resulting from full scale researches are supposed to be used as one of building structure stability criteria. (authors)

  8. The application of mechanical desktop in the design of the reactor core structure of China advanced research reactor

    International Nuclear Information System (INIS)

    Lang Ruifeng

    2002-01-01

    The three-dimensional parameterization design method is introduced to the design of reactor core structure for China advanced research reactor. Based on the modeling and dimension variable driving of the main parts as well as the modification of dimension variable, the preliminary design and modification of reactor core is carried out with high design efficiency and quality as well as short periods

  9. Behaviour of a reactor PWR containment submitted to an external explosion

    International Nuclear Information System (INIS)

    Barbe, B.; Avet-Flancard, R.; Perrot, J.; Berriaud, C.; Dulac, J.

    1981-01-01

    The aims of this study are to obtain experimental data and theoretical evaluation of the transient field pressure existing on importants buildings of the plant. The knowledge of the pressure loading permits then to predict the structure mechanical behaviour. For this purpose the cylindrical reactor building and the parallelepipedic fuel building have been modelized to a 1/40 scale. These models were realized as carefully as possible with prestressing in the thickness of microconcrete walls and were submitted to incident shock waves obtained by T.N.T. explosions. Several characteristics explosion directions have been tested. Experimental data were recorded with pressure and displacement transducers and also by accelerometers. The results show that: 1) the geometrical dihedral between reactor and fuel building induces local overpressures five times the incident pressures; 2) no apparent damage occurred on the structure, for the range of field pressure tested so far; this may related to only small effects of resonances. Simultaneously a tridimensional, acoustic code has been developed an conveniently correlates experimental data. (orig./HP)

  10. An approach to build a knowledge base for reactor accident diagnostic expert system

    International Nuclear Information System (INIS)

    Yoshida, K.; Fujii, M.; Fujiki, K.; Yokobayashi, M.; Kohsaka, A.; Aoyagi, T.; Hirota, Y.

    1987-01-01

    In the development of a rule based expert system, one of the key issues is how to acquire knowledge and to build knowledge base (KB). On building the KB of DISKET, which is an expert system for nuclear reactor accident diagnosis developed in JAERI, several problems have been experienced as follows. To write rules is a time consuming task, and it is difficult to keep the objectivity and consistency of rules as the number of rules increase. Further, certainty factors (CFs) must be often determined according to engineering judgment, i.e., empirically or intuitively. A systematic approach was attempted to handle these difficulties and to build an objective KB efficiently. The approach described in this paper is based on the concept that a prototype KB, colloquially speaking an initial guess, should first be generated in a systematic way and then is to be modified and/or improved by human experts for practical use. Statistical methods, principally Factor Analysis, were used as the systematic way to build a prototype KB for the DISKET using a PWR plant simulator data. The source information is a number of data obtained from the simulation of transients, such as the status of components and annunciator etc., and major process parameters like pressures, temperatures and so on

  11. Vibration system identification of Paks and Kozloduy reactor buildings on the basis of the blast test results

    International Nuclear Information System (INIS)

    Varpasuo, P.

    1999-01-01

    System identification allows to build mathematical models of a dynamic system based on measured data. System identification is carried out by adjusting parameters within a given model until its output coincides as well as possible with the measured output. The aim of this study is to investigate and model the behavior of complex vibratory systems on the basis of measured excitation and response. The first part of the study describes the theory used in the analysis and the software tools used in the analysis. The second part of the study describes the investigation and modeling of the response of single degree of freedom oscillator excited by sinusoidal and blast excitation. In the third part of the study the system identification of the Kozloduy NPP unit 5 reactor building and Paks NPP unit 1 reactor building is studied and the models are estimated using the method of segmentation of excitation and response. System identification is carried out using MATLAB software by adjusting parameters within a given model until its output coincides as well as possible with the measured output. The types of models used for the were: l) ARX models; 2) ARMAX model; 3) Output-Error (OE) models; 4) Box-Jenkins (BJ) models; 5) State-space models. The model coefficients for different models were calculated using the least-squares and maximum likelihood estimation methods available in MATLAB system identification toolbox. Excitation was in both Paks and Kozloduy case the measured free-field excitation and responses were the vibration responses of the building on the foundation slab level and top of the building. By examining the established models the frequency characteristics of vibration systems were determined with 95 % accuracy and the amplitude response with 80 % accuracy. In case of the steady state response of sinusoidally excited single dof oscillator the modelling gave almost exact results. But in the case of the blast response of the reactor building the obtaining of the

  12. Integrated structural design of nuclear power plants for high seismic areas

    International Nuclear Information System (INIS)

    Rieck, P.J.

    1979-01-01

    A design approach which structurally interconnects NPP buildings to be located in high seismic areas is described. The design evolution of a typical 600 MWe steel cylindrical containment PWR is described as the plant is structurally upgraded for higher seismic requirements, while maintaining the original plant layout. The plant design is presented as having separate reactor building and auxiliary structures for a low seismic area (0.20 g) and is structurally combined at the foundation for location in a higher seismic area (0.30 g). The evolution is completed by a fully integrated design which structurally connects the reactor building and auxiliary structures at superstructure elevations as well as foundation levels for location in very severe seismic risk areas (0.50 g). (orig.)

  13. Thermal-structural response of EBR-II major components under reactor operational transients

    International Nuclear Information System (INIS)

    Chang, L.K.; Lee, M.J.

    1983-01-01

    Until recently, the LMFBR safety research has been focused primarily on severe but highly unlikely accident, such as hypothetical-core-disruptive accidents (HCDA's), and not enough attention has been given to accident prevention, which is less severe but more likely sequence. The objective of the EBR-II operational reliability testing (ORT) is to demonstrate that the reactor can be designed and operated to prevent accident. A series of mild duty cycles and overpower transients were designed for accident prevention tests. An assessment of the EBR-II major plant components has been performed to assure structural integrity of the reactor plant for the ORT program. In this paper, the thermal-structural response and structural evaluation of the reactor vessel, the reactor-vessel cover, the intermediate heat exchanger (IHX) and the superheater are presented

  14. Surface activity and radiation field measurements of the TMI-2 reactor building gross decontamination experiment

    International Nuclear Information System (INIS)

    McIsaac, C.V.

    1983-10-01

    Surface samples were collected from concrete and metal surfaces within the Three Mile Island Unit 2 Reactor Building on December 15 and 17, 1981 and again on March 25 and 26, 1982. The Reactor Building was decontaminated by hydrolasing during the period between these dates. The collected samples were analyzed for radionuclide concentration at the Idaho National Engineering Laboratory. The sampling equipment and procedures, and the analysis methods and results are discussed. The measured mean surface concentrations of 137 Cs and 90 Sr on the 305-ft elevation floor before decontamination were, respectively, 3.6 +- 0.9 and 0.17 +- 0.04 μCi/cm 2 . Their mean concentrations on the 347-ft elevation floor were about the same. On both elevations, walls were found to be considerably less contaminated than floors. The fractions of the core inventories of 137 Cs, 90 Sr, and 129 I deposited on Reactor Building surfaces prior to decontamination were calculated using their mean concentrations on various types of surfaces. The calculated values for these three nuclides are 3.5 +- 0.4 E-4, 2.4 +- 0.8 E-5, and 5.7 +- 0.5 E-4, respectively. The decontamination operations reduced the 137 Cs surface activity on the 305- and 347-ft elevations by factors of 20 and 13, respectively. The 90 Sr surface activity reduction was the same for both floors, that being a factor of 30. On the whole, decontamination of vertical surfaces was not achieved. Beta and gamma exposure rates that were measured during surface sampling were examined to determine the degree to which they correlated with measured surface activities. The data were fit with power functions of the form y = ax/sup b/. As might be expected, the beta exposure rates showed the best correlation. Of the data sets fit with the power function, the set of December 1981 beta exposure exhibited the least scatter. The coefficient of determination for this set was calculated to be 0.915

  15. On the classification of structures, systems and components of nuclear research and test reactors

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel

    2009-01-01

    The classification of structures, systems and components of nuclear reactors is a relevant issue related to their design because it is directly associated with their safety functions. There is an important statement regarding quality standards and records that says Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. The definition of the codes, standards and technical requirements applied to the nuclear reactor design, fabrication, inspection and tests may be seen as the main result from this statement. There are well established guides to classify structures, systems and components for nuclear power reactors such as the Pressurized Water Reactors but one can not say the same for nuclear research and test reactors. The nuclear reactors safety functions are those required to the safe reactor operation, the safe reactor shutdown and continued safe conditions, the response to anticipated transients, the response to potential accidents and the control of radioactive material. So, it is proposed in this paper an approach to develop the classification of structures, systems and components of these reactors based on their intended safety functions in order to define the applicable set of codes, standards and technical requirements. (author)

  16. Characterization of the 309 building fuel transfer pit and storage basin

    International Nuclear Information System (INIS)

    Hale, N.S.

    1998-01-01

    This document identifies radiological, chemical and physical conditions inside the Fuel Transfer Pit and Fuel Storage Basins. These spaces are located inside the Plutonium Recycle Test Reactor structure (309 Building.) The fuel handling and storage feature of the PRTR were primarily located in these spaces. The conditions were assessed as part of overall 309 Building transition

  17. Integrity assessment of grouted posttensioning cables and reinforced concrete of a nuclear containment building

    OpenAIRE

    Shenton B.; Philipose K.

    2011-01-01

    The Containment Buildings of CANDU Nuclear Generating Stations were designed to house nuclear reactors and process equipment and also to provide confinement of releases from a potential nuclear accident such as a Loss Of Coolant Accident (LOCA). To meet this design requirement, a post-tensioning system was designed to induce compressive stresses in the structure to counteract the internal design pressure. The CANDU reactor building at Gentilly-1 (G-1), Quebec, Canada (250 MWe) was built in th...

  18. Building on success. The foreign research reactor spent nuclear fuel acceptance program

    International Nuclear Information System (INIS)

    Huizenga, David G.; Mustin, Tracy P.; Saris, Elizabeth C.; Massey, Charles D.

    1998-01-01

    The second year of implementation of the research reactor spent nuclear fuel acceptance program was marked by significant challenges and achievements. In July 1998, the Department of Energy completed by significant challenges and achievements. In July 1998, the Department of Energy completed its first shipment of spent fuel from Asia via the Concord Naval Weapons Station in California to the Idaho National Engineering and Environmental (INEEL). This shipment, which consisted of three casks of spent nuclear fuel from two research reactors in the Republic of Korea, presented significant technical, legal, and political challenges in the United States and abroad. Lessons learned will be used in the planning and execution of our next significant milestone, a shipment of TRIGA spent fuel from research reactors in Europe to INEEL, scheduled for the summer of 1999. This shipment will include transit across the United States for over 2,000 miles. Other challenges and advances include: clarification of the fee policy to address changes in the economic status of countries during the life of the program; resolution of issues associated with cask certification and the specific types and conditions of spent fuel proposed for transport; revisions to standard contract language in order to more clearly address unique shipping situations; and priorization and scheduling of shipments to most effectively implement the program. As of this meeting, eight shipments, consisting of nearly 2,000 spent fuel assemblies from fifteen countries, have been successfully completed. With the continued cooperation of the international research reactor community, we are committed to building on this success in the remaining years of the program. (author)

  19. The intelligent customer: considerations around build-own-operate business and licensing models for small modular reactors in Canada

    International Nuclear Information System (INIS)

    Jones, K.

    2014-01-01

    An organization planning a proposal for a build-own-operate business model needs to address expanded licensee responsibilities under this model, associated regulatory impacts and how this affects their role as an 'intelligent customer'. This is particularly important for cases where builder-owner-operators plan to manufacture factory-fuelled designs and ship them to a site for installation and operation. The primary responsibility for safe conduct of licensed activities rests with the licensee. A build-own-operate model expands the scope of licensed activities to include design, manufacturing, transport, construction, and operation. The licensee must be able to demonstrate they are qualified to conduct all licensed activities including sufficient competent resources within the licensee's organization to oversee('Intelligent Customer') any work it commissions externally and the subsequent flow down through of the supply chain. This paper examines aspects that organizations need to assess the suitability of approaches that it may take to maintain in-house expertise for the control and oversight of licensed activities at all times. It considers the approach to identification of: core capabilities the licensee would need to understand its safety case under a build-own-operate model to manage licensed activities in accordance with requirements under the Nuclear Safety and Control Acta licensee's 'intelligent customer' capabilities in particular around understanding, specifying, overseeing and accepting work undertaken on its behalf by contractors. While this paper is focused on small modular reactors, being an intelligent customer applies to large commercial or research reactors equally; the size of reactor is immaterial.

  20. The intelligent customer: considerations around build-own-operate business and licensing models for small modular reactors in Canada

    Energy Technology Data Exchange (ETDEWEB)

    Jones, K., E-mail: kenneth.jones@cnsc-ccsn.gc.ca [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

    2014-07-01

    An organization planning a proposal for a build-own-operate business model needs to address expanded licensee responsibilities under this model, associated regulatory impacts and how this affects their role as an 'intelligent customer'. This is particularly important for cases where builder-owner-operators plan to manufacture factory-fuelled designs and ship them to a site for installation and operation. The primary responsibility for safe conduct of licensed activities rests with the licensee. A build-own-operate model expands the scope of licensed activities to include design, manufacturing, transport, construction, and operation. The licensee must be able to demonstrate they are qualified to conduct all licensed activities including sufficient competent resources within the licensee's organization to oversee('Intelligent Customer') any work it commissions externally and the subsequent flow down through of the supply chain. This paper examines aspects that organizations need to assess the suitability of approaches that it may take to maintain in-house expertise for the control and oversight of licensed activities at all times. It considers the approach to identification of: core capabilities the licensee would need to understand its safety case under a build-own-operate model to manage licensed activities in accordance with requirements under the Nuclear Safety and Control Acta licensee's 'intelligent customer' capabilities in particular around understanding, specifying, overseeing and accepting work undertaken on its behalf by contractors. While this paper is focused on small modular reactors, being an intelligent customer applies to large commercial or research reactors equally; the size of reactor is immaterial.

  1. Space-time reactor kinetics for heterogeneous reactor structure

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1969-11-15

    An attempt is made to formulate time dependent diffusion equation based on Feinberg-Galanin theory in the from analogue to the classical reactor kinetic equation. Parameters of these equations could be calculated using the existing codes for static reactor calculation based on the heterogeneous reactor theory. The obtained kinetic equation could be analogues in form to the nodal kinetic equation. Space-time distribution of neutron flux in the reactor can be obtained by solving these equations using standard methods.

  2. Seismic analysis and structure capacity evaluation of the Belene nuclear power plant

    International Nuclear Information System (INIS)

    Johnson, J.J.; Hashimoto, P.S.; Campbell, R.D.; Baltus, R.S.

    1993-01-01

    The seismic analysis and structure capacity evaluation of the Belene Nuclear Power Plant, a two-unit WWER 1000, was performed. The principal objective of the study was to review the major aspects of the seismic design including ground motion specification, foundation concept and materials, and the Unit I main reactor building structure response and capacity. The main reactor building structure /foundation/soil were modeled and analyzed by a substructure approach to soil-structure interaction (SSI) analysis. The elements of the substructure approach, implemented in the family of computer programs CLASSI, are: Specification of the free-field ground motion; Modeling the soil profile; SSI parameters; Modeling the structure; SSI-response analyses. Each of these aspects is discussed. The Belene Unit 1 main reactor building structure was evaluated to verify the seismic design with respect to current western criteria. The structural capacity evaluation included criteria development, element load distribution analysis, structural element selection, and structural element capacity evaluation. Equipment and commodity design criteria were similarly reviewed and evaluated. Methodology results and recommendations are presented. (author)

  3. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    Takahashi, Kazumi

    1998-01-01

    The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

  4. Liquid metal systems development: reactor vessel support structure evaluation

    International Nuclear Information System (INIS)

    McEdwards, J.A.

    1981-01-01

    Results of an evaluation of support structures for the reactor vessel are reported. The U ring, box ring, integral ring, tee ring and tangential beam supports were investigated. The U ring is the recommended vessel support structure configuration

  5. A structure for the protection of nuclear-reactor pressurized-vessels against rupture

    International Nuclear Information System (INIS)

    Marcellin, J.-P.; Aubert, Gilles

    1974-01-01

    Description is given of a structure for the protection of nuclear-reactor pressurized-vessels against rupture. Said structure comprises a pre-stressed concrete tank adapted to surround the tank side-wall and bottom, said tank being higher than said vessel, said tank being provided with ports for passing cooling fluid ducts therethrough, and a crown adapted to rest along the periphery of the reactor-cover and made integral therewith. This can be applied to reactors of the PWR type [fr

  6. Safety Research Experiment Facility Project. Conceptual design report. Volume VII. Reactor cooling

    International Nuclear Information System (INIS)

    1975-12-01

    The Reactor Cooling System (RCS) will provide the required cooling during test operations of the Safety Research Experiment Facility (SAREF) reactor. The RCS transfers the reactor energy generated in the core to a closed-loop water storage system located completely inside the reactor containment building. After the reactor core has cooled to a safe level, the stored heat is rejected through intermediate heat exchangers to a common forced-draft evaporative cooling tower. The RCS is comprised of three independent cooling loops of which any two can remove sufficient heat from the core to prevent structural damage to the system components

  7. Effect of Additional Structure on Effective Stack Height of Gas Dispersion in Atmosphere

    Directory of Open Access Journals (Sweden)

    Takenobu Michioka

    2016-03-01

    Full Text Available Wind-tunnel experiments were conducted to evaluate the effect of additional structure (building, sea wall and banking on the effective stack height, which is usually used in safety analyses of nuclear power facilities in Japan. The effective stack heights were estimated with and without the additional structure in addition to the reactor building while varying several conditions such as the source height, the height of additional structure and the distance between the source position and the additional structure. When the source height is equivalent to the reactor building height, the additional structure enhances both the vertical and horizontal gas dispersion widths and decreases the ground gas concentration, and it means that the additional structure does not decrease the effective stack height. When the source height is larger than the reactor height, the additional structures might affect the effective stack height. As the distance between the source and the additional structure decreases, or as the height of the additional structure increases, the structure has a larger effect on the effective stack height.

  8. Report on design and technical standard planning of vibration controlling structure on the buildings, in the Tokai Reprocessing Facility, Power Reactor and Nuclear Fuel Development Corporation

    International Nuclear Information System (INIS)

    Uryu, Mitsuru; Terada, Shuji; Shinohara, Takaharu; Yamazaki, Toshihiko; Nakayama, Kazuhiko; Kondo, Toshinari; Hosoya, Hisashi

    1997-10-01

    The Tokai reprocessing facility buildings are constituted by a lower foundation, vibration controlling layers, and upper structure. At the vibration controlling layer, a laminated rubber aiming support of the building load and extension of the eigenfrequency and a damper aiming absorption of earthquake energy are provided. Of course, the facility buildings are directly supported at the arenaceous shale (Taga Layer) of the Miocene in the Neogene confirmed to the stablest ground, as well the buildings with high vibration resistant importance in Japan. This report shows that when the vibration controlling structure is adopted for the reprocessing facility buildings where such high vibration resistance is required, reduction of input acceleration for equipments and pipings can be achieved and the earthquake resistant safety can also be maintained with sufficient tolerance and reliability. (G.K.)

  9. Structural design and dynamic analysis of underground nuclear reactor containments

    International Nuclear Information System (INIS)

    Kierans, T.W.; Reddy, D.V.; Heale, D.G.

    1975-01-01

    Present actual experience in the structural design of undeground containments is limited to only four rather small reactors all located in Europe. Thus proposals for future underground reactors depend on the transposition of applicable design specifications, constraints and criteria from existing surface nuclear power plants to underground, and the use of many years of experience in the structural design of large underground cavities and cavity complexes for other purposes such as mining, hydropower stations etc. An application of such considerations in a recent input for the Underground Containment sub-section of the Seismic Task Group Report to the ASCE Committee for Nuclear Structures and Materials is presented as follows: underground concept considerations, siting criteria and structural selection, structural types, analytical and semi-analytical approaches, design and other miscellaneous considerations

  10. RA Reactor

    International Nuclear Information System (INIS)

    1989-01-01

    This chapter includes the following: General description of the RA reactor, organization of work, responsibilities of leadership and operators team, regulations concerning operation and behaviour in the reactor building, regulations for performing experiments, regulations and instructions for inserting samples into experimental channels [sr

  11. Identification and characterization of passive safety system and inherent safety feature building blocks for advanced light-water reactors

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1989-01-01

    Oak Ridge National Laboratory (ORNL) is investigating passive and inherent safety options for Advanced Light-Water Reactors (ALWRs). A major activity in 1989 includes identification and characterization of passive safety system and inherent safety feature building blocks, both existing and proposed, for ALWRs. Preliminary results of this work are reported herein. This activity is part of a larger effort by the US Department of Energy, reactor vendors, utilities, and others in the United States to develop improved LWRs. The Advanced Boiling Water Reactor (ABWR) program and the Advanced Pressurized Water Reactor (APWR) program have as goals improved, commercially available LWRs in the early 1990s. The Advanced Simplified Boiling Water Reactor (ASBWR) program and the AP-600 program are developing more advanced reactors with increased use of passive safety systems. It is planned that these reactors will become commercially available in the mid 1990s. The ORNL program is an exploratory research program for LWRs beyond the year 2000. Desired long-term goals for such reactors include: (1) use of only passive and inherent safety, (2) foolproof against operator errors, (3) malevolence resistance against internal sabotage and external assault and (4) walkaway safety. The acronym ''PRIME'' [Passive safety, Resilient operation, Inherent safety, Malevolence resistance, and Extended (walkaway) safety] is used to summarize these desired characteristics. Existing passive and inherent safety options are discussed in this document

  12. Seismic analysis of a NPP reactor building using spectrum-compatible power spectral density functions

    International Nuclear Information System (INIS)

    Venancio Filho, F.; DeCarvalho Santos, S.H.; Joia, L.A.

    1987-01-01

    A numerical methodology to obtain Power Spectral Density Functions (PSDF) of ground accelerations, compatible with a given design response spectrum is presented. The PSDF's are derived from the statistical analysis of the amplitudes of the frequency components in a set of artificially generated time-histories matching the given spectrum. A so obtained PSDF is then used in the stochastic analysis of a NPP Reactor Building. The main results of this analysis are compared with the ones obtained by deterministic methods

  13. Seismic analysis of a NPP reactor building using spectrum-compatible power spectral density functions

    International Nuclear Information System (INIS)

    Venancio Filho, F.; Joia, L.A.

    1987-01-01

    A numerical methodology to obtain Power Spectral Density Functions (PSDF) of ground accelerations, compatible with a given design response spectrum is presented. The PSDF's are derived from the statistical analysis of the amplitudes of the frequency components in a set of artificially generated time-histories matching the given spectrum. A so obtained PSDF is then used in the stochastic analysis of a reactor building. The main results of this analysis are compared with the ones obtained by deterministic methods. (orig./HP)

  14. The Performance of Structured Packings in Trickle-Bed Reactors

    NARCIS (Netherlands)

    Frank, M.J.W.; Kuipers, J.A.M.; Versteeg, G.F.; Swaaij, W.P.M. van

    1999-01-01

    An experimental study was carried out to investigate whether the use of structured packings might improve the mass transfer characteristics and the catalyst effectiveness of a trickle-bed reactor. Therefore, the performances of a structured packing, consisting of KATAPAK elements, and a dumped

  15. Nuclear reactor construction with bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1987-01-01

    This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

  16. Structural design of nuclear reactor machinery and equipment

    International Nuclear Information System (INIS)

    Hara, Hideki

    1992-01-01

    Since the machinery, equipment and piping which compose nuclear power station facilities are diverse, when those are designed, consideration is given sufficiently to the objective of use and the importance of the object machinery and equipment so that those can maintain the soundness over the design life. In this report, on the contents and the design standard in the design techniques for nuclear reactor machinery and equipment, the way of thinking is shown, taking an example of reactor pressure vessel which is stipulated as the vessel kind 1 in the 'Technical standard of structures and others regarding nuclear facilities for electric power generation', Notice No. 501 of the Ministry of International Trade and Industry. The reactor pressure vessel of 1350 MWe improved type BWR (ABWR) is used under the condition of 87.9 kg/cm 2 and 302 degC, and the inside diameter is about 7.2 m, the inside height is about 21 m, and the wall thickness is about 170 mm. The design standard for reactor pressure vessels and its way of thinking, breakdown prevention design and the design techniques for reactor pressure vessels are described. (K.I.)

  17. Structural wood products in onshore buildings at Naval Station Norfolk, 2000.

    Science.gov (United States)

    David B. McKeever

    2003-01-01

    As of December 31, 2000, there were 603 buildings at Naval Station (NAVSTA) Norfolk with a combined floor area of nearly 17.3 million ft2. In one-third of these buildings, structural wood products were used in one or more major structural building applications, utilizing an estimated 11.6 million board feet of lumber, 0.4 million ft2 (3/8-in. basis) of structural...

  18. INTRANS. A computer code for the non-linear structural response analysis of reactor internals under transient loads

    International Nuclear Information System (INIS)

    Ramani, D.T.

    1977-01-01

    The 'INTRANS' system is a general purpose computer code, designed to perform linear and non-linear structural stress and deflection analysis of impacting or non-impacting nuclear reactor internals components coupled with reactor vessel, shield building and external as well as internal gapped spring support system. This paper describes in general a unique computational procedure for evaluating the dynamic response of reactor internals, descretised as beam and lumped mass structural system and subjected to external transient loads such as seismic and LOCA time-history forces. The computational procedure is outlined in the INTRANS code, which computes component flexibilities of a discrete lumped mass planar model of reactor internals by idealising an assemblage of finite elements consisting of linear elastic beams with bending, torsional and shear stiffnesses interacted with external or internal linear as well as non-linear multi-gapped spring support system. The method of analysis is based on the displacement method and the code uses the fourth-order Runge-Kutta numerical integration technique as a basis for solution of dynamic equilibrium equations of motion for the system. During the computing process, the dynamic response of each lumped mass is calculated at specific instant of time using well-known step-by-step procedure. At any instant of time then, the transient dynamic motions of the system are held stationary and based on the predicted motions and internal forces of the previous instant. From which complete response at any time-step of interest may then be computed. Using this iterative process, the relationship between motions and internal forces is satisfied step by step throughout the time interval

  19. Radar Mapping of Building Structures Applying Sparse Reconstruction

    NARCIS (Netherlands)

    Tan, R.G.; Wit, J.J.M. de; Rossum, W.L. van

    2012-01-01

    The ability to map building structures at a certain stand-off distance allows intelligence, reconnaissance, and clearance tasks to be performed in a covert way by driving around a building. This will greatly improve security, response time, and reliability of aforementioned tasks. Therefore,

  20. An approach to build a knowledge base for reactor diagnostic system using statistical method

    International Nuclear Information System (INIS)

    Yokobayashi, Masao; Matsumoto, Kiyoshi; Kohsaka, Atsuo

    1988-01-01

    In the development of a rule-based expert system, one of the key issues is how to acquire knowledge and to build a knowledge base. When the knowledge base of DISKET was built, which is an expert system for nuclear reactor accident diagnosis developed in Japan Atomic Energy Research Institute, several problems have been experienced. To write rules is a time-consuming task, and it was difficult to keep the objectivity and consistency of rules as the number of rules increased. Certainty factors must be determined often according to engineering judgement, i.e. empirically or intuitively. A systematic approach was attempted to cope with these difficulties and to build efficiently an objective knowledge base. The approach described in this paper is based on the concept that a prototype knowledge base, colloquially speaking an initial guess, should first be generated in a systematic way, then it is modified or improved by human experts for practical use. Factor analysis was used as the systematic way. DISKET system, the procedure of building a knowledge base, and the verification of the approach are reported. (Kako, I.)

  1. Constructive systems, load-bearing and enclosing structures of high-rise buildings

    Science.gov (United States)

    Anatol'evna Korol', Elena; Olegovna Kustikova, Yuliya

    2018-03-01

    As the height of the building increases, loads on load-carrying structures increase dramatically, and as a result of the development of high-rise construction, several structural systems of such buildings have been developed: frame, frame-frame, cross-wall, barrel, box-type, box-to-wall ("pipe in pipe", "Trumpet in the farm"), etc. In turn, the barrel systems have their own versions: cantilever support of the ceilings on the trunk, suspension of the outer part of the overlap to the upper carrying console "hanging house" or its support by means of the walls on the lower bearing cantilever, intermediate position of the supporting cantilevers in height to the floor, from a part of floors. The object of the study are the structural solutions of high-rise buildings. The subject of the study is the layout of structural schemes of high-rise buildings, taking into account the main parameters - altitude (height), natural climatic conditions of construction, materials of structural elements and their physical and mechanical characteristics. The purpose of the study is to identify the features and systematization of structural systems of high-rise buildings and the corresponding structural elements. The results of the research make it possible, at the stage of making design decisions, to establish rational parameters for the correspondence between the structural systems of high-rise buildings and their individual elements.

  2. Ethical Guidelines for Structural Interventions to Small-Scale Historic Stone Masonry Buildings.

    Science.gov (United States)

    Hurol, Yonca; Yüceer, Hülya; Başarır, Hacer

    2015-12-01

    Structural interventions to historic stone masonry buildings require that both structural and heritage values be considered simultaneously. The absence of one of these value systems in implementation can be regarded as an unethical professional action. The research objective of this article is to prepare a guideline for ensuring ethical structural interventions to small-scale stone historic masonry buildings in the conservation areas of Northern Cyprus. The methodology covers an analysis of internationally accepted conservation documents and national laws related to the conservation of historic buildings, an analysis of building codes, especially Turkish building codes, which have been used in Northern Cyprus, and an analysis of the structural interventions introduced to a significant historic building in a semi-intact state in the walled city of Famagusta. This guideline covers issues related to whether buildings are intact or ruined, the presence of earthquake risk, the types of structural decisions in an architectural conservation project, and the values to consider during the decision making phase.

  3. An analysis of reactor structural response to fuel sodium interaction in a hypothetical core disruptive accident

    International Nuclear Information System (INIS)

    Suzuki, K.; Tashiro, M.; Sasanuma, K.; Nagashima, K.

    1976-01-01

    This study shows the effect of constraints around FSI zone on FSI phenomena and deformations of reactor structures. SUGAR-PISCES code system has been developed to evaluate the phenomena of FSI and the response of reactor structure. SUGAR calculates the phenomena of FSI. PISCES, developed by Physics International Company in U.S.A., calculates the dynamic response of reactor structure in two-dimensional, time-dependent finite-difference Lagrangian model. The results show that the peak pressure and energy by FSI and the deformation of reactor structures are about twice in case of FSI zone surrounding by blanket than by coolant. The FSI phenomena highly depend on the reactor structure and the realistic configuration around core must be considered for analyzing hypothetical core disruptive accident. This work was supported by a grant from Power Reactor and Nuclear Fuel Development Corporation. (auth.)

  4. The fast breeder reactor

    International Nuclear Information System (INIS)

    Davis, D.A.; Baker, M.A.W.; Hall, R.S.

    1990-01-01

    Following submission of written evidence, the Energy Committee members asked questions of three witnesses from the Central Electricity Generating Board and Nuclear Electric (which will be the government owned company running nuclear power stations after privatisation). Both questions and answers are reported verbatim. The points raised include where the responsibility for the future fast reactor programme should lie, with government only or with private enterprise or both and the viability of fast breeder reactors in the future. The case for the fast reactor was stated as essentially strategic not economic. This raised the issue of nuclear cost which has both a construction and a decommissioning element. There was considerable discussion as to the cost of building a European Fast reactor and the cost of the electricity it would generate compared with PWR type reactors. The likely demand for fast reactors will not arrive for 20-30 years and the need to build a fast reactor now is questioned. (UK)

  5. Soil Structure Interaction Effect on High Rise and Low Rise Buildings

    OpenAIRE

    Divya Pathak; PAresh H. SHAH

    2000-01-01

    Effect of supporting soil on the response of structure has been analyzed in the present study. A low rise (G+ 5 storey) and a high rise (G+12 storey) building has been taken for the analysis. For both type of buildings, the response of building with and without consideration of soil structure interaction effect has been compared.Without interaction case is the case in which ends of the structure are assumed to be fixed while in interaction case, structure is assumed to be...

  6. Influence of temperature on strain monitoring of degradation in concrete containment buildings

    International Nuclear Information System (INIS)

    Ding, Y.; Jaffer, S.; Angell, P.

    2015-01-01

    Concrete containment buildings (CCBs) are important safety structures in a nuclear power plant (NPP). The CCBs can be made of reinforced and post-tensioned (P-T) concrete. Post-tensioning concrete induces compressive stresses, which have to be overcome for the concrete to crack under tensile loads. However, post-tensioned CCBs may undergo pre-stressing losses as they age, which could affect their performance under accident conditions. CANDU 6 reactor buildings contain grouted post-tensioned tendons as the primary reinforcement. The grouting of the tendons makes direct monitoring of pre-stressing losses via lift-off testing impossible. Therefore, instruments have been installed on an existing reactor building to measure and monitor strains and stresses in the concrete and the deformation of the concrete structure to detect aging degradation and indirectly evaluate the pre-stressing losses. However, the instrumentation readings are affected by temporary volume changes in the concrete caused by the influence of environmental factors, particularly temperature, on concrete. In this work, the focus is on developing an understanding of the effect of temperature on the interpretation of instrumentation data from a reactor building. Vibrating Wire Strain Gauge (VWSG) data has been analysed. The influence of concrete coefficient of thermal expansion and temperature distribution within the reactor building walls, on VWSG data, is discussed based on the analysis of the available instrumentation data and available numerical simulation results. The present study demonstrates that temperature distribution within the containment concrete has a significant impact on the VWSG measurements and the coefficient of thermal expansion of concrete is an important factor in the correction of VWSG data for thermal strain. It is recommended that VWSG data obtained over small temperature variations be considered for interpretation to assess pre-stressing losses. (authors)

  7. Helium effect on mechanical property of fusion reactor structural materials

    International Nuclear Information System (INIS)

    Yamamoto, Norikazu; Chuto, Toshinori; Murase, Yoshiharu; Nakagawa, Johsei

    2004-01-01

    High-energy neutrons produced in fusion reactor core caused helium in the structural materials of fusion reactors, such as blankets. We injected alpha particles accelerated by the cyclotron to the samples of martensite steel (9Cr3WVTaB). Equivalent helium doses injected to the sample is estimated to be up to 300 ppm, which were estimated to be equivalent to helium accumulation after the 1-year reactor operation. Creep tests of the samples were made to investigate helium embrittlement. There were no appreciable changes in the relation between the stresses and the rupture time, the minimum creep rate and the applied stress. Grain boundary effect by helium was not observed in ruptured surfaces. Fatigue tests were made for SUS304 samples, which contain helium up to 150 ppm. After 0.05 Hz cyclic stress tests, it was shown that the fatigue lifetime (cycles to rupture and extension to failure) are 1/5 in 150 ppm helium samples compared with no helium samples. The experimental results suggest martensite steel is promising for structural materials of fusion reactors. (Y. Tanaka)

  8. Significance assessment of small-medium sized reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kanno, Minoru [Japan Atomic Power Co., Research and Development Dept., Tokyo (Japan)

    2002-12-01

    Preliminary assessment for deployment of small-medium sized reactor (S and M reactor) as a future option has been conducted at the JAPCO (Japan Atomic Power Company) under the cooperation with the CRIERI (Central Research Institute of Electric Power Industry). Significance of the S and M reactor introduction is listed as follows; lower investment cost, possible siting near demand side, enlarged freedom of siting, shorter transmission line, good compatibility with slow increase of demand and plain explanation of safety using simpler system such as integral type vessel without piping, natural convection core cooling and passive safety system. The deployment of simpler plant system, modular shop fabrication, ship-shell structured building and longer operation period can assure economics comparable with that of a large sized reactor, coping with scale-demerit. Also the S and M reactor is preferable in size for the nuclear heat utilization such as hydrogen production. (T. Tanaka)

  9. The energy-saving anaerobic baffled reactor membrane bioreactor (EABR-MBR) system for recycling wastewater from a high-rise building.

    Science.gov (United States)

    Ratanatamskul, Chavalit; Charoenphol, Chakraphan

    2015-01-01

    A novel energy-saving anaerobic baffled reactor-membrane bioreactor (EABR-MBR) system has been developed as a compact biological treatment system for reuse of water from a high-rise building. The anaerobic baffled reactor (ABR) compartment had five baffles and served as the anaerobic degradation zone, followed by the aerobic MBR compartment. The total operating hydraulic retention time (HRT) of the EABR-MBR system was 3 hours (2 hours for ABR compartment and very short HRT of 1 hour for aerobic MBR compartment). The wastewater came from the Charoen Wisawakam building. The results showed that treated effluent quality was quite good and highly promising for water reuse purposes. The average flux of the membrane was kept at 30 l/(m2h). The EABR-MBR system could remove chemical oxygen demand, total nitrogen and total phosphorus from building wastewater by more than 90%. Moreover, it was found that phosphorus concentration was rising in the ABR compartment due to the phosphorus release phenomenon, and then the concentration decreased rapidly in the aerobic MBR compartment due to the phosphorus uptake phenomenon. This implies that phosphorus-accumulating organisms inside the EABR-MBR system are responsible for biological phosphorus removal. The research suggests that the EABR-MBR system can be a promising system for water reuse and reclamation for high-rise building application in the near future.

  10. A study on nonlinear behavior of reactor containment structures during ultimate accident condition(I)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sun Hoon; Kim, Young Jin; Park, Joo Yeon [Youngdong Univ., Yeongdong (Korea, Republic of)] (and others)

    2003-03-15

    In this study, the following scope and contents are established for first year's study of determining ultimate pressure capacity of CANDU-type reactor containment. State-of-arts on the prediction of the ultimate pressure capacity of prestressed concrete reactor containment. Comparative study on structural characteristics and analysis model of CANDU-type reactor containment. State-of-arts on evaluation method of the ultimate pressure capacity of prestressed concrete reactor containment. Enhancement of evaluation method of the ultimate pressure capacity for PWR containment structure. In order to determine a realistic lower bound of a typical reactor containment structural capacity for internal pressure, modelling techniques and analytical investigation to predict its non-linear behavior up to ultimate capacity are required. Especially, the in-depth evaluation of modeling technique and analysis procedure for determining ultimate pressure capacity of CANDU-type reactor containment is required. Therefore, modelling techniques and analytical investigation to predict its non-linear behavior up to ultimate pressure capacity of CANDU-type reactor containment for internal pressure will be suggested in this study.

  11. A study on nonlinear behavior of reactor containment structures during ultimate accident condition(I)

    International Nuclear Information System (INIS)

    Kim, Sun Hoon; Kim, Young Jin; Park, Joo Yeon

    2003-03-01

    In this study, the following scope and contents are established for first year's study of determining ultimate pressure capacity of CANDU-type reactor containment. State-of-arts on the prediction of the ultimate pressure capacity of prestressed concrete reactor containment. Comparative study on structural characteristics and analysis model of CANDU-type reactor containment. State-of-arts on evaluation method of the ultimate pressure capacity of prestressed concrete reactor containment. Enhancement of evaluation method of the ultimate pressure capacity for PWR containment structure. In order to determine a realistic lower bound of a typical reactor containment structural capacity for internal pressure, modelling techniques and analytical investigation to predict its non-linear behavior up to ultimate capacity are required. Especially, the in-depth evaluation of modeling technique and analysis procedure for determining ultimate pressure capacity of CANDU-type reactor containment is required. Therefore, modelling techniques and analytical investigation to predict its non-linear behavior up to ultimate pressure capacity of CANDU-type reactor containment for internal pressure will be suggested in this study

  12. Adoption of Smart Structures for Prevention of Health Hazards in Buildings

    Science.gov (United States)

    Oke, Ayodeji; Aigbavboa, Clinton; Ngema, Wiseman

    2017-11-01

    The importance of building quality to the health and well-being of occupants and surrounding neighbors cannot be overemphasized. Smart structures were construed to proffer solution to various issues of sustainable development including social factors that is concerned with health and safety of people. Based on existing literature materials on building quality, smart structures and general aspect of sustainable developments, this study examined the benefits of smart structures in the prevention of various health issues in infrastructural buildings, which has been a concern for stakeholders in the architecture, engineering and construction industry. The criterion for indoor environmental quality was adopted and various health and bodily issues related to building quality were explained. The adoption of smart structure concept will help to manage physical, chemical, biological and psychological factors of building with a view to enhancing better quality of life of occupants.

  13. Deployable bamboo structure project: A building life-cycle report

    Science.gov (United States)

    Firdaus, Adrian; Prastyatama, Budianastas; Sagara, Altho; Wirabuana, Revian N.

    2017-11-01

    Bamboo is considered as a sustainable material in the world of construction, and it is vastly available in Indonesia. The general utilization of the material is increasingly frequent, however, its usage as a deployable structure-a recently-developed use of bamboo, is still untapped. This paper presents a report on a deployable bamboo structure project, covering the entire building life-cycle phase. The cycle encompasses the designing; fabrication; transportation; construction; operation and maintenance; as well as a plan for future re-use. The building is made of a configuration of the structural module, each being a folding set of bars which could be reduced in size to fit into vehicles for easy transportation. Each structural module was made of Gigantochloa apus bamboo. The fabrication, transportation, and construction phase require by a minimum of three workers. The fabrication and construction phase require three hours and fifteen minutes respectively. The building is utilized as cafeteria stands, the operation and maintenance phase started since early March 2017. The maintenance plan is scheduled on a monthly basis, focusing on the inspection of the locking mechanism element and the entire structural integrity. The building is designed to allow disassembly process so that it is reusable in the future.

  14. Limit the effects of secondary circuit water or steam piping breaks in the reactor building

    International Nuclear Information System (INIS)

    Nachev, N.

    2001-01-01

    The existing design of the WWER-1000 Model 320 does not include provisions against the local mechanical effects of pipe ruptures of the secondary system piping. This situation may lead to accidental effects beyond the design basis of the plant in case of a postulated secondary pipe rupture event. The aim of the present safety enhancement measure is to overcome this safety deficit, that means to carry out some analyses and to suggest protection measures, by which the specified design basis of the plant concerning secondary circuit design basis accidents will be assured. The systems to be considered include the main steam lines (MSL) and the main feedwater lines (MFWL) in the safety related system areas. These areas are the system portions, which are located in the reactor building (containment and room A820 outside the containment). The pipe rupture effects to be considered include the local effects, that means pipe whip impact and jet forces on the adjacent equipment and structures, as well as reaction forces due to blowdown thrust forces and pressure waves in the broken piping system. (author)

  15. Specificities of micro-structured reactors for hydrogen production and purification

    Energy Technology Data Exchange (ETDEWEB)

    Dupont, N.; Germani, G.; Van Veen, A.C.; Schuurman, Y.; Mirodatos, C. [Institut de Recherches sur la Catalyse - CNRS, 2, Avenue Albert Einstein, 69626 Villeurbanne Cedex (France); Schaefer, G. [Atotech Deutschland GmbH, PO Box 210780, 10507 Berlin (Germany)

    2007-07-15

    This paper presents the specificities of micro-structured reactors as compared to conventional fixed-bed reactors through two case studies devoted to (i) hydrogen production by methanol steam reforming, (ii) hydrogen purification by water-gas shift (WGS). Key features like catalyst coating stability, temperature and pressure management, effects of operating conditions (residence time, pressure drops, etc.) are well identified as controlling the micro-reactor performances for methanol reforming. These devices are also shown to be excellent tools for fast access to reaction kinetics as exemplified for the WGS reaction, subject to operating conditions carefully chosen to ensure proper hydrodynamics, in order to use conventional plug flow reactor models for extracting rate constants. (author)

  16. Earthquake-proof support structures for the recycling pump in FBR type reactors

    International Nuclear Information System (INIS)

    Nakagawa, Masaki; Shigeta, Masayuki.

    1984-01-01

    Purpose: To improve the earthquake proofness of the recycling pump for use in FBR type reactors upon earthquake by reducing the vibration response of the pump. Constitution: The outer casing of a recycle pump suspended into liquid sodium is extended to the portion that penetrates a reactor core support structures. Support structures surrounding the outer side of the recycling pump are disposed with a gap not restraining the free thermal deformations of the recycling pump to the inside of the partition wall structures and the portion of the recycling pump penetrating the reator core support structures, to integrate the support structures with the reactor core support structures. Accordingly, there are no interferences between the recycling pump and the support structures with respect to the thermal deformations that change gradually with time. Upon vibrating under the rapidly changing external forces of earthquakes, however, the pressure resulted to the liquid in the gap due to the vibrations of the recycling pump is transmitted with no escape to the support structures, the recycling pump and the support structures integrally resist the vibrations thereby enabling to reduce the vibrations in the recycling pumps. (Horiuchi, T.)

  17. Composites as structural materials in fusion reactors

    International Nuclear Information System (INIS)

    Megusar, J.

    1989-01-01

    In fusion reactors, materials are used under extreme conditions of temperature, stress, irradiation, and chemical environment. The absence of adequate materials will seriously impede the development of fusion reactors and might ultimately be one of the major difficulties. Some of the current materials problems can be solved by proper design features. For others, the solution will have to rely on materials development. A parallel and balanced effort between the research in plasma physics and fusion-related technology and in materials research is, therefore, the best strategy to ultimately achieve economic, safe, and environmentally acceptable fusion. The essential steps in developing composites for structural components of fusion reactors include optimization of mechanical properties followed by testing under fusion-reactor-relevant conditions. In optimizing the mechanical behavior of composite materials, a wealth of experience can be drawn from the research on ceramic matrix and metal matrix composite materials sponsored by the Department of Defense. The particular aspects of this research relevant to fusion materials development are methodology of the composite materials design and studies of new processing routes to develop composite materials with specific properties. Most notable examples are the synthesis of fibers, coatings, and ceramic materials in their final shapes form polymeric precursors and the infiltration of fibrous preforms by molten metals

  18. Calculation of DPA in the Reactor Internal Structural Materials of Nuclear Power Plant

    International Nuclear Information System (INIS)

    Kim, Yong Deong; Lee, Hwan Soo

    2014-01-01

    The embrittlement is mainly caused by atomic displacement damage due to irradiations with neutrons, especially fast neutrons. The integrity of the reactor internal structural materials has to be ensured over the reactor life time, threatened by the irradiation induced displacement damage. Accurate modeling and prediction of the displacement damage is a first step to evaluate the integrity of the reactor internal structural materials. Traditional approaches for analyzing the displacement damage of the materials have relied on tradition model, developed initially for simple metals, Kinchin and Pease (K-P), and the standard formulation of it by Norgett et al. , often referred to as the 'NRT' model. An alternative and complementary strategy for calculating the displacement damage is to use MCNP code. MCNP uses detailed physics and continuous-energy cross-section data in its simulations. In this paper, we have performed the evaluation of the displacement damage of the reactor internal structural materials in Kori NPP unit 1 using detailed Monte Carlo modeling and compared with predictions results of displacement damage using the classical NRT model. The evaluation of the displacement damage of the reactor internal structural materials in Kori NPP unit 1 using detailed Monte Carlo modeling has been performed. The maximum value of the DPA rate was occurred at the baffle side of the reactor internal where the node has the maximum neutron flux

  19. Structural assessments of plate type support system for APR1400 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, Anh Tung; Namgung, Ihn, E-mail: inamgung@kings.ac.kr

    2017-04-01

    Highlights: • This paper investigates plate-type support structure for the reactor vessel of the APR 1400. • The tall column supports of APR1400 reactor challenges in seismic and severe accident events. • A plate-type support of reactor vessel was proposed and evaluated based on ASME code. • The plate-type support was assessed to show its higher rigidity than column-type. - Abstract: This paper investigates an alternative form of support structure for the reactor vessel of the APR 1400. The current reactor vessel adopts a four-column support arrangement locating on the cold legs of the vessel. Although having been successfully designed, the tall column structure challenges in seismic events. In addition, for the mitigation of severe accident consequences, the columns inhibit ex-vessel coolant flow, hence the elimination of the support columns proposes extra safety advantages. A plate-type support was proposed and evaluated for the adequacy of meeting the structural stiffness by Finite Element Analysis (FEA) approach. ASME Boiler and Pressure Vessel Code was used to verify the design. The results, which cover thermal and static structural analysis, show stresses are within allowable limits in accordance with the design code. Even the heat conduction area is increased for the plate-type of support system, the results showed that the thermal stresses are within allowable limits. A comparison of natural frequencies and mode shapes for column support and plate-type support were presented as well which showed higher fundamental frequencies for the plate-type support system resulting in greater rigidity of the support system. From the outcome of this research, the plate-type support is proven to be an alternative to current APR column type support design.

  20. Mitigate Strategy of Very High Temperature Reactor Air-ingress Accident

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Tae Kyu [KHNP CRI, Daejeon (Korea, Republic of); Arcilesi, David J.; Sun, Xiaodong; Christensen, Richard N. [The Ohio State University, Columbus (United States); Oh, Chang H.; Kim, Eung S. [Idaho National Laboratory, Idaho (United States)

    2016-10-15

    A critical safety event of the Very High Temperature Reactor (VHTR) is a loss-of-coolant accident (LOCA). Since a VHTR uses graphite as a core structure, if there is a break on the pressure vessel, the air in the reactor cavity could ingress into the reactor core. The worst case scenario of the accident is initiated by a double-ended guillotine break of the cross vessel that connects the reactor vessel and the power conversion unit. The operating pressures in the vessel and containment are about 7 and 0.1 MPa, respectively. In the VHTR, the reactor pressure vessel is located within a reactor cavity which is filled with air during normal operation. Therefore, the air-helium mixture in the cavity may ingress into the reactor pressure vessel after the depressurization process. In this paper, a commercial computational fluid dynamics (CFD) tool, FLUENT, was used to figure out air-ingress mitigation strategies in the gas-turbine modular helium reactor (GT-MHR) designed by General Atomics, Inc. After depressurization, there is almost no air in the reactor cavity; however, the air could flow back to the reactor cavity since the reactor cavity is placed in the lowest place in the reactor building. The heavier air could flow to the reactor cavity through free surface areas in the reactor building. Therefore, Argon gas injection in the reactor cavity is introduced. The injected argon would prevent the flow by pressurizing the reactor cavity initially, and eventually it prevents the flow by making the gas a heavier density than air in the reactor cavity. The gate opens when the reactor cavity is pressurized during the depressurization and it closes by gravity when the depressurization is terminated so that it can slow down the air flow to the reactor cavity.

  1. Study on layout and construction concept of DMS (modular simplified medium small reactor)

    International Nuclear Information System (INIS)

    Shizuka Hirako; Yuusuke Shimizu; Shigeru Yokouchi; Yoshinori Iimura; Yuuji Yasuda; Kumiaki Moriya; Takahiko Hida

    2005-01-01

    Nuclear power is expected to become the main source of electric power generation in Japan for reasons of energy security and prevention of CO 2 emissions. In addition, the recent slowdown of electric power demand and the liberalization of the electric power market are accelerating medium and small sized reactor development. Under these circumstances, DMS's (modular simplified and medium small reactors) have been developed as 400 MWe class LWR's supported by the Japan Atomic Power Company. In the development of medium and small sized reactors, the most important point is how to overcome the scale demerits. To this end, we have pursued not only the simplification of systems and equipment but also the standardization of layout and construction. The main technical feature of DMS's is the adoption of a natural circulation reactor with short length fuel. Short length fuel enables the reduction of RPV height as well as construction volume of the PCV and building volume. A natural circulation reactor has considerable rationalizing effects such as the elimination of re-circulation pumps and their drive power source. By applying simplified systems and equipment, a rationalized layout and construction method are adopted. To improve the constructability by means of modular construction methods, steel containment is applied. The PCV size is reduced to 17 m in diameter and 24 m in height by applying a dish-shaped drywell and eccentric RPV arrangement. By applying a compact PCV and concentrated equipment arrangement in building, it can be confirmed that the ratio of building volume per unit power is equivalent to that of existing large sized ABWRs. Furthermore, a steel plate reinforced concrete structure (SC structure) is applied to the building layout. The application of the compact PCV (steel containment) and the SC structure makes it easier to apply a large-scale module, such as an integrated steel containment and SC structure module, and an integrated multi-layer BM (building

  2. Data Quality Objective Summary Report for Phase II of the 105-F and DR Reactor Buildings

    International Nuclear Information System (INIS)

    Bauer, R.G.

    1998-01-01

    This data quality objective (DQO) process is to support planning and decision-making activities of Phase II decontamination and decommissioning (D and D) activities for the 105-F and 105-DR Reactor Buildings.The objective of this DQO is to determine the survey and characterization requirements for these rooms to provide the necessary information for worker safety, waste designation, recycle, reuse, and clean landfill disposal decisions during D and D

  3. The SPHINX reactor for engineering tests

    International Nuclear Information System (INIS)

    Adamov, E.O.; Artamkin, K.N.; Bovin, A.P.; Bulkin, Y.M.; Kartashev, E.F.; Korneev, A.A.; Stenbok, I.A.; Terekhov, A.S.; Khmel'Shehikov, V.V.; Cherkashov, Y.M.

    1990-01-01

    A research reactor known as SPHINX is under development in the USSR. The reactor will be used mainly to carry out tests on mock-up power reactor fuel assemblies under close-to-normal parameters in experimental loop channels installed in the core and reflector of the reactor, as well as to test samples of structural materials in ampoule and loop channels. The SPHINX reactor is a channel-type reactor with light-water coolant and moderator. Maximum achievable neutron flux density in the experimental channels (cell composition 50% Fe, 50% H 2 O) is 1.1 X 10 15 neutrons/cm 2 · s for fast neutrons (E > 0.1 MeV) and 1.7 X 10 15 for thermal neutrons at a reactor power of 200 MW. The design concepts used represent a further development of the technical features which have met with approval in the MR and MIR channel-type engineering test reactors currently in use in the USSR. The 'in-pond channel' construction makes the facility flexible and eases the carrying out of experimental work while keeping discharges of radioactivity into the environment to a low level. The reactor and all associated buildings and constructions conform to modern radiation safety and environmental protection requirements

  4. Homogenization of the internal structures of a reactor with the cooling fluid

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F. [CEA Saclay, SEMT, 91 - Gif sur Yvette (France); Bliard, F. [Socotec Industrie, Service AME, 78 - Montigny le Bretonneux (France)

    2001-07-01

    To take into account the influence of a structure net among a fluid flow, without modelling exactly the structure shape, a concept of ''equivalent porosity method'' was developed. The structures are considered as solid pores inside the fluid. The structure presence is represented by three parameters: a porosity, a shape coefficient and a pressure loss coefficient. The method was studied for an Hypothetical Core Disruptive Accident in a Liquid Metal Fast Breeder Reactor, but it can be applied to any problem involving fluid flow getting through a solid net. The model was implemented in the computer code CASTEM-PLEXUS and validated on an analytical shock tube test, simulating an horizontal slice of a schematic LMFBR in case of a HCDA (bubble at high pressure, liquid sodium and internal structures of the reactor). A short parametric study shows the influence of the porosity and the structure shape on the pressure wave impacting the shock tube bottom. These results were used to simulate numerically the HCDA mechanical effects in a small scale reactor mock-up. (author)

  5. Homogenization of the internal structures of a reactor with the cooling fluid

    International Nuclear Information System (INIS)

    Robbe, M.F.; Bliard, F.

    2001-01-01

    To take into account the influence of a structure net among a fluid flow, without modelling exactly the structure shape, a concept of ''equivalent porosity method'' was developed. The structures are considered as solid pores inside the fluid. The structure presence is represented by three parameters: a porosity, a shape coefficient and a pressure loss coefficient. The method was studied for an Hypothetical Core Disruptive Accident in a Liquid Metal Fast Breeder Reactor, but it can be applied to any problem involving fluid flow getting through a solid net. The model was implemented in the computer code CASTEM-PLEXUS and validated on an analytical shock tube test, simulating an horizontal slice of a schematic LMFBR in case of a HCDA (bubble at high pressure, liquid sodium and internal structures of the reactor). A short parametric study shows the influence of the porosity and the structure shape on the pressure wave impacting the shock tube bottom. These results were used to simulate numerically the HCDA mechanical effects in a small scale reactor mock-up. (author)

  6. Artificial intelligence applications in fixed area monitor for TRIGA reactor building and service building

    International Nuclear Information System (INIS)

    Talpalariu, C.; Talpalariu, J.; Vaja, N.; Matei, C.

    2008-01-01

    This system is intended for the protection of personnel working in those areas of the Reactor Building and Service Building where high gamma radiation fields are expected. A detector, sensitive to gamma radiation, is installed in each of the areas to be monitored. The detector will send a signal, proportional to the radiation level in the area, to a corresponding electronic module (Alarm Unit), where the signal will be amplified and checked against alarm set points for possible alarming conditions. In case the field exceeds the alarm set values, the Alarm Unit will produce a signal that will trigger the field alarms (Horn and Beacon) located in the area where the condition occurred. Each Alarm Unit will send a numerical input to central computer command. he system is required to accomplish the following tasks: - Monitors the level of gamma radiation in those areas of the Station where high radiation fields are expected; - Provides a continuous and centralized display of the radiation level in each of the monitored areas. The display shall be in exposure rate units (R/h); - Provides a visual and audible alarm in each monitored areas; Allows the control room operator to check at any time the radiation levels and alarm conditions in each of the monitored areas; - Control room operator shall be alerted of any alarm conditions that occurs in the Station. A typical monitoring loop is composed of the following components: Detector Assembly type: CI-MA - 522 two channels, two ranges; Horn and Beacon Assembly; Remote Indicating Meter with Warning Lights; Central computer; common equipment for all 40 loops. (authors)

  7. Building with electromagnetic shield structure for individual floors

    International Nuclear Information System (INIS)

    Takahashi, T.; Nakamura, M.; Yabana, Y.; Ishikawa, T.; Nagata, K.

    1991-01-01

    This invention relates to a building having a floor-by-floor electromagnetic shield structure well-suited for application to an information network system in which an electromagnetically shielded space is divided by individual floors and electric waves are utilized within the building on a floor-by-floor basis. (author). 8 figs

  8. Building with electromagnetic shield structure for individual floors

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, T; Nakamura, M; Yabana, Y; Ishikawa, T; Nagata, K

    1991-09-10

    This invention relates to a building having a floor-by-floor electromagnetic shield structure well-suited for application to an information network system in which an electromagnetically shielded space is divided by individual floors and electric waves are utilized within the building on a floor-by-floor basis. (author). 8 figs.

  9. Renewal of reactor cooling system of JMTR. Reactor building site

    International Nuclear Information System (INIS)

    Onoue, Ryuji; Kawamata, Takanori; Otsuka, Kaoru; Sekine, Katsunori; Koike, Sumio; Gorai, Shigeru; Nishiyama, Yutaka; Fukasaku, Akitomi

    2012-03-01

    The Japan Materials Testing Reactor (JMTR) is a light water moderated and cooled tank-type reactor, and its thermal power is 50 MW. The JMTR is categorized as high flux testing reactors in the world. The JMTR has been utilized for irradiation experiments of nuclear fuels and materials, as well as for radioisotope productions since the first criticality in March 1968 until August 2006. JAEA is decided to refurbish the JMTR as an important fundamental infrastructure to promote the nuclear research and development. And The JMTR refurbishment work is carried out for 4 years from 2007. Before refurbishment work, from August 2006 to March 2007, all concerned renewal facilities were selected from evaluation on their damage and wear in terms of aging. Facilities which replacement parts are no longer manufactured or not likely to be manufactured continuously in near future, are selected as renewal ones. Replace priority was decided with special attention to safety concerns. A monitoring of aging condition by the regular maintenance activity is an important factor in selection of continuous using after the restart. In this report, renewal of the cooling system within refurbishment facilities in the JMTR is summarized. (author)

  10. Probabilistic Assessment of Structural Seismic Damage for Buildings in Mid-America

    International Nuclear Information System (INIS)

    Bai, Jong-Wha; Hueste, Mary Beth D.; Gardoni, Paolo

    2008-01-01

    This paper provides an approach to conduct a probabilistic assessment of structural damage due to seismic events with an application to typical building structures in Mid-America. The developed methodology includes modified damage state classifications based on the ATC-13 and ATC-38 damage states and the ATC-38 database of building damage. Damage factors are assigned to each damage state to quantify structural damage as a percentage of structural replacement cost. To account for the inherent uncertainties, these factors are expressed as random variables with a Beta distribution. A set of fragility curves, quantifying the structural vulnerability of a building, is mapped onto the developed methodology to determine the expected structural damage. The total structural damage factor for a given seismic intensity is then calculated using a probabilistic approach. Prediction and confidence bands are also constructed to account for the prevailing uncertainties. The expected seismic structural damage is assessed for a typical building structure in the Mid-America region using the developed methodology. The developed methodology provides a transparent procedure, where the structural damage factors can be updated as additional seismic damage data becomes available

  11. Flow induced vibrational excitation of nuclear reactor structures

    International Nuclear Information System (INIS)

    Gibert, R.J.

    1979-01-01

    The pressure fluctuations generated by disturbed flows, encountered in nuclear reactors induce vibrations in the structures. In order to make forecastings for these vibrational levels, it is necessary to know the characteristics of the random pressure fluctuations induced in the walls by the main flow peculiarities of the circuits. This knowledge is essentially provided by experimentation which shows that most of the energy from these fluctuations is in the low frequency area. It is also necessary to determine the transfer functions of the fluid-structure coupled system. Given the frequency range of the excitations, a calculation of the characteristics of the first eigenmodes is generally sufficient. This calculation is carried out by finite element codes, the modal dampings being assessed separately. In this paper, emphasis is placed mainly on the analysis of the sources of excitation due to flow peculiarities. Some examples will also be given of assessments of vibrations in real structures (pipes, reactor internals, etc.) and of comparisons with the experimental results obtained on models or on a site [fr

  12. Trends in large-scale testing of reactor structures

    International Nuclear Information System (INIS)

    Blejwas, T.E.

    2003-01-01

    Large-scale tests of reactor structures have been conducted at Sandia National Laboratories since the late 1970s. This paper describes a number of different large-scale impact tests, pressurization tests of models of containment structures, and thermal-pressure tests of models of reactor pressure vessels. The advantages of large-scale testing are evident, but cost, in particular limits its use. As computer models have grown in size, such as number of degrees of freedom, the advent of computer graphics has made possible very realistic representation of results - results that may not accurately represent reality. A necessary condition to avoiding this pitfall is the validation of the analytical methods and underlying physical representations. Ironically, the immensely larger computer models sometimes increase the need for large-scale testing, because the modeling is applied to increasing more complex structural systems and/or more complex physical phenomena. Unfortunately, the cost of large-scale tests is a disadvantage that will likely severely limit similar testing in the future. International collaborations may provide the best mechanism for funding future programs with large-scale tests. (author)

  13. Effect of structural design on traffic-induced building vibrations

    DEFF Research Database (Denmark)

    Persson, Peter; Andersen, Lars Vabbersgaard; Persson, Kent

    2017-01-01

    Population growth and urbanization results in densified cities, where new buildings are being built closer to existing vibration sources such as road-, tram- and rail traffic. In addition, new transportation systems are constructed closer to existing buildings. Potential disturbing vibrations...... are one issue to consider in planning urban environment and densification of cities. Vibrations can be disturbing for humans but also for sensitive equipment in, for example, hospitals. In determining the risk for disturbing vibrations, the distance between the source and the receiver, the ground...... properties, and type and size of the building are governing factors. In the paper, a study is presented aiming at investigating the influence of various parameters of the building's structural design on vibration levels in the structure caused by ground surface loads, e.g. traffic. Parameters studied...

  14. CRBR reactor structures design. BRC meeting presentation

    International Nuclear Information System (INIS)

    Pennell, W.E.

    1975-01-01

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

  15. Structural strength during severe reactor accidents of the VVER- 91 nuclear power plant

    International Nuclear Information System (INIS)

    Varpasuo, P.

    1999-12-01

    The report summarises the studies carried out in Fortum Engineering (formerly IVO Power Engineering) between the years of 1992 and 1997 concerning ultimate strength of structures designed to mitigate and contain the consequences of various core melt accident scenarios. The report begins with the description of containment loading situations arising from core melt accidents. These situations are divided to fast and slow loads. Fast loads include ex-vessel steam explosions, steam spikes, hydrogen burns, direct containment heating and missiles. Slow loads are connected with pressure rise inside the containment in case when the containment heat removal system is not functioning. First part of report describes the analyses of reactor cavity based on axi-symmetric load assumptions. These studies are performed with various models like one degree of freedom idealisation, axi-symmetric modelling of geometry and full three-dimensional modelling of geometry. Second part of report describes the analyses of cavity based on non-axi-symmetric load assumptions. Here full 3D- geometry model is used combined with various physical models for the behaviour of reinforced concrete. Third part of report gives short account of the analysis of containment ultimate pressure capacity. The containment model in this case includes pre-stressing tendons and mild steel reinforcing bars. The load is assumed to axi-symmetric internal static pressure. The capacity of the reactor cavity against the ex-vessel steam explosion scenarios for VVER-91 plant concept is established for both axi-symmetric and non-axi-symmetric load models using ANACAP structural analysis code. The validation of the cavity response to ex-vessel steam explosion load using different commercially available codes gave mixed results for both axisymmetric and non-axi-symmetric load presentations.The ultimate static overpressure capacity of the VVER-91 reactor cavity structure was established to be of the order of 10 MPa. This result

  16. Study on dynamic characteristics of reduced analytical model for PWR reactor internal structures

    International Nuclear Information System (INIS)

    Yoo, Bong; Lee, Jae Han; Kim, Jong Bum; Koo, Kyeong Hoe

    1993-01-01

    The objective of this study is to establish the procedure of the reduced analytical modeling technique for the PWR reactor internal(RI) structures and to carry out the sensitivity study of the dynamic characteristics of the structures by varying the structural parameters such as the stiffness, the mass and the damping. Modeling techniques for the PWR reactor internal structures and computer programs used for the dynamic analysis of the reactor internal structures are briefly investigated. Among the many components of RI structures, the dynamic characteristics for CSB was performed. The sensitivity analysis of the dynamic characteristics for the reduced analytical model considering the variations of the stiffnesses for the lower and upper flanges of the CSB and for the RV Snubber were performed to improve the dynamic characteristics of the RI structures against the external loadings given. In order to enhance the structural design margin of the RI components, the nonlinear time history analyses were attempted for the RI reduced models to compare the structural responses between the reference model and the modified one. (Author)

  17. Reanalysis and evaluation of seismic response of reactor building

    International Nuclear Information System (INIS)

    Li Zhongcheng; Li Zhongxian

    2005-01-01

    For the Ling Ao phase-I (LA-I) Nuclear Power Plant (NPP), its' seismic analysis of nuclear island was in accordance with the approaches in RCC-G standard for the model M310 in France, in which the Simplified impedance method was employed for the consideration of SSI. Thanks to the rapid progress being made in upgrading the evaluation technology and the capability of data processing systems, methods and software tools for the SSI analysis have experienced significant development all over the world. Focused on the model of reactor building of the LA-I NPP, in this paper the more sophisticated 3D half-space continuum impedance method based on the Green functions is used to analyze the functions of the soil, and then the seismic responses of the coupled SSI system are calculated and compared with the corresponding design values. It demonstrates that the design method provides a set of conservatively safe results. The conclusions from the study are hopefully to provide some important references to the assessment of seismic safety margin for LA-I NPP. (authors)

  18. Development of radiation resistant structural materials utilizing fission research reactors in Japan (Role of research reactors)

    International Nuclear Information System (INIS)

    Shikama, T.; Tanigawa, H.; Nozawa, T.; Muroga, T.; Aoyama, T.; Kawamura, H.; Ishihara, M.; Ito, C.; Kaneda, S.; Mimura, S.

    2009-01-01

    Structural materials for next-generation nuclear power systems should have a good radiation resistance, where the expected accumulation dose will largely exceed 10 dpa. Among several candidate materials, materials of five categories, 1. Austenitic steels, including high nickel alloys, 2. Low activation ferritic martensitic steels, 3. ODS steels (austenitic and ferritic), 4. Vanadium based alloys, 5. Silicon carbide composites (SiC/SiCf). All have been most extensively studied in Japan, in collaboration among industries, national institutes such as Japan Atomic Energy Agency (JAEA), National Institute for Fusion Science (NIFS) and National Institute for Materials Science (NIMS), and universities. The high nickel base alloys were studied for their low swelling behaviors mainly by the NIMS and the austenitic steels are studied for their reliable engineering data base and their reliable performance in irradiation environments mainly by the JAEA, mainly for their application in the near-term projects such as the ITER and the Sodium Cooled Fast Reactors. The most extensive studies are now concentrated on the Low Activation Ferritic Marsensitic steels and ODS steels, for their application in a demonstration fusion reactor and prototype sodium cooled fast reactors. Fundamental studies on radiation effects are carried out, mainly utilizing Japan Materials Testing Rector (JMTR) with its flexible irradiation ability, up to a few dpa. For higher dpa irradiation, a fast test reactor, JOYO is utilized up to several 10s dpa. Some international collaborations such as Japan/USA and Japan/France are effective to utilize reactors abroad, such as High Flux Isotope Reactor (HFIR) of Oak Ridge National Laboratory, and sodium cooled high flux fast reactors in France. Silicon carbide based composites are extensively studied by university groups led by Kyoto University and the JAEA. For their performance in heavy irradiation environments, the Japan/USA collaboration plays an important role

  19. Fundamental Understanding of Crack Growth in Structural Components of Generation IV Supercritical Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Iouri I. Balachov; Takao Kobayashi; Francis Tanzella; Indira Jayaweera; Palitha Jayaweera; Petri Kinnunen; Martin Bojinov; Timo Saario

    2004-11-17

    This work contributes to the design of safe and economical Generation-IV Super-Critical Water Reactors (SCWRs) by providing a basis for selecting structural materials to ensure the functionality of in-vessel components during the entire service life. During the second year of the project, we completed electrochemical characterization of the oxide film properties and investigation of crack initiation and propagation for candidate structural materials steels under supercritical conditions. We ranked candidate alloys against their susceptibility to environmentally assisted degradation based on the in situ data measure with an SRI-designed controlled distance electrochemistry (CDE) arrangement. A correlation between measurable oxide film properties and susceptibility of austenitic steels to environmentally assisted degradation was observed experimentally. One of the major practical results of the present work is the experimentally proven ability of the economical CDE technique to supply in situ data for ranking candidate structural materials for Generation-IV SCRs. A potential use of the CDE arrangement developed ar SRI for building in situ sensors monitoring water chemistry in the heat transport circuit of Generation-IV SCWRs was evaluated and proved to be feasible.

  20. The seismic response and floor spectra of OL3 NPP buildings in Finland

    International Nuclear Information System (INIS)

    Pentti Varpasuo

    2005-01-01

    The purpose of the present work is the computation of seismic response and floor spectra of the nuclear power plant OL3 buildings in Olkiluoto. The following OL3 plant buildings were included in the analysis: 1. the Reactor Building UJA/UJB; 2. the Safeguard Buildings UJH/UJK 1-4; 3. and the Fuel Building UFA The in-structure spectra were generated using the ground motion response spectra documented in YVL GUIDE 2.6 'Seismic events at nuclear power plants' issued by Finnish Centre of Radiation Protection. The floor spectra were computed for the following equipment damping values: 2%, 4%, 7%, and 10%. The joint model for the plant buildings was generated. All analyses were linear and the direct time integration method was used with time step of 0.001 sec. All response runs were carried out with MSC/Nastran general purpose structural analysis program. The development of floor spectra has been carried out in accordance with the US NRC -Regulatory Guide 1.122: 'Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components'. The response results show that the dominant frequencies of the reactor building are located around 5 Hz in frequency space and that the typical amplification of spectral peaks for 4% damping is from 8 -10 times when compared to peak ground acceleration. (authors)

  1. Micro structured reactors for synthesis/decomposition of hazardous chemicals. Challenging prospects for micro structured reaction architectures (4)

    NARCIS (Netherlands)

    Rebrov, E.V.; Croon, de M.H.J.M.; Schouten, J.C.

    2004-01-01

    A review. This paper completes a series of four publications dealing with the different aspects of the applications of micro reactor technol. This article focuses on the application of micro structured reactors in the processes for synthesis/decompn. of hazardous chems., such as unsym.

  2. Decentralized Networked Control of Building Structures

    Czech Academy of Sciences Publication Activity Database

    Bakule, Lubomír; Rehák, Branislav; Papík, Martin

    2016-01-01

    Roč. 31, č. 11 (2016), s. 871-886 ISSN 1093-9687 R&D Projects: GA ČR GA13-02149S Institutional support: RVO:67985556 Keywords : decentralized control * networked control * building structures Subject RIV: BC - Control Systems Theory Impact factor: 5.786, year: 2016

  3. Structural materials for Gen-IV nuclear reactors: Challenges and opportunities

    Science.gov (United States)

    Murty, K. L.; Charit, I.

    2008-12-01

    Generation-IV reactor design concepts envisioned thus far cater toward a common goal of providing safer, longer lasting, proliferation-resistant and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-IV reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses and extremely corrosive environment, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This paper presents a summary of various Gen-IV reactor concepts, with emphasis on the structural materials issues depending on the specific application areas. This paper also discusses the challenges involved in using the existing materials under both service and off-normal conditions. Tasks become increasingly complex due to the operation of various fundamental phenomena like radiation-induced segregation, radiation-enhanced diffusion, precipitation, interactions between impurity elements and radiation-produced defects, swelling, helium generation and so forth. Further, high temperature capability (e.g. creep properties) of these materials is a critical, performance-limiting factor. It is demonstrated that novel alloy and microstructural design approaches coupled with new materials processing and fabrication techniques may mitigate the challenges, and the optimum system performance may be achieved under much demanding conditions.

  4. Main features of buildings and structures important to safety of units V1 and V2 of Bohunice NPP

    International Nuclear Information System (INIS)

    David, M.

    1993-01-01

    The program of seismic upgrading of Bohunice NPPs has been started in the year 1989 (after finishing of new seismic input). Since that time the seismic upgrading of Main building of NPP V1 has already been realized, structural as well as technological parts. Beside that the designs of seismic upgrading of other structures of NPP V1 and V2 have been completed. It has been proved that the seismic upgrading of NPPs with reactors WWER 440 is very complicated, but still possible, even in the case with high seismic intensity. It would be not possible to fulfill this complicated task without the help of IAEA Missions. The activities of IAEA experts in the program of Bohunice NPPs upgrading are appreciated very much

  5. Surface modification method for reactor incore structural component

    International Nuclear Information System (INIS)

    Obata, Minoru; Sudo, Akira.

    1996-01-01

    A large number of metal or ceramic small spheres accelerated by pressurized air are collided against a surface of a reactor incore structures or a welded surface of the structural components, and then finishing is applied by polishing to form compression stresses on the surface. This can change residual stresses into compressive stress without increasing the strength of the surface. Accordingly, stress corrosion crackings of the incore structural components or welded portions thereof can be prevented thereby enabling to extend the working life of equipments. (T.M.)

  6. Proposal for the use of new materials in the TOKAMAK building cover; Contrato de ingenieria/arquitectura para el proyecto ITER

    Energy Technology Data Exchange (ETDEWEB)

    Chiva, L.

    2011-07-01

    It was considered relevant and innovative to apply new structural materials to the construction of the roof of the building that lodged the TOKAMAK reactor, with the aim of achieving a severe reduction of the weight of the roof structure that result in greater ease of mounting, minor charges on the walls and foundations of the building and a reduced impact on the distribution of masses of the building scheme.

  7. Synthesis of vibration control and health monitoring of building structures under unknown excitation

    International Nuclear Information System (INIS)

    He, Jia; Huang, Qin; Xu, You-Lin

    2014-01-01

    The vibration control and health monitoring of building structures have been actively investigated in recent years but often treated separately according to the primary objective pursued. In this study, a time-domain integrated vibration control and health monitoring approach is proposed based on the extended Kalman filter (EKF) for identifying the physical parameters of the controlled building structures without the knowledge of the external excitation. The physical parameters and state vectors of the building structure are then estimated and used for the determination of the control force for the purpose of the vibration attenuation. The interaction between the health monitoring and vibration control is revealed and assessed. The feasibility and reliability of the proposed approach is numerically demonstrated via a five-story shear building structure equipped with magneto-rheological (MR) dampers. Two types of excitations are considered: (1) the EI-Centro ground excitation underneath of the building and (2) a swept-frequency excitation applied on the top floor of the building. Results show that the structural parameters as well as the unknown dynamic loadings could be identified accurately; and, at the same time, the structural vibration is significantly reduced in the building structure. (paper)

  8. KEY ASPECTS OF ENSURING ENERGY EFFICIENCY OF BUILDINGS AND STRUCTURES

    Directory of Open Access Journals (Sweden)

    S.G. Abramyan

    2017-06-01

    Full Text Available The paper is based on the review of the foreign and national academic literature and intended to emphasize the issues of ensuring energy efficiency of buildings and structures applicable to all the countries as for reconstruction of existing buildings as for erection of new ones . The author highlights the key aspects of the provision of energy efficiency of buildings and structures in some foreign countries. The conclusion is made that the studies are mainly aimed at discovering new heat insulation materials, whereby polystyrene insulation is found to be the most widespread wall insulation material in a number of countries. At the same time, it is observed that the ongoing research is focused on solutions to optimize the structure of walling systems in terms of both insulant thickness and the number and sequence of insulation layers in the walling structure. A conclusion is made that hyper insulation of external walls leads to considerable expenses arising due to cooling during the summer season. The use of prefabricated vacuum panels as a heat insulation layer and off-the-shelf single-layer structures, subject to their heat insulation characteristics, appears a more constructive way to meet the energy efficiency requirements, as the arrangement of ideal air space in multilayered walls proves a significant challenge today. One of the most promising ways to ensure energy efficiency is the use of multifunctional polyvalent walls and provision of polyvalent heat supply from renewable energy sources. Since energy efficiency depends on the spatial arrangement of buildings, construction must ensure a minimum ratio of the area of enclosing structures to the overall building volume (by adding on new facilities in case of reconstruction. It is noted that a systemic approach to ensuring energy efficiency of buildings is impossible without proper regard to the environmental parameters of heat insulation materials.

  9. Reliability of redundant structures of nuclear reactor protection systems

    International Nuclear Information System (INIS)

    Vojnovic, B.

    1983-01-01

    In this paper, reliability of various redundant structures of PWR protection systems has been analysed. Structures of reactor tip systems as well as the systems for activation of safety devices have been presented. In all those systems redundancy is achieved by means of so called majority voting logic ('r out of n' structures). Different redundant devices have been compared, concerning probability of occurrence of safe as well as unsafe failures. (author)

  10. The fundamentals of structural building codes

    NARCIS (Netherlands)

    Vrouwenvelder, A.C.W.M.

    2001-01-01

    Partial Factor Design is nowadays a generally accepted design method for building and civil engineering structures. For most engineers the general philosophy that the safety factors depend on the type of the load and on the limit state under consideration makes sense. However, the background, in

  11. Damage by radiation in structural materials of BWR reactor vessels

    International Nuclear Information System (INIS)

    Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

    2002-01-01

    The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA Mark III Salazar reactor and separately with Ni +3 ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A 2 ). (Author)

  12. Earthquake response of nuclear reactor building deeply embedded in soil

    International Nuclear Information System (INIS)

    Masao, T.; Hirasawa, M.; Yamamoto, S.; Koori, Y.

    1977-01-01

    Regarding the earthquake response of nuclear reactor building embedded in soil, experimental and theoretical investigations has been performed on a model of height-3.75 meter, bottom cross section-5x5 meter, weight-173 ton made of conrete under the financial support of Japanese government (The Science and Technology Agency). The top of model was excited by an eccentric mass vibration that can generate maximum force of 3 tons. Earthpressures were measured at the bottom and side wall of model, and displacements were also measured at the top of model (6 components) and ground surface changed in the steps which were 0, 20, 47, 73, 100% (against the height of model). Using these experimental results and soil properties, dynamical characteristics were studied, including resonant frequency, radiation damping, vibrational mode, frequency response and earthpressure distribution around the model at varying embedment by lumped model, cyclindrical elastic wave model and FEM models (thin layer elements). (Auth.)

  13. A study on ex-vessel steam explosion for a flooded reactor cavity of reactor scale - 15216

    International Nuclear Information System (INIS)

    Song, S.; Yoon, E.; Kim, Y.; Cho, Y.

    2015-01-01

    A steam explosion can occur when a molten corium is mixed with a coolant, more volatile liquid. In severe accidents, corium can come into contact with coolant either when it flows to the bottom of the reactor vessel and encounters the reactor coolant, or when it breaches the reactor vessel and flows into the reactor containment. A steam explosion could then threaten the containment structures, such as the reactor vessel or the concrete walls/penetrations of the containment building. This study is to understand the shortcomings of the existing analysis code (TEXAS-V) and to estimate the steam explosion loads on reactor scale and assess the effect of variables, then we compared results and physical phenomena. Sensitivity study of major parameters for initial condition is performed. Variables related to melt corium such as corium temperature, falling velocity and diameter of melt are more important to the ex-vessel steam explosion load and the steam explosion loads are proportional to these variables related to melt corium. Coolant temperature on reactor cavity has a specific area to increase the steam explosion loads. These results will be used to evaluate the steam explosion loads using ROAAM (Risk Oriented Accident Analysis Methodology) and to develop the evaluation methodology of ex-vessel steam explosion. (authors)

  14. Progress in the development of the blanket structural material for fusion reactors

    International Nuclear Information System (INIS)

    Scott, J.L.; Bloom, E.E.; Grossbeck, M.L.; Maziasz, P.J.; Wiffen, F.W.; Gold, R.E.; Holmes, J.J.; Reuther, P.C. Jr.; Rosenwasser, S.N.

    1981-01-01

    The Alloy Development for Irradiation Performance Program has become more focused since the last Fusion Reactor Technology Conference two years ago. Since austenitic stainless steels and ferritic steels are candidate structural materials for the near-term reactors ETF and INTOR and austenitic stainless steel is also the preferred structural material for the steady-state commercial fusion reactor, STARFIRE, a vigorous experimental program is under way to identify the best alloy from each of these alloy classes and to provide the engineering data base in a timely manner. In addition the comprehensive program that includes high-strength Fe-Ni-Cr alloys, reactive and refractory metals, and advanced concepts continues in an orderly fashion

  15. Improved nuclear reactor construction with bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1987-01-01

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

  16. Structural performance of a graphite blanket in fusion reactors

    International Nuclear Information System (INIS)

    Wolfer, W.G.; Watson, R.D.

    1978-01-01

    Irradiation of graphite in a fusion reactor causes dimensional changes, enhanced creep, and changes in elastic properties and fracture strength. Temperature and flux gradients through the graphite blanket structure produce differential distortions and stress gradients. An inelastic stress analysis procedure is described which treats these variations of the graphite properties in a consistent manner as dictated by physical models for the radiation effects. Furthermore, the procedure follows the evolution of the stress and fracture strength distributions during the reactor operation as well as for possible shutdowns at any time. The lifetime of the graphite structure can be determined based on the failure criterion that the stress at any location exceeds one-half of the fracture strength. This procedure is applied to the most critical component of the blanket module in the SOLASE design

  17. Report of the reactor Operators Service - Annex F

    International Nuclear Information System (INIS)

    Zivotic, Z.

    1992-01-01

    RA reactor operators service is organized in two groups: permanent staff (chief operator, chief shift operators and operators) and changeable group which is formed according to the particular operation needs for working in shifts. For continuous training of the existing operator staff the Service has prepared and published eleven booklets: Nuclear reactor; RA reactor primary coolant loop; System for purification of heavy water; reactor helium system; system for technical water; electric power system; control and operation; ventilation system in the reactor building; special sewage system; construction properties of the reactor core; reactor building and installations. During the reporting period there have been no accidents nor incidents that could affect the reactor personnel [sr

  18. Seismic evaluation and strengthening of Bohunice nuclear power plant structures

    International Nuclear Information System (INIS)

    Shipp, J.G.; Short, S.A.; Grief, T.; Borov, V.; Kuzma, J.

    2001-01-01

    A seismic assessment and strengthening investigation is being performed for selected structures at the Bohunice V1 Nuclear Power Plant in Slovakia. Structures covered in this paper include the reactor building complex and the emergency generator station. The emergency generator station is emphasized in the paper as work is nearly complete while work on the reactor building complex is ongoing at this time. Seismic evaluation and strengthening work is being performed by a cooperative effort of Siemens and EQE along with local contractors. Seismic input is the interim Review Level Earthquake (horizontal peak ground acceleration of 0.3 g). The Bohunice V1 reactor building complex is a WWER 4401230 nuclear power plant that was originally built in the mid-1970s but had extensive seismic upgrades in 1991. Siemens has performed three dimensional dynamic analyses of the reactor building complex to develop seismic demand in structural elements. EQE is assessing seismic capacities of structural elements and developing strengthening schemes, where needed. Based on recent seismic response analyses for the interim Review Level Earthquake which account for soil-structure interaction in a rigorous manner, the 1991 seismic upgrade has been found to be inadequate in both member/connection strength and in providing complete load paths to the foundation. Additional strengthening is being developed. The emergency generator station was built in the 1970s and is a two-story unreinforced brick masonry (URM) shear wall building above grade with a one story reinforced concrete shear wall basement below grade. Seismic analyses and testing of the URM walls has been performed to assess the need for building strengthening. Required structural strengthening for in-plane forces consists of revised and additional vertical steel framing and connections, stiffening of horizontal roof bracing, and steel connections between the roof and supporting walls and pointing of two interior transverse URM

  19. Reactors licensing: proposal of an integrated quality and environment regulatory structure for nuclear research reactors in Brazil

    International Nuclear Information System (INIS)

    Serra, Reynaldo Cavalcanti

    2014-01-01

    A new integrated regulatory structure based on quality and integrated issues has been proposed to be implemented on the licensing process of nuclear research reactors in Brazil. The study starts with a literature review about the licensing process in several countries, all of them members of the International Atomic Energy Agency. After this phase it is performed a comparative study with the Brazilian licensing process to identify good practices (positive aspects), the gaps on it and to propose an approach of an integrated quality and environmental management system, in order to contribute with a new licensing process scheme in Brazil. The literature review considered the following research nuclear reactors: Jules-Horowitz and OSIRIS (France), Hanaro (Korea), Maples 1 and 2 (Canada), OPAL (Australia), Pallas (Holand), ETRR-2 (Egypt) and IEA-R1 (Brazil). The current nuclear research reactors licensing process in Brazil is conducted by two regulatory bodies: the Brazilian National Nuclear Energy Commission (CNEN) and the Brazilian Institute of Environment and Renewable Natural Resources (IBAMA). CNEN is responsible by nuclear issues, while IBAMA by environmental one. To support the study it was applied a questionnaire and interviews based on the current regulatory structure to four nuclear research reactors in Brazil. Nowadays, the nuclear research reactor’s licensing process, in Brazil, has six phases and the environmental licensing process has three phases. A correlation study among these phases leads to a proposal of a new quality and environmental integrated licensing structure with four harmonized phases, hence reducing potential delays in this process. (author)

  20. Assessment of Technogenic Accident Risk of Industrial Building Structures

    Science.gov (United States)

    Baiburin, D. A.; Baiburin, A. Kh

    2017-11-01

    A methodology for assessing the risk of an industrial building accident was developed taking into account the damage caused by various localization of collapse. Before the beginning of the survey of a facility technical condition, groups including the same type of building structures are selected. Further, assessment is made for the reduction in their load-carrying capacity from the strength and stability conditions taking into account defects. The characteristics of the influence of defects and structural damage on a building safety is the degree of compliance with the standards expressed by the reliability level. Reliability levels assignment is carried out on the basis of calculations, operating experience and inspection of a particular type of structure according to the formalized rules. The risk of collapse according to a separate scenario is calculated for structures that are capable and incapable of causing a progressive ossification. The results of the technique application are based on the analysis of the accident risk at the welding shop “Vysota (Height) 239” of the Chelyabinsk Pipe Rolling Plant.

  1. Nuclear reactor PBMR and cogeneration; Reactor nuclear PBMR y cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Alonso V, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    In recent years the nuclear reactor designs for the electricity generation have increased their costs, so that at the moment costs are managed of around the 5000 US D for installed kw, reason for which a big nuclear plant requires of investments of the order of billions of dollars, the designed reactors as modular of low power seek to lighten the initial investment of a big reactor dividing the power in parts and dividing in modules the components to lower the production costs, this way it can begin to build a module and finished this to build other, differing the long term investment, getting less risk therefore in the investment. On the other hand the reactors of low power can be very useful in regions where is difficult to have access to the electric net being able to take advantage of the thermal energy of the reactor to feed other processes like the water desalination or the vapor generation for the processes industry like the petrochemical, or even more the possible hydrogen production to be used as fuel. In this work the possibility to generate vapor of high quality for the petrochemical industry is described using a spheres bed reactor of high temperature. (Author)

  2. On Directionality of Phrase Structure Building

    Science.gov (United States)

    Chesi, Cristiano

    2015-01-01

    Minimalism in grammatical theorizing (Chomsky in "The minimalist program." MIT Press, Cambridge, 1995) led to simpler linguistic devices and a better focalization of the core properties of the structure building engine: a lexicon and a free (recursive) phrase formation operation, dubbed Merge, are the basic components that serve in…

  3. RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1986

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1986-01-01

    In order to enable future reliable operation of the RA reactor, according to new licensing regulations, three major tasks started in 1984 were fulfilled: building of the new emergency system, reconstruction of the existing ventilation system, and reconstruction of the power supply system. Simultaneously in 1985/1986 renewal of the instrumentation and reconstruction of the system for handling and storage of the spent fuel in the reactor building have started. Design projects for these tasks are almost finished and the reconstruction of both systems is expected to be finished until 1988 and mid 1989 respectively. RA reactor Safety report was finished according to the recommendations of the IAEA. Investments in 1986 were used for 8000 kg of heavy water, maintenance of reactor systems and supply of new components, reconstruction of reactor systems. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr

  4. A model for structural analysis of nuclear reactor pressure vessel flanges

    International Nuclear Information System (INIS)

    Oliveira, C.A. de.

    1987-01-01

    Due to the recent Brazilian advances in the nuclear technology area, it has been necessary the development of design and analysis methods for pressurized water reactor components, also as other components of a nuclear plant. This work proposes a methodology for the structural analysis of large diameter nuclear reactor pressure vessel flanges. In the analysis the vessel is divided into shell-of-revolution elements, the flanges are represented by rigid rings, and the bolts are treated as beams. The flexibility method is used for solving the problem. A computer program is shown, and the given results (displacements and stresses) are compared with results obtained by the finite element method. Although developed for nuclear reactor pressure vessel calculations, the program is more general, being possible its use for the analysis of any structure composed by shells of revolution. (author)

  5. Structural integrity analysis of an INPP building under external loading

    International Nuclear Information System (INIS)

    Dundulis, G.; Karalevicius, R.; Uspuras, E.; Kulak, R.F.; Marchertas, A.

    2005-01-01

    After the terrorist attacks in New York and Washington D. C. using civil airplanes, the evaluation of civil airplane crashes into civil and NPP structures has become very important. The interceptions of many terrorists' communications reveal that the use of commandeered commercial aircraft is still a major part of their plans for destruction. Aircraft crash or other flying objects in the territory of the Ignalina Nuclear Power Plant (INPP) represents a concern to the plant. Aircraft traveling at high velocity have a destructive potential. The aircraft crash may damage the roof and walls of buildings, pipelines, electric motors, cases of power supplies, power cables of electricity transmission and other elements and systems, which are important for safety. Therefore, the evaluation of the structural response to an of aircraft crash is important and was selected for analysis. The structural integrity analysis due to the effects of an aircraft crash on an NPP building structure is the subject of this paper. The finite element method was used for the structural analysis of a typical Ignalina NPP building. The structural integrity analysis was performed for a portion of the ALS using the dynamic loading of an aircraft crash impact model. The computer code NEPTUNE was used for this analysis. The local effects caused by impact of the aircraft's engine on the building wall were evaluated independently by using an empirical formula. (authors)

  6. Adaptive building skin structures

    International Nuclear Information System (INIS)

    Del Grosso, A E; Basso, P

    2010-01-01

    The concept of adaptive and morphing structures has gained considerable attention in the recent years in many fields of engineering. In civil engineering very few practical applications are reported to date however. Non-conventional structural concepts like deployable, inflatable and morphing structures may indeed provide innovative solutions to some of the problems that the construction industry is being called to face. To give some examples, searches for low-energy consumption or even energy-harvesting green buildings are amongst such problems. This paper first presents a review of the above problems and technologies, which shows how the solution to these problems requires a multidisciplinary approach, involving the integration of architectural and engineering disciplines. The discussion continues with the presentation of a possible application of two adaptive and dynamically morphing structures which are proposed for the realization of an acoustic envelope. The core of the two applications is the use of a novel optimization process which leads the search for optimal solutions by means of an evolutionary technique while the compatibility of the resulting configurations of the adaptive envelope is ensured by the virtual force density method

  7. Topics to be covered in safety analysis reports for nuclear power plants with pressurized water reactors or boiling water reactors in the F.R.G

    International Nuclear Information System (INIS)

    Kohler, H.A.G.

    1977-01-01

    This manual aims at defining the standards to be used in Safety Analysis Reports for Nuclear Power Plants with Pressurized Water Reactors or Boiling Water Reactors in the Federal Republic of Germany. The topics to be covered are: Information about the site (geographic situation, settlement, industrial and military facilities, transport and communications, meteorological conditions, geological, hydrological and seismic conditions, radiological background), description of the power plant (building structures, safety vessel, reactor core, cooling system, ventilation systems, steam power plant, electrical facilities, systems for measurement and control), indication of operation (commissioning, operation, safety measures, radiation monitoring, organization), incident analysis (reactivity incidents, loss-of-coolant incidents, external impacts). (HP) [de

  8. Efforts onto nuclear research and development such as new reactor and so forth

    International Nuclear Information System (INIS)

    Onishi, Tuneji

    2000-01-01

    The Japan Atomic Power Co. which is one of specified business company on nuclear power generation, has carried out construction and operation of power plants with different types of reactor such as boiling light water reactor (BWR), pressurized light water rector (PWR), and so forth. And, by actively using technical powers and experiences accumulated before then, additional construction of a new power unit, and researches and developments on a simplified light water reactor, a future type rector, and a high breeder proof reactor have been made some efforts. Here were introduced some outlines on development of an improved type PWR, development of a new type reactor for example, deep embedded plant), future type reactor (for example, revolutionary middle and small type reactor, simplified PWR, and simplified BWR), a fast breeder reactor, and a reactor building suitable for a ship shell structure. (G.K.)

  9. Structural design of the turbine building of Angra Nuclear Power Station, Unit 1

    International Nuclear Information System (INIS)

    Varella, L.N.; Reis, F.J.C.; Jurkiewicz, W.J.

    1978-01-01

    The Turbine Building of the Angra Nuclear Power Plant, Unit 1, and particularly its structure and structural design are described. The Turbine Building, as far as its structure is concerned, deviates from the standard structure of any turbine building due to the fact that huge ducts are provided in the foundation mat as to accomodate the circulating water system. This aspect and the fact that the building is founded upon a very deep strata of compacted and controlled fill, makes out of the building structure 'a concrete ship floating in the sea of sand', and by the same reason presents by itself an interesting structure, worth to be known to all engineers involved in design of power plants. This pape, suplemented by a few slides shown during presentation of the paper at the conference, covers the subject mainly from the designers' point of view. (Author)

  10. Transactions of the 10th international conference on structural mechanics in reactor technology

    International Nuclear Information System (INIS)

    Hadjian, A.H.

    1989-01-01

    This book covers all aspects of engineering mechanics pertaining to mechanical and structural components and the relevant systems in nuclear reactors. Subjects covered include: theoretical developments in structural mechanics, loading conditions, behavior of materials, fluid mechanics, operating experience, accident sequences, and calculational procedures. Problems of structural mechanics analysis are focused within the general context of the design, reliability, and safety of nuclear reactors. Operating plant performance and life extension, waste repository technology and regulatory research have been formalized as distinct Divisions

  11. Reactors set for mini market

    International Nuclear Information System (INIS)

    Knox, Richard.

    1988-01-01

    Commercial nuclear power generation on a large-scale has an uncertain future. However, it is hoped that a small nuclear reactor could form the basis for providing heating, cooling or electricity in large buildings. Based on the Slowpoke research reactor, the Slowpoke energy system concept is simple. The concept and the way in which the small-scale reactor would work are discussed. The system consists of a stainless steel tank surrounded by reinforced concrete and let into the ground. The tank is full of light water which is heated to about 90 deg C by a central core of 2.4 percent enriched uranium fuel. The resulting natural circulation causes the water to pass through a heat exchanger. The water from the heat exchanger can be used for building or district heating, to operate air-conditioners or to generate small quantities of electricity. It is hoped to automate the operation of the reactor so that continuous supervision by a team of engineers would be unnecessary. A single operator on call in the building would be able to take control actions if the reactor's safety system failed. (UK)

  12. Calculation methods of Structure-Soil-Structure Interaction (3SI) for embedded buildings: Application to NUPEC tests

    International Nuclear Information System (INIS)

    Clouteau, D.; Broc, D.; Devesa, G.; Guyonvarh, V.; Massin, P.

    2012-01-01

    This work aims at improving and validating methods coupling Finite Element (FE) and Boundary Element (BE) Methods in the context of Soil-Structure Interaction (SSI) and Structure-Soil-Structure Interaction (3SI) tests performed by NUPEC on mock-up structures built on an unmade ground. Several cases have been tested: single and juxtaposed buildings, shallow and embedded foundations, with various loading conditions: forced and natural seismic loadings. The numerical simulations of forced vibration tests are in good agreement with the results of the NUPEC experiments in the case of two embedded buildings either in terms of amplitude and resonance. The numerical simulation of seismic response tests by FEM and BEM allows for a proper choice of the 'reference point' where the computed and the experimental displacements coincide. A parametric analysis of Structure-Soil-Structure Interaction carried out by the FEM has allowed to determine the influence of some parameters on SSI. Most of them like the position of the building in the excavation, the direction of the load, the quality of the contact between the sidewalls of the buildings and the soil for embedded foundations, do not show to have a strong influence on the dynamic system behaviour, which is mainly governed by the stiffness of the first soil layer. As far as 3SI is concerned, this paper shows that when the cross interaction has a small effect on the building response in the case of surface foundations, it has a strong influence in the case of embedded foundations with an important decrease of the response at the top of the buildings. (authors)

  13. Full scale dynamic testing of Kozloduy NPP unit 5 structures

    International Nuclear Information System (INIS)

    Da Rin, E.M.

    1999-01-01

    As described in this report, the Kozloduy NPP western site has been subjected to low level earthquake-like ground shaking - through appropriately devised underground explosions - and the resulting dynamic response of the NPP reactor Unit 5 important structures appropriately measured and digitally recorded. In-situ free-field response was measured concurrently more than 100 m aside the main structures of interest. The collected experimental data provide reference information on the actual dynamic characteristics of the Kozloduy NPPs main structures, as well as give some useful indications on the dynamic soil-structure interaction effects for the case of low level excitation. Performing the present full-scale dynamic structural testing activities took advantage of the experience gained by ISMES during similar tests, lately performed in Italy and abroad (in particular, at the Paks NPP in 1994). The IAEA promoted dynamic testing of the Kozloduy NPP Unit 5 by means of pertinently designed buried explosion-induced ground motions which has provided a large amount of data on the dynamic structural response of its major structures. In the present report, the conducted investigation is described and the acquired digital data presented. A series of preliminary analyses were undertaken for examining in detail the ground excitation levels that were produced by these weak earthquake simulation experiments, as well as for inferring some structural characteristics and behaviour information from the collected data. These analyses ascertained the high quality of the collected digital data. Presumably due to soil-structure dynamic interaction effects, reduced excitation levels were observed at the reactor building foundation raft level with respect to the concurrent free-field ground motions. measured at a 140 m distance from the reactor building centre. Further more detailed and systematic analyses are worthwhile to be performed for extracting more complete information about the

  14. Economic evaluation of small modular nuclear reactors and the complications of regulatory fee structures

    International Nuclear Information System (INIS)

    Vegel, Benjamin; Quinn, Jason C.

    2017-01-01

    Carbon emission concerns and volatility in fossil fuel resources have renewed world-wide interest in nuclear energy as a solution to growing energy demands. Several large nuclear reactors are currently under construction in the United States, representing the first new construction in over 30 years. Small Modular Reactors (SMRs) have been in design for many years and offer potential technical and economic advantages compared with traditionally larger reactors. Current SMR capital and operational expenses have a wide range of uncertainty. This work evaluates the potential for SMRs in the US, develops a robust techno-economic assessment of SMRs, and leverages the model to evaluate US regulatory fees structures. Modeling includes capital expenses of a factory facility and capital and operational expenses with multiple scenarios explored through a component-level capital cost model. Policy regarding the licensing and regulation of SMRs is under development with proposed annual US regulatory fees evaluated through the developed techno-economic model. Results show regulatory fees are a potential barrier to the economic viability of SMRs with an alternate fee structure proposed and evaluated. The proposed fee structure is based on the re-distribution of fees for all nuclear reactors under a single structure based on reactor thermal power rating. - Highlights: • Potential demand for new small modular nuclear power in the US is established. • Capital costs are broken down on component level and include factory production. • US regulatory fees structures are evaluated, results show potential barrier. • An additional fee structure is proposed and compared with current US fee structures.

  15. Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the Technical Meeting is to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials

  16. Dynamic soil-structure interactions on embedded buildings

    International Nuclear Information System (INIS)

    Kobarg, J.; Werkle, H.; Henseleit, O.

    1983-01-01

    The dynamic soil-structure interaction on the horizontal seismic excitation is investigated on two typical embedded auxiliary buildings of a nuclear power plant. The structure and the soil are modelled by various analytical and numerical methods. Under the condition of the linear viscoelastic theory, i.e. soil characteristic constant in time and independent of strain, the interaction influences between a homogenous soil layer and a structure are analysied for the following parameters: 4) mathematical soil modells; 4) mathematical structure modells; 4) shear wave velocities; 3) embedment conditions; 4) earthquake time histories. (orig.) [de

  17. Thermal performance of an insulating structure for a reactor vessel

    International Nuclear Information System (INIS)

    Aranovitch, E.; Crutzen, S.; LeDet, M.; Denis, R.

    This report describes the installations used to test the HTGR reactor vessel insulating structure called ''Casali'' and details the experimental results in 3 groups: general experiments, systematic study, and technological experiments. The results obtained make it possible to satisfactorily predict the behavior of the structure in a practical application

  18. Extent of moisture and mould damage in structures of public buildings

    Directory of Open Access Journals (Sweden)

    Petri J. Annila

    2017-06-01

    Full Text Available The study concentrated on the extent of moisture and mould damage in different structures in 25 public buildings in Finland. Users of all the buildings had health symptoms suspected to be the result of moisture and mould damage, which is why moisture performance assessments had been performed. The assessment reports on each building were available as research material. The reports indicated that the examined buildings suffered from multiple moisture and mould problems in several different structures. On average, however, a relatively small proportion of the total number of structures had suffered damage. On the basis of the research material, damage was most extensive in walls in soil contact (16.3% and base floor structures (12.5%. The lowest damage rates were found in partition walls (2.4%, external walls (2.6% and intermediate floors (2.5%. The results of the study underline the importance of thorough moisture performance assessments to ensure that all point-sized moisture and mould damage is detected.

  19. Waste generated by the future decommissioning of the Magurele VVR-S Research Reactor

    International Nuclear Information System (INIS)

    Dragolici, F.; Turcanu, C.N.; Dragolici, A.C.

    2001-01-01

    Nuclear Research Reactor WWR-S from the National Institute of Research and Development for Physics and Nuclear Engineering 'Horia Hulubei', Bucharest-Magurele, was commissioned in July 1957 and it was shut down in December 1997. At the moment the reactor is in conservation state. During its operation this reactor worked at an average power of 2MW, almost 3216 h/year, producing a total thermal power of 230 x 10 3 MWh. No major modifications or improvements were made during the 40 years of operation to the essential parts of the reactor, respective to the primary cooling system, reactor vessel, active core and electronic devices. So, all components of the measure, control and protection systems are old, generally at the technical level of the 1950s, therefore a reason why in December 1997 the operation was ceased. At present, the reactor can be considered, by IAEA definition in the first stage (reactor shut down, but the vital functions are maintained and monitored). The survey is related to the second stage - restrictive use of the area. To develop a real decommissioning project, it was first necessary to evaluate the volume and the characteristics of the radioactive waste which will be generated. Radioactive waste generated during the decommissioning of Magurele WR-S research reactor may be classified as: Activated wastes (internal structures, horizontal channels and thermal column, biological shield); Contaminated wastes (primary circuit non-activated components, hot cells, some technological rooms as main hall, pumps room, radioactive material transfer areas, ventilation building and stack); Possibly contaminated materials from any area of reactor building and ventilation building. After 40 years of nuclear research activities, all such areas are suspected of contamination. The volume of wastes that will result from WWR-S Research Reactor decommissioning is summarized

  20. Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the TM on “Liquid metal reactor concept: core design and structural materials” was to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials. Main results arising from national and international R&D programmes and projects in the field were reviewed, and new activities to be carried out under the IAEA aegis were identified on the basis of the analysis of current research and technology gaps

  1. Estimation of release of tritium from measurements of air concentrations in reactor building of PHWR

    International Nuclear Information System (INIS)

    Purohit, R.G.; Sarkar, P.K.

    2010-01-01

    In this paper an attempt has been made to estimate the releases from measured air concentrations of tritium at various locations in Reactor Building (RB). Design data of Kaiga Generating Station and sample measurements of tritium concentrations at various locations in RB and discharges for a period of fortnight were used. A comparison has also been made with actual measurements. It has been observed that there is good matching in estimated and actual measurements of tritium release on some days while on some days there is high difference

  2. The French nuclear power plant reactor building containment contributions of prestressing and concrete performances in reliability improvements and cost savings

    International Nuclear Information System (INIS)

    Rouelle, P.; Roy, F.

    1998-01-01

    The Electricite de France's N4 CHOOZ B nuclear power plant, two units of the world's largest PWR model (1450 Mwe each), has earned the Electric Power International's 1997 Powerplant Award. This lead NPP for EDF's N4 series has been improved notably in terms of civil works. The presentation will focus on the Reactor Building's inner containment wall which is one of the main civil structures on a technical and safety point of view. In order to take into account the necessary evolution of the concrete technical specification such as compressive strength low creep and shrinkage, the HSC/HPC has been used on the last N4 Civaux 2 NPP. As a result of the use of this type of professional concrete, the containment withstands an higher internal pressure related to severe accident and ensures higher level of leak-tightness, thus improving the overall safety of the NPP. On that occasion, a new type of prestressing has been tested locally through 55 C 15 S tendons using a new C 1500 FE Jack. These updated civil works techniques shall allow EDF to ensure a Reactor Containment lifespan for more than 50 years. The gains in terms of reliability and cost saving of these improved techniques will be developed hereafter

  3. Structural analysis of the reactor pool for the RRRP

    International Nuclear Information System (INIS)

    Alberro, J.G.; Abbate, A.D.

    2005-01-01

    The purpose of the present document is to describe the structural design of the Reactor Pool relevant to the RRRP (Replacement Research Reactor Project) for the Australian Nuclear Science and Technology Organisation. The structural analysis required coordinated design, engineering, analysis, and fabrication efforts. The pool has been designed, manufactured, and inspected following as guideline the ASME Boiler and Pressure Vessel Code, which defines the requirements for the pool to withstand hydrostatic and mechanical forces, ensuring its integrity throughout its lifetime. Standard off-the-shelf finite element programs (Nastran and Ansys codes) were used to evaluate the pool and further qualify the design and its construction. Both global and local effect analyses were carried out. The global analysis covers the structural integrity of the pool wall (6 mm thick) considering the different load states acting on it, namely hydrostatic pressure, thermal expansion, and seismic event. The local analysis evaluates the structural behaviour of the pool at specific points resulting from the interaction among components. It is confirmed that maximum stresses and displacements fall below the allowable values required by the ASME Boiler and Pressure Vessel Code. The water pressure analysis was validated by means of a hydrostatic test. (authors)

  4. Structural integrity and management of aging in internal components of BWR reactors

    International Nuclear Information System (INIS)

    Arganis J, C.R.

    2004-01-01

    Presently work the bases to apply structural integrity and the handling of the aging of internal components of the pressure vessel of boiling water reactors of water are revised and is carried out an example of structural integrity in the horizontal welding H4 of the encircling one of the core of a reactor, taking data reported in the literature. It is also revised what is required to carry out the handling program or conduct of the aging (AMP). (Author)

  5. The FRJ-1 (MERLIN) research reactor: its main activity inventory has been removed by successful demolition of the reactor block

    International Nuclear Information System (INIS)

    Stahn, B.; Printz, R.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

    2004-01-01

    The FRJ-1 (MERLIN) research reactor was decommissioned in 1985 after twenty-three years of operation. Demolition of the plant was begun in 1996. The article contains a survey of the demolition steps carried out so far within the framework of three partial permits. The main activity is the demolition of the reactor core structures as a precondition for subsequent measures to ensure clearance measurements of the building. The core structures are demolished which were exposed to high neutron fluxes during reactor operation and now show the highest activity and dose rate levels, except for the core internals. For demolition and disassembly of the metal structures in this part of the plant, the tools specially designed and made include a remotely operated sawing system and a pipe cutting system for internal segmentation of the beam lines. The universal demolition tool for use also above and beyond the concrete structures has been found to be a remotely controlled electrohydraulic demolition shovel. Spreading contamination in the course of the demolition work was avoided. One major reason for this success was the fact that no major airborne contamination existed at any time as a consequence of the quality of the material demolished and also of the consistent use of technical tools. While the reactor block was being demolished, an application for clearance measurement of the reactor hall and subsequent release from the scope of the Atomic Energy Act was filed as early as in mid-2003. The fourth partial permit covering these activities is expected to be issued in the spring of 2004. (orig.)

  6. Structural dynamics in fast reactor accident analysis

    International Nuclear Information System (INIS)

    Fistedis, S.H.

    1975-01-01

    Analyses and codes are under development combining the hydrodynamics and solid mechanics (and more recently the bubble dynamics) phenomena to gage the stresses, strains, and deformations of important primary components, as well as the overall adequacy of primary and secondary containments. An arbitrary partition of the structural components treated evolves into (1) a core mechanics effort; and (2) a primary system and containment program. The primary system and containment program treats the structural response of components beyond the core, starting with the core barrel. Combined hydrodynamics-solid mechanics codes provide transient stresses and strains and final deformations for components such as the reactor vessel, reactor cover, cover holddown bolts, as well as the pulses for which the primary piping system is to be analyzed. Both, Lagrangian and Eulerian two-dimensional codes are under development, which provide greater accuracy and longer durations for the treatment of HCDA. The codes are being augmented with bubble migration capability pertaining to the latter stages of the HCDA, after slug impact. Recent developments involve the adaptation of the 2-D Eulerian primary system code to the 2-D elastic-plastic treatment of primary piping. Pulses are provided at the vessel-primary piping interfaces of the inlet and outlet nozzles, calculation includes the elbows and pressure drops along the components of the primary piping system. Recent improvements to the primary containment codes include introduction of bending strength in materials, Langrangian mesh regularization techniques, and treatment of energy absorbing materials for the slug impact. Another development involves the combination of a 2-D finite element code for the reactor cover with the hydrodynamic containment code

  7. Dynamic soil-structure interaction analysis based on discretized Green function

    International Nuclear Information System (INIS)

    Muto, K.; Kobayashi, T.; Nakahara, M.

    1983-01-01

    In the seismic design of massive and rigid structure such as a nuclear reactor building, it is important to evaluate the dynamic interaction effect between soil and structure. The authors developed an advanced and practical method to evaluate the interaction effect between the soil which is considered to be semi-infinite elastic medium, and the structure in which flexibility is considered. In this report, this method is applied to a seismic analysis of the full size BWR Mark I type reactor building. For horizontal input earthquake, the vibrational degrees of freedom shall be considered both horizontal and vertical as the rocking response occurs because of the overturning moment caused by the building's horizontal response. The results of earthquake response analysis show that the floors deform in-place and the response acceleration at the center of the floor is larger than that of at the side wall. The response spectra also differ each other even if on the same floor because of the in-place deformation of the floor slab. It means that in analytical modeling of the reactor building, multi-stick model considering deformation of floor slab is required instead of single-stick model. The ratio of the peak acceleration response of the roof floor to the input earthquake is about 2.5. (orig./HP)

  8. Energy-averaged neutron cross sections of fast-reactor structural materials

    International Nuclear Information System (INIS)

    Smith, A.; McKnight, R.; Smith, D.

    1978-02-01

    The status of energy-averaged cross sections of fast-reactor structural materials is outlined with emphasis on U.S. data programs in the neutron-energy range 1-10 MeV. Areas of outstanding accomplishment and significant uncertainty are noted with recommendations for future efforts. Attention is primarily given to the main constituents of stainless steel (e.g., Fe, Ni, and Cr) and, secondarily, to alternate structural materials (e.g., V, Ti, Nb, Mo, Zr). Generally, the mass regions of interest are A approximately 50 to 60 and A approximately 90 to 100. Neutron total and elastic-scattering cross sections are discussed with the implication on the non-elastic-cross sections. Cross sections governing discrete-inelastic-neutron-energy transfers are examined in detail. Cross sections for the reactions (n;p), (n;n',p), (n;α), (n;n',α) and (n;2n') are reviewed in the context of fast-reactor performance and/or diagnostics. The primary orientation of the discussion is experimental with some additional attention to the applications of theory, the problems of evaluation and the data sensitivity of representative fast-reactor systems

  9. Structural materials for fusion reactors

    International Nuclear Information System (INIS)

    Victoria, M.; Baluc, N.; Spaetig, P.

    2001-01-01

    In order to preserve the condition of an environmentally safe machine, present selection of materials for structural components of a fusion reactor is made not only on the basis of adequate mechanical properties, behavior under irradiation and compatibility with other materials and cooling media, but also on their radiological properties, i.e. activity, decay heat, radiotoxicity. These conditions strongly limit the number of materials available to a few families of alloys, generically known as low activation materials. We discuss the criteria for deciding on such materials, the alloys resulting from the application of the concept and the main issues and problems of their use in a fusion environment. (author)

  10. Structural response of steel high rise buildings to fire

    DEFF Research Database (Denmark)

    Gentili, Filippo; Giuliani, Luisa; Bontempi, Franco

    2013-01-01

    Due to the significant vertical elevation and complexity of the structural system, high rise buildings may suffer from the effects of fire more than other structures. For this reason, in addition to evacuation strategies and active fire protection, a careful consideration of structural response t...

  11. Analysis for mechanical consequences of a core disruptive accident in Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Chellapandi, P.; Velusamy, K.; Chetal, S.C.; Bhoje, S.B.; Lal, H.; Sethi, V.S.

    2003-01-01

    The mechanical consequences of a core disruptive accident (CDA) in a fast breeder reactor are described. The consequences are development of deformations and strains in the vessels, intermediate heat exchangers (IHX) and decay heat exchangers (DHX), impact of sodium slug on the bottom surface of the top shield, sodium release to reactor containment building through top shield penetrations, sodium fire and consequent temperature and pressure rise in reactor containment building (RCB). These are quantified for 500 MWe Prototype Fast Breeder Reactor (PFBR) for a CDA with 100 MJ work potential. The results are validated by conducting a series of experiments on 1/30 and 1/13 scaled down models with increasing complexities. Mechanical energy release due to nuclear excursion is simulated by chemical explosion of specially developed low density explosive charge. Based on these studies, structural integrity of primary containment, IHX and DHX is demonstrated. The sodium release to RCB is 350 kg which causes pressure rise of 12 kPa in RCB. (author)

  12. Multi functional roof structures of the energy efficient buildings

    Directory of Open Access Journals (Sweden)

    Krstić Aleksandra

    2006-01-01

    Full Text Available Modern architectural concepts, which are based on rational energy consumption of buildings and the use of solar energy as a renewable energy source, give the new and significant role to the roofs that become multifunctional structures. Various energy efficient roof structures and elements, beside the role of protection, provide thermal and electric energy supply, natural ventilation and cooling of a building, natural lighting of the indoor space sunbeam protection, water supply for technical use, thus according to the above mentioned functions, classification and analysis of such roof structures and elements are made in this paper. The search for new architectural values and optimization in total energy balance of a building or the likewise for the urban complex, gave to roofs the role of "climatic membranes". Contemporary roof forms and materials clearly exemplify their multifunctional features. There are numerous possibilities to achieve the new and attractive roof design which broadens to the whole construction. With such inducement, this paper principally analyze the configuration characteristics of the energy efficient roof structures and elements, as well as the visual effects that may be achieved by their application.

  13. Assessment of fission product release from the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Generino, G.

    1984-07-01

    Fission product releases from the RCB associated with hypothetical core-melt accidents ABβ, S 2 CDβ and TLBβ in a PWR-900 MWe have been performed using French computer codes (in particular, the JERICHO Code for containment response analysis and AEROSOLS/B1 for aerosol behavior in the containment) related to thermalhydraulics and fission product behavior in the primary system and in the reactor containment building

  14. Los Alamos National Laboratory case studies on decommissioning of research reactors and a small nuclear facility

    International Nuclear Information System (INIS)

    Salazar, M.D.

    1998-01-01

    Approximately 200 contaminated surplus structures require decommissioning at Los Alamos National Laboratory. During the last 10 years, 50 of these structures have undergone decommissioning. These facilities vary from experimental research reactors to process/research facilities contaminated with plutonium-enriched uranium, tritium, and high explosives. Three case studies are presented: (1) a filter building contaminated with transuranic radionuclides; (2) a historical water boiler that operated with a uranyl-nitrate solution; and (3) the ultra-high-temperature reactor experiment, which used enriched uranium as fuel

  15. Los Alamos National Laboratory case studies on decommissioning of research reactors and a small nuclear facility

    Energy Technology Data Exchange (ETDEWEB)

    Salazar, M.D.

    1998-12-01

    Approximately 200 contaminated surplus structures require decommissioning at Los Alamos National Laboratory. During the last 10 years, 50 of these structures have undergone decommissioning. These facilities vary from experimental research reactors to process/research facilities contaminated with plutonium-enriched uranium, tritium, and high explosives. Three case studies are presented: (1) a filter building contaminated with transuranic radionuclides; (2) a historical water boiler that operated with a uranyl-nitrate solution; and (3) the ultra-high-temperature reactor experiment, which used enriched uranium as fuel.

  16. Statistical techniques for the identification of reactor component structural vibrations

    International Nuclear Information System (INIS)

    Kemeny, L.G.

    1975-01-01

    The identification, on-line and in near real-time, of the vibration frequencies, modes and amplitudes of selected key reactor structural components and the visual monitoring of these phenomena by nuclear power plant operating staff will serve to further the safety and control philosophy of nuclear systems and lead to design optimisation. The School of Nuclear Engineering has developed a data acquisition system for vibration detection and identification. The system is interfaced with the HIFAR research reactor of the Australian Atomic Energy Commission. The reactor serves to simulate noise and vibrational phenomena which might be pertinent in power reactor situations. The data acquisition system consists of a small computer interfaced with a digital correlator and a Fourier transform unit. An incremental tape recorder is utilised as a backing store and as a means of communication with other computers. A small analogue computer and an analogue statistical analyzer can be used in the pre and post computational analysis of signals which are received from neutron and gamma detectors, thermocouples, accelerometers, hydrophones and strain gauges. Investigations carried out to date include a study of the role of local and global pressure fields due to turbulence in coolant flow and pump impeller induced perturbations on (a) control absorbers, (B) fuel element and (c) coolant external circuit and core tank structure component vibrations. (Auth.)

  17. Mechanical components: fabrication of major reactor structures

    International Nuclear Information System (INIS)

    Nicholson, S.

    1985-01-01

    The paper examines the validity of criticisms of quality assurance of mechanical plant and welded products within major reactor structures, taking into account experience gained on the AGR's. Various constructive recommendations are made aimed at furthering the objectives of quality assurance in the nuclear industry and making it more cost-effective. Current levels of quality related costs in the fabrication industry are provided as a basis for discussion. (U.K.)

  18. Finding the displacement of wood structure in heritage building by 3D laser scanner

    Science.gov (United States)

    Lee, M. C.; Tsai, Y. L.; Wang, R. Z.; Lin, M. L.

    2015-08-01

    Heritage buildings are highly prone to long term damage from the microclimate, scourge and vandalism, which can result in damaged materials, structures, painting and cultural heritage items. This study will focus on finding the displacement of wood structural members through the use of a 3D laser scanner and the 4D concept of time. The results will compare the scans from different periods to find the difference (if any) in the structural member position. Wood structures usually consist of numerous wood members connected to form the structure. However, these members can be damaged in various ways such as physical mechanisms, chemical reactions, and biological corrosion. When damage to the wood structure occurs, the structural displacement can be affected, and if affected severely, can lead to a building collapse. Monitoring of the structural displacement is the best way to discover damage immediately and to preserve the heritage building. However, the Cultural Heritage Preservation Law in Taiwan prohibits the installation of monitoring instruments (e.g strain gauge, accelerometer) in historic structures (heritage buildings). Scanning the wood structure with 3D lasers is the most non-intrusive method and quickly achieves displacement through visualization. The displacement scan results can be compared with different periods and different members to analyze the severity of damage. Once the 3D scanner is installed, the whole building is scanned, and point clouds created to build the visual building model. The structural displacement can be checked via the building model and the differences are measured between each member to find the high risk damaged areas or members with large displacement. Early detection of structural damage is the most effective way means of preservation.

  19. LIFE-CYCLE COST MODEL AND DESIGN OPTIMIZATION OF BASE ISOLATED BUILDING STRUCTURES

    Directory of Open Access Journals (Sweden)

    Chara C. Mitropoulou

    2016-11-01

    Full Text Available Design of economic structures adequately resistant to withstand during their service life, without catastrophic failures, all possible loading conditions and to absorb the induced seismic energy in a controlled fashion, has been the subject of intensive research so far. Modern buildings usually contain extremely sensitive and costly equipment that are vital in business, commerce, education and/or health care. The building contents frequently are more valuable than the buildings them-selves. Furthermore, hospitals, communication and emergency centres, police and fire stations must be operational when needed most: immediately after an earthquake. Conventional con-struction can cause very high floor accelerations in stiff buildings and large interstorey drifts in flexible structures. These two factors cause difficulties in insuring the safety of both building and its contents. For this reason base-isolated structures are considered as an efficient alternative design practice to the conventional fixed-base one. In this study a systematic assessment of op-timized fixed and base-isolated reinforced concrete buildings is presented in terms of their initial and total cost taking into account the life-cycle cost of the structures.

  20. Measurement of fatigue crack growth rate of reactor structural material in air based on DCPD method

    International Nuclear Information System (INIS)

    Du Donghai; Chen Kai; Yu Lun; Zhang Lefu; Shi Xiuqiang; Xu Xuelian

    2014-01-01

    The principles and details of direct current potential drop (DCPD) in monitoring the crack growth of reactor structural materials was introduced in this paper. Based on this method, the fatigue crack growth rate (CGR) of typical structural materials in nuclear power systems was measured. The effects of applied load, load ratio and loading frequency on the fatigue crack growth rate of reactor structural materials were discussed. The result shows that the fatigue crack growth rate of reactor structural materials depends on the hardness of materials, and the harder the material is, the higher the rate of crack growth is. (authors)

  1. RANDOM FUNCTIONS AND INTERVAL METHOD FOR PREDICTING THE RESIDUAL RESOURCE OF BUILDING STRUCTURES

    Directory of Open Access Journals (Sweden)

    Shmelev Gennadiy Dmitrievich

    2017-11-01

    Full Text Available Subject: possibility of using random functions and interval prediction method for estimating the residual life of building structures in the currently used buildings. Research objectives: coordination of ranges of values to develop predictions and random functions that characterize the processes being predicted. Materials and methods: when performing this research, the method of random functions and the method of interval prediction were used. Results: in the course of this work, the basic properties of random functions, including the properties of families of random functions, are studied. The coordination of time-varying impacts and loads on building structures is considered from the viewpoint of their influence on structures and representation of the structures’ behavior in the form of random functions. Several models of random functions are proposed for predicting individual parameters of structures. For each of the proposed models, its scope of application is defined. The article notes that the considered approach of forecasting has been used many times at various sites. In addition, the available results allowed the authors to develop a methodology for assessing the technical condition and residual life of building structures for the currently used facilities. Conclusions: we studied the possibility of using random functions and processes for the purposes of forecasting the residual service lives of structures in buildings and engineering constructions. We considered the possibility of using an interval forecasting approach to estimate changes in defining parameters of building structures and their technical condition. A comprehensive technique for forecasting the residual life of building structures using the interval approach is proposed.

  2. Finite Element Method in the Three Dimensions Deformation Computation ofKartini Reactor Stack

    International Nuclear Information System (INIS)

    Supriyono; Syarip; Wibisono, I

    2000-01-01

    The calculation of the Kartini reactor stack i.e. one of the nuclearinstallations in P3TM-BATAN Yogyakarta by using SAP 90 software have beendone. The calculation is done as a safety review of building towards theearthquake style in Yogyakarta. The 3-dimension deformation calculation isperformed by the numeric method i.e. finite element method with the form ofelements is the shell. The result obtained showed that the construction oftower safe to the existing earthquake, where the moment exerted as a resultof earthquake style was different under the moment having been kept by thebuilding structure. By knowing the deformation on the stack it is expectedcould be used for concluding the strength of the whole reactor building.(author)

  3. Decommissioning of the MTR-605 process water building at the Idaho National Engineering Laboratory. Final report

    International Nuclear Information System (INIS)

    Browder, J.H.; Wills, E.L.

    1985-01-01

    Decontamination and decommissioning (D and D) of the unused radioactively contaminated portions of the MTR-605 building at the Test Reactor Area of the Idaho National Engineering Laboratory has been completed; this final report describes the D and D project. The building is a two-story concrete structure that was used to house piping systems to channel and control coolant water flow for the Materials Testing Reactor (MTR), a 40 MW (thermal) light water test reactor that was operated from 1952 until 1970 and then deactivated. D and D project objectives were to reduce potential environmental and radioactive contamination hazards to levels as low a reasonably achievable. Primary tasks of the D and D project were: to remove contaminated piping (about 400 linear ft of 36- and 30-in.-dia stainless steel pipe) and valves from the primary coolant pipe tunnels, to remove a primary coolant pump and piping, and to remove the three 8-ft-dia by 25-ft-long evaporators from the building second floor

  4. Radon entry into buildings: Effects of atmospheric pressure fluctuations and building structural factors

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Allen Lantham [Univ. of California, Berkeley, CA (United States). Dept. of Mechanical Engineering

    1996-05-01

    An improved understanding of the factors that control radon entry into buildings is needed in order to reduce the public health risks caused by exposure to indoor radon. This dissertation examines three issues associated with radon entry into buildings: (1) the influence of a subslab gravel layer and the size of the openings between the soil and the building interior on radon entry; (2) the effect of atmospheric pressure fluctuations on radon entry; and (3) the development and validation of mathematical models which simulate radon and soil-gas entry into houses. Experiments were conducted using two experimental basements to examine the influence of a subslab gravel layer on advective radon entry driven by steady indoor-outdoor pressure differences. These basement structures are identical except that in one the floor slab lies directly on native soil whereas in the other the slab lies on a high-permeability gravel layer. The measurements indicate that a high permeability subslab gravel layer increases the advective radon entry rate into the structure by as much as a factor of 30. The magnitude of the enhancement caused by the subslab gravel layer depends on the area of the openings in the structure floor; the smaller the area of these openings the larger the enhancement in the radon entry rate caused by the subslab gravel layer. A three-dimensional, finite-difference model correctly predicts the effect of a subslab gravel layer and open area configuration on advective radon entry driven by steady indoor-outdoor pressure differences; however, the model underpredicts the absolute entry rate into each structure by a factor of 1.5.

  5. Radon entry into buildings: Effects of atmospheric pressure fluctuations and building structural factors

    International Nuclear Information System (INIS)

    Robinson, A.L.

    1996-05-01

    An improved understanding of the factors that control radon entry into buildings is needed in order to reduce the public health risks caused by exposure to indoor radon. This dissertation examines three issues associated with radon entry into buildings: (1) the influence of a subslab gravel layer and the size of the openings between the soil and the building interior on radon entry; (2) the effect of atmospheric pressure fluctuations on radon entry; and (3) the development and validation of mathematical models which simulate radon and soil-gas entry into houses. Experiments were conducted using two experimental basements to examine the influence of a subslab gravel layer on advective radon entry driven by steady indoor-outdoor pressure differences. These basement structures are identical except that in one the floor slab lies directly on native soil whereas in the other the slab lies on a high-permeability gravel layer. The measurements indicate that a high permeability subslab gravel layer increases the advective radon entry rate into the structure by as much as a factor of 30. The magnitude of the enhancement caused by the subslab gravel layer depends on the area of the openings in the structure floor; the smaller the area of these openings the larger the enhancement in the radon entry rate caused by the subslab gravel layer. A three-dimensional, finite-difference model correctly predicts the effect of a subslab gravel layer and open area configuration on advective radon entry driven by steady indoor-outdoor pressure differences; however, the model underpredicts the absolute entry rate into each structure by a factor of 1.5

  6. An investigation of structural design methodology for HTGR reactor internals with ceramic materials (Contract research)

    International Nuclear Information System (INIS)

    Sumita, Junya; Shibata, Taiju; Nakagawa, Shigeaki; Iyoku, Tatsuo; Sawa, Kazuhiro

    2008-03-01

    To advance the performance and safety of HTGR, heat-resistant ceramic materials are expected to be used as reactor internals of HTGR. C/C composite and superplastic zirconia are the promising materials for this purpose. In order to use these new materials as reactor internals in HTGR, it is necessary to establish a structure design method to guarantee the structural integrity under environmental and load conditions. Therefore, C/C composite expected as reactor internals of VHTR is focused and an investigation on the structural design method applicable to the C/C composite and a basic applicability of the C/C composite to representative structures of HTGR were carried out in this report. As the results, it is found that the competing risk theory for the strength evaluation of the C/C composite is applicable to design method and C/C composite is expected to be used as reactor internals of HTGR. (author)

  7. Jacking mechanism for upper internals structure of a liquid metal nuclear reactor

    International Nuclear Information System (INIS)

    Gillett, J.E.; Wineman, A.L.

    1984-01-01

    A jacking mechanism is described for raising the upper internals structure of a liquid metal nuclear reactor which jacking mechanism uses a system of gears and drive shafts to transmit force from a single motor to four mechanically synchronized ball jacks to raise and lower support columns which support the upper internals structure. The support columns have a pin structure which rides up and down in a slot in a housing fixed to the reactor head. The pin has two locking plates which can be rotated around the pin to bring bolt holes through the locking plates into alignment with a set of bolt holes in the housing, there being a set of such housing bolt holes corresponding to both a raised and a lowered position of the support column. When the locking plate is so aligned, a surface of the locking plate mates with a surface in the housing such that the support column is then supported by the locking plate and not by the ball jacks. Since the locking plates are to be installed and bolted to the housing during periods of reactor operation, the ball jacks need not be sized to react the large forces which occur or potentially could occur on the upper internals structure of the reactor during operation. The locking plates react these loads. The ball jacks, used only during refueling, can be smaller, which enable conventionally available equipment to fulfill the precision requirements for the task within available space

  8. Design study of an IHX support structure for a POOL-TYPE Sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Park, Chang Gyu; Kim, Jong Bum; Lee, Jae Han

    2009-01-01

    The IHX (Intermediate Heat eXchanger) for a pool-type SFR (Sodium-cooled Fast Reactor) system transfers heat from the primary high temperature sodium to the intermediate cold temperature sodium. The upper structure of the IHX is a coaxial structure designed to form a flow path for both the secondary high temperature and low temperature sodium. The coaxial structure of the IHX consists of a central downcomer and riser for the incoming and outgoing intermediate sodium, respectively. The IHX of a pool-type SFR is supported at the upper surface of the reactor head with an IHX support structure that connects the IHX riser cylinder to the reactor head. The reactor head is generally maintained at the low temperature regime, but the riser cylinder is exposed in the elevated temperature region. The resultant complicated temperature distribution of the co-axial structure including the IHX support structure may induce a severe thermal stress distribution. In this study, the structural feasibility of the current upper support structure concept is investigated through a preliminary stress analysis and an alternative design concept to accommodate the IHTS (Intermediate Heat Transport System) piping expansion loads and severe thermal stress is proposed. Through the structural analysis it is found that the alternative design concept is effective in reducing the thermal stress and acquiring structural integrity

  9. Typological diversity of tall buildings and complexes in relation to their functional structure

    Science.gov (United States)

    Generalov, Viktor P.; Generalova, Elena M.; Kalinkina, Nadezhda A.; Zhdanova, Irina V.

    2018-03-01

    The paper focuses on peculiarities of tall buildings and complexes, their typology and its formation in relation to their functional structure. The research is based on the analysis of tall buildings and complexes and identifies the following main functional elements of their formation: residential, administrative (office), hotel elements. The paper also considers the following services as «disseminated» in the space-planning structure: shops, medicine, entertainment, kids and sports facilities, etc., their location in the structure of the total bulk of the building and their impact on typological diversity. Research results include suggestions to add such concepts as «single-function tall buildings» and «mixed-use tall buildings and complexes» into the classification of tall buildings. In addition, if a single-function building or complex performs serving functions, it is proposed to add such concepts as «a residential tall building (complex) with provision of services», «an administrative (public) tall building (complex) with provision of services» into the classification of tall buildings. For mixed-use buildings and complexes the following terms are suggested: «a mixed-use tall building with provision of services», «a mixed-use tall complex with provision of services».

  10. Space-planning and structural solutions of low-rise buildings: Optimal selection methods

    Science.gov (United States)

    Gusakova, Natalya; Minaev, Nikolay; Filushina, Kristina; Dobrynina, Olga; Gusakov, Alexander

    2017-11-01

    The present study is devoted to elaboration of methodology used to select appropriately the space-planning and structural solutions in low-rise buildings. Objective of the study is working out the system of criteria influencing the selection of space-planning and structural solutions which are most suitable for low-rise buildings and structures. Application of the defined criteria in practice aim to enhance the efficiency of capital investments, energy and resource saving, create comfortable conditions for the population considering climatic zoning of the construction site. Developments of the project can be applied while implementing investment-construction projects of low-rise housing at different kinds of territories based on the local building materials. The system of criteria influencing the optimal selection of space-planning and structural solutions of low-rise buildings has been developed. Methodological basis has been also elaborated to assess optimal selection of space-planning and structural solutions of low-rise buildings satisfying the requirements of energy-efficiency, comfort and safety, and economical efficiency. Elaborated methodology enables to intensify the processes of low-rise construction development for different types of territories taking into account climatic zoning of the construction site. Stimulation of low-rise construction processes should be based on the system of approaches which are scientifically justified; thus it allows enhancing energy efficiency, comfort, safety and economical effectiveness of low-rise buildings.

  11. Lining facility for FBR type reactor

    International Nuclear Information System (INIS)

    Shimano, Kunio.

    1991-01-01

    In a lining facility for protecting structural material concretes for concrete buildings in an FBR type power plant, sodium-resistant and heat-resistant first and second coating layers are lined at the surface of concretes, and steam releasing materials are disposed between the first and the second coating layers for releasing water contents evaporated from the concretes to the outside. With such a constitution, since there is no structures for welding steel plates to each other as in the prior art, the fabrication is made easy. Further, since cracks of coating materials can be suppressed, reactor safety is improved. (T.M.)

  12. Utilities/industries joint study on seismic isolation systems for LWR: Part II. Observed behaviors of base-isolated general buildings under real earthquakes

    International Nuclear Information System (INIS)

    Matsumura, Takao; Sato, Shoji; Kato, Muneaki

    1989-01-01

    This paper describes the observed behavior of base-isolated buildings under real earthquake conditions. These buildings were constructed by five construction companies participating in the Joint Study on Seismic Isolation Systems for lightwater reactors. All the buildings are medium- or low-height buildings of reinforced-concrete structures with combinations of laminated rubber bearing or sliding bearings and various damping devices

  13. Passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  14. Regulations for RA reactor operation

    International Nuclear Information System (INIS)

    1980-09-01

    Regulations for RA reactor operation are written in accordance with the legal regulations defined by the Law about radiation protection and related legal acts, as well as technical standards according to the IAEA recommendations. The contents of this book include: fundamental data about the reactor; legal regulations for reactor operation; organizational scheme for reactor operation; general and detailed instructions for operation, behaviour in the reactor building, performing experiments; operating rules for operation under steady state and accidental conditions [sr

  15. Design study of blanket structure for tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    1979-11-01

    Design study of the blanket structure for JAERI Experimental Fusion Reactor (JXFR) has been carried out. Studied here were fabrication and testing of the blanket structure (blanket cells, blanket rings, piping and blanket modules), assembly and disassembly of the blanket module, and monitering and testing technique. Problems in design and fabrication of the blanket structure could be revealed. Research and development problems for the future were also disclosed. (author)

  16. Windscale advanced gas-cooled reactor (WAGR) decommissioning project overview

    International Nuclear Information System (INIS)

    Pattinson, A.

    2003-01-01

    The current BNFL reactor decommissioning projects are presented. The projects concern power reactor sites at Berkely, Trawsfynydd, Hunterstone, Bradwell, Hinkley Point; UKAEA Windscale Pile 1; Research reactors within UK Scottish Universities at East Kilbride and ICI (both complete); WAGR. The BNFL environmental role include contract management; effective dismantling strategy development; implementation and operation; sentencing, encapsulation and transportation of waste. In addition for the own sites it includes strategy development; baseline decommissioning planning; site management and regulator interface. The project objectives for the Windscale Advanced Gas-Cooled Reactor (WAGR) are 1) Safe and efficient decommissioning; 2) Building of good relationships with customer; 3) Completion of reactor decommissioning in 2005. The completed WAGR decommissioning campaigns are: Operational Waste; Hot Box; Loop Tubes; Neutron Shield; Graphite Core and Restrain System; Thermal Shield. The current campaign is Lower Structures and the remaining are: Pressure vessel and Insulation; Thermal Columns and Outer Vault Membrane. An overview of each campaign is presented

  17. Structural evaluation of the 2736Z Building for seismic loads

    International Nuclear Information System (INIS)

    Giller, R.A.

    1994-01-01

    The 2736Z building structure is evaluated for high-hazard loads. The 2736Z building is analyzed herein for normal and seismic loads and is found to successfully meet the guidelines of UCRL-15910 along with the related codes requirements

  18. Reactor control rod supporting structure

    International Nuclear Information System (INIS)

    Akimoto, Tokuzo; Miyata, Hiroshi.

    1984-01-01

    Purpose: To enable stable reactor core control even in extremely great vertical earthquakes, as well as under normal operation conditions in FBR type reactors. Constitution: Since a mechanism for converting the rotational movement of a control rod into vertical movement is placed at the upper portion of the reactor core at high temperature, the mechanism should cause fusion or like other danger after the elapse of a long period of time. In view of the above, the conversion mechanism is disposed to the lower portion of the reactor core at a lower temperature region. Further, the connection between the control rod and the control rod drive can be separated upon great vertical earthquakes. (Seki, T.)

  19. Structural capacity assessment of WWER-1000 MW reactor containment. Progress report

    International Nuclear Information System (INIS)

    Jordanov, M.

    1999-01-01

    The objective of the project is to provide assessment of the structural behaviour and safety capacity of the WWER-1000 MW Reactor Building Containment at Kozloduy NPP under critical combination of loads according to the current international requirements. The analysis is focused on a realistic assessment of the Containment taking into account the non-linear shell behaviour of the pre-stressed reinforced concrete structure. Previous assessments of the status of pre stressing cables pointed out that the efficiency of the Containment as a final defence barrier for internal and external events depends on their reliability. Due to this, the experimental data obtained from embedded sensors (gauges) at pre-stressed shell structure is to be compared with the results from analytical investigations. The reliability of the WWER-1000 MW accident prevention system is under evaluation in the project. The Soviet standard design WWER-1000 MW type units installed in Kozloduy NPP were originally designed for a Safe Shutdown Earthquake (SSE) with a peak ground acceleration (PGA) of 0.1g. The new site seismicity studies revealed that the seismic hazard for the site significantly exceeds the originally estimated and a Review Level Earthquake (RLE) anchored to PGA=0.20g was proposed for re-assessment of the structures and equipment at Kozloduy NPP. The scope of the study is a re-assessment of the Containment structure under critical combination of loads according to the current safety and reliability requirements, including comparison between the Russian design requirements and the international regulations. Additionally, an investigation of the pre-stressing technology and the annual control of the cables' pre-stressing of the Containment is to be made. The crane influence on the dynamic behaviour of the Containment will be done as well as a study of the integrity of the Containment as a final defence barrier

  20. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'JOYO'. 2. Replacement of upper core structure

    International Nuclear Information System (INIS)

    Ushiki, Hiroshi; Ito, Hiromichi; Okuda, Eiji; Suzuki, Nobuhiro; Sasaki, Jun; Oota, Katsu; Kawahara, Hirotaka; Takamatsu, Misao; Nagai, Akinori; Okawa, Toshikatsu

    2015-01-01

    In the experimental fast reactor Joyo, it was confirmed that the top of the irradiation test sub-assembly of MARICO-2 (material testing rig with temperature control) had bent onto the in-vessel storage rack as an obstacle and had damaged the upper core structure (UCS) in 2007. As a part of the restoration work, UCS replacement was begun at March 24, 2014 and was completed at December 17. In-vessel repair (including observation) for sodium-cooled fast reactors (SFRs) is distinct from that for light water reactors and necessitates independent development. Application of developed in-vessel repair techniques to operation and maintenance of SFRs enhanced their safety and integrity. There is little UCS replacement experience in the world and this experience and insights, which were accumulated in the replacement work of in-vessel large structure (UCS) used for more than 30 years, are expected to improve the in-vessel repair techniques in SFRs. (author)

  1. Determining building interior structures using compressive sensing

    Science.gov (United States)

    Lagunas, Eva; Amin, Moeness G.; Ahmad, Fauzia; Nájar, Montse

    2013-04-01

    We consider imaging of the building interior structures using compressive sensing (CS) with applications to through-the-wall imaging and urban sensing. We consider a monostatic synthetic aperture radar imaging system employing stepped frequency waveform. The proposed approach exploits prior information of building construction practices to form an appropriate sparse representation of the building interior layout. We devise a dictionary of possible wall locations, which is consistent with the fact that interior walls are typically parallel or perpendicular to the front wall. The dictionary accounts for the dominant normal angle reflections from exterior and interior walls for the monostatic imaging system. CS is applied to a reduced set of observations to recover the true positions of the walls. Additional information about interior walls can be obtained using a dictionary of possible corner reflectors, which is the response of the junction of two walls. Supporting results based on simulation and laboratory experiments are provided. It is shown that the proposed sparsifying basis outperforms the conventional through-the-wall CS model, the wavelet sparsifying basis, and the block sparse model for building interior layout detection.

  2. Dose rate in the reactor room and environment during maintenance in fusion reactors

    International Nuclear Information System (INIS)

    Maki, Koichi; Satoh, Satoshi; Takatsu, Hideyuki; Seki, Yasushi

    1995-01-01

    According to the International Thermonuclear Experimental Reactor (ITER) conceptual design activity, after reactor shutdown, damaged segments are pulled up from the reactor and hung from the reactor room ceiling by a remote handling device. The dose rate in the reactor room and the environment is estimated for this situation, and the following results are obtained. First, the dose rate in the room is > 10 8 μSv/h. Since this dose rate is 10 7 times greater than the biological radiation shielding design limit of 25 μSv/h, workers cannot enter the room. Second, lenses and optical fiber composed of glass that is radiation resistant up to 10 6 Gy would be damaged after <100 h near the segment, and devices using semiconductors could not work after several hours or so in the aforementioned dose-rate conditions. Third, during suspension of one blanket segment from the ceiling, the dose rate in the site boundary can be reduced by one order by a 23-cm-thicker reactor building roof. To reduce dose rate in public exposure to a value that is less than one-tenth of the public exposure radiation shielding design limit of 100 μSv/yr, the distance of the site boundary from the reactor must be greater than 200 m for a reactor building with a 160-cm-thick concrete roof. 9 refs., 6 figs., 2 tabs

  3. Damping in building structures during earthquakes: test data and modeling

    International Nuclear Information System (INIS)

    Coats, D.W. Jr.

    1982-01-01

    A review and evaluation of the state-of-the-art of damping in building structures during earthquakes is presented. The primary emphasis is in the following areas: 1) the evaluation of commonly used mathematical techniques for incorporating damping effects in both simple and complex systems; 2) a compilation and interpretation of damping test data; and 3) an evaluation of structure testing methods, building instrumentation practices, and an investigation of rigid-body rotation effects on damping values from test data. A literature review provided the basis for evaluating mathematical techiques used to incorporate earthquake induced damping effects in simple and complex systems. A discussion on the effectiveness of damping, as a function of excitation type, is also included. Test data, from a wide range of sources, has been compiled and interpreted for buidings, nuclear power plant structures, piping, equipment, and isolated structural elements. Test methods used to determine damping and frequency parameters are discussed. In particular, the advantages and disadvantages associated with the normal mode and transfer function approaches are evaluated. Additionally, the effect of rigid-body rotations on damping values deduced from strong-motion building response records is investigated. A discussion of identification techniques typically used to determine building parameters (frequency and damping) from strong motion records is included. Finally, an analytical demonstration problem is presented to quantify the potential error in predicting fixed-base structural frequency and damping values from strong motion records, when rigid-body rotations are not properly accounted for

  4. Object-Based Dense Matching Method for Maintaining Structure Characteristics of Linear Buildings.

    Science.gov (United States)

    Su, Nan; Yan, Yiming; Qiu, Mingjie; Zhao, Chunhui; Wang, Liguo

    2018-03-29

    In this paper, we proposed a novel object-based dense matching method specially for the high-precision disparity map of building objects in urban areas, which can maintain accurate object structure characteristics. The proposed framework mainly includes three stages. Firstly, an improved edge line extraction method is proposed for the edge segments to fit closely to building outlines. Secondly, a fusion method is proposed for the outlines under the constraint of straight lines, which can maintain the building structural attribute with parallel or vertical edges, which is very useful for the dense matching method. Finally, we proposed an edge constraint and outline compensation (ECAOC) dense matching method to maintain building object structural characteristics in the disparity map. In the proposed method, the improved edge lines are used to optimize matching search scope and matching template window, and the high-precision building outlines are used to compensate the shape feature of building objects. Our method can greatly increase the matching accuracy of building objects in urban areas, especially at building edges. For the outline extraction experiments, our fusion method verifies the superiority and robustness on panchromatic images of different satellites and different resolutions. For the dense matching experiments, our ECOAC method shows great advantages for matching accuracy of building objects in urban areas compared with three other methods.

  5. Object-Based Dense Matching Method for Maintaining Structure Characteristics of Linear Buildings

    Directory of Open Access Journals (Sweden)

    Nan Su

    2018-03-01

    Full Text Available In this paper, we proposed a novel object-based dense matching method specially for the high-precision disparity map of building objects in urban areas, which can maintain accurate object structure characteristics. The proposed framework mainly includes three stages. Firstly, an improved edge line extraction method is proposed for the edge segments to fit closely to building outlines. Secondly, a fusion method is proposed for the outlines under the constraint of straight lines, which can maintain the building structural attribute with parallel or vertical edges, which is very useful for the dense matching method. Finally, we proposed an edge constraint and outline compensation (ECAOC dense matching method to maintain building object structural characteristics in the disparity map. In the proposed method, the improved edge lines are used to optimize matching search scope and matching template window, and the high-precision building outlines are used to compensate the shape feature of building objects. Our method can greatly increase the matching accuracy of building objects in urban areas, especially at building edges. For the outline extraction experiments, our fusion method verifies the superiority and robustness on panchromatic images of different satellites and different resolutions. For the dense matching experiments, our ECOAC method shows great advantages for matching accuracy of building objects in urban areas compared with three other methods.

  6. Criteria of choosing building structures for rooftop boiler rooms

    Directory of Open Access Journals (Sweden)

    Plotnikov Artyom

    2018-01-01

    Full Text Available The paper investigates parameters of noise and vibration distribution in the territory of residential area depending on the structural materials and power of independent heat supply systems. Rooftop boiler rooms are decentralized heat supply systems in buildings. Today, residential areas are strongly affected by noise and vibrations. Adverse effects are isolated by buildings materials, protective shields and floating floors. Rooftop boiler rooms located in Tyumen city were investigated within this research. Structures of rooftop boiler rooms were analyzed. Acoustic analysis results and the parameters of equivalent continuous sound level are presented. An option for improvement of rooftop boiler rooms structures is suggested. Comparison of capital investments in construction and installation activities is carried out. Conclusion on capital investments required for noise protection is made.

  7. TMI-2 [Three Mile Island Unit 2] reactor building dose reduction task force

    International Nuclear Information System (INIS)

    Daniels, R.S.

    1988-01-01

    In late October 1982, the director of Three Mile Island Unit 2 (TMI-2) created the dose reduction task force with the objective of identifying the principal radiological sources in the reactor building and recommending actions to minimize the dose to workers on labor-intensive projects. Members of the task force were drawn form various groups at TMI. Findings and recommendations were presented to the US Nuclear Regulatory Commission in a briefing on November 18, 1982. The task force developed a three-step approach toward dose reduction. Step 1 identified the radiological sources. Step 2 modeled the source and estimated its contribution to the general area dose rates. Step 3 recommended actions to achieve dose reductions consistent with general exposure rate goals

  8. Seismic upgrading of the Brookhaven High Flux Beam Research Reactor

    International Nuclear Information System (INIS)

    Subudhi, M.

    1985-01-01

    In recent years the High Flux Beam Research (HFBR) reactor facility at Brookhaven National Laboratory (BNL) was upgraded from 40 to 50 MW power level. The reactor plant was built in the early sixties to the seismic design requirements of the period, using the static load approach. While the plant power level was upgraded, the seismic design was also improved according to current design criteria. This included the development of new floor response spectra for the facility and an overall seismic analysis of those systems important to the safe shutdown of the reactor. Items included in the reanalysis are the containment building with its internal structure, the piping systems, tanks, equipment, and heat exchangers. This paper describes the procedure utilized in developing the floor response spectra for the existing facility. Also included in the paper are the findings and recommendations, based on the seismic analysis, regarding the seismic adequacy of structural and mechanical systems vital to achieving the safe shutdown of the reactor. 11 references, 4 figures, 1 table

  9. Remote repairs and refurbishment of reactor internal structures of magnox plant

    International Nuclear Information System (INIS)

    Barnes, S.A.; Kelly, D.E.

    1992-01-01

    The original designers of the UK Magnox reactor plant made provision for the then perceived time dependent processes that could have influenced the operational life of the plant. Changes in graphite properties with irradiation, particularly dimensional change, were well understood and in-core samples were provided for subsequent laboratory examination to monitor the processes throughout plant life. The tendency towards embrittlement with irradiation of the steel of the reactor pressure vessels was also acknowledged and again in-core samples were provided for monitoring changes in materials properties in-service and thus provide data in support of structural analyses to sustain the reactor safety cases. (author)

  10. Preparations for decommissioning the TRIGA Mark III Berkeley Research Reactor

    International Nuclear Information System (INIS)

    Denton, Michael M.; Lim, Tek. H.

    1988-01-01

    On December 20, 1986 the chancellor of UC Berkeley announced his decision to decommission the 20 year old Berkeley Research Reactor citing as principal reasons a decline in use and a need to erect a new computer science building over the reactor's site. In order to meet the University's construction timetable for the new building, the reactor staff together with other units of the campus administration have initiated a program to remove the reactor structure and clear the room for unlicensed use as expediently as possible. Due to the sequence of events which must occur in a limited amount of time, the University adopted a policy to contract out as much of the work as possible, including generation of the defueling and decommissioning plans.The first physical step in the decommissioning project is the removal of the irradiated fuel. This task is largely contracted out to a commercial firm with experience in the transport of radioactive materials and reactor fuel. As suggested by the NRC, the reactor will be defueled under the current operating license. This requires that all fuel must be off-site before the DP can be approved. Therefore any delay in defueling in-turn delays the decommissioning. The NRC has given no commitment or date for completion of their review. Informal discussion with NRC project managers and the experience from other facilities indicate that the review process will take between six and nine months

  11. Design of the segment structure and coolant ducts for a fusion reactor blanket and shield

    International Nuclear Information System (INIS)

    Briaris, D.A.; Stanbridge, J.R.

    1978-05-01

    An outline design and analysis of a support structure for the replaceable first wall of a helium cooled fusion reactor blanket has been undertaken. The proposed structure supports all the segment gravitational loads with maximum deflections limited to < 10 mm, and is itself supported off the outer shield by a simple vee-in-groove arrangement. It is a feature of the design that the coaxial coolant pipes and the segment structure operate at the same temperature, making it possible for them to be integrated, thereby avoiding the necessity for pipe bellows. The requirements of cooling the inner arm of the structure and increasing the major radius of the torus by approximately = 0.5 m, have been identified as problems associated with the 'horseshoe' shaped structure applicable to the reactor with divertor. For a ring structure, i.e. reactor without divertor, these problems do not arise. (author)

  12. Damage analysis of TRIGA MARK II Bandung reactor tank material structure

    International Nuclear Information System (INIS)

    Soedardjo; Sumijanto

    2000-01-01

    Damage of Triga Mark II Bandung reactor tank material structure has been analyzed. The analysis carried out was based on ultrasonic inspection result in 1996 and the monthly reports of reactor operation by random data during 1988 up to 1995. Ultrasonic test data had shown that thinning processes on south and west region of reactor out side wall at upper part of water level had happened. Reactor operation data had shown the demineralized water should be added monthly to the reactor and bulk shielding water tank. Both reactor and bulk shielding tank are shielded by concrete of Portland type I cement consisting of CaO content about 58-68 %. The analysis result shows that the reaction between CaO and seepage water from bulk shielding wall had taken place and consequently the reactor out sidewall surroundings became alkaline. Based on Pourbaix diagram, the aluminum reactor tank made of aluminum alloy 6061 T6 would be corroded easily at pH equal an greater than 8.6. The passive layer AI 2 O 3 aluminum metal surface would be broken due to water reaction taken place continuously at high pH and produces hydrogen gas. The light hydrogen gas would expand the concrete cement and its expanding power would open the passive layer of aluminum metal upper tank. The water sea pages from adding water into reactor tank could indicate the upper water level tank corrosion is worse than the lower water level tank. (author)

  13. Research on making reactor buildings of irregular plan and elevation forms aseismatic

    International Nuclear Information System (INIS)

    Okawa, Izuru; Yamauchi, Yasuyuki

    1997-01-01

    The necessity of pursuing the possibility of irregular form buildings as the condition of location for construction is limited, and the rational and economical arrangement of equipment and piping is considered. In order to know the effect that irregular forms exert to the aseismatic ability of buildings, it is indispensable to develop the program for precision three-dimensional elastoplastic analysis at the time of earthquakes. As the means of solving the problem, the introduction of seismic insulation structure is conceivable. The investigation of seismic insulator and its modeling and the analysis of earthquake response were carried out, and the irregular form and the effect of seismic insulation were investigated, and the results of vibration test using test specimens were summarized. The concrete items of investigation were the characteristics of input earthquake motion, the techniques of analysis, the parametric study taking the input and various characteristics of buildings in consideration, and the synthetic assessment. The vibration table experiment and the static loading experiment for the purpose of grasping the response behavior in the case of irregular form of wall type and seismic insulation type structures were carried out, and the results are reported. (K.I.)

  14. Seismic Category I Structures Program

    International Nuclear Information System (INIS)

    Endebrock, E.G.; Dove, R.C.; Anderson, C.A.

    1984-01-01

    The Seismic Category I Structures Program currently being carried out at the Los Alamos National Laboratory is sponsored by the Mechanical/Structural Engineering Branch, Division of Engineering Technology of the Nuclear Regulatory Commission (NRC). This project is part of a program designed to increase confidence in the assessment of Category I nuclear power plant structural behavior beyond the design limit. The program involves the design, construction, and testing of heavily reinforced concrete models of auxiliary buildings, fuel-handling buildings, etc., but doe not include the reactor containment building. The overall goal of the program is to supply to the Nuclear Regulatory Commission experimental information and a validated procedure to establish the sensitivity of the dynamic response of these structures to earthquakes of magnitude beyond the design basis earthquake

  15. RB research reactor Safety Report

    International Nuclear Information System (INIS)

    Sotic, O.; Pesic, M.; Vranic, S.

    1979-04-01

    This RB reactor safety report is a revised and improved version of the Safety report written in 1962. It contains descriptions of: reactor building, reactor hall, control room, laboratories, reactor components, reactor control system, heavy water loop, neutron source, safety system, dosimetry system, alarm system, neutron converter, experimental channels. Safety aspects of the reactor operation include analyses of accident causes, errors during operation, measures for preventing uncontrolled activity changes, analysis of the maximum possible accident in case of different core configurations with natural uranium, slightly and highly enriched fuel; influence of possible seismic events

  16. Increased SRP reactor power

    International Nuclear Information System (INIS)

    MacAfee, I.M.

    1983-01-01

    Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

  17. Impact on breeding rate of different Molten Salt reactor core structures

    International Nuclear Information System (INIS)

    Wang Haiwei; Mei Longwei; Cai Xiangzhou; Chen Jingen; Guo Wei; Jiang Dazhen

    2013-01-01

    Background: Molten Salt Reactor (MSR) has several advantages over the other Generation IV reactor. Referred to the French CNRS research and compared to the fast reactor, super epithermal neutron spectrum reactor type is slightly lower and beading rate reaches 1.002. Purpose: The aim is to explore the best conversion zone layout scheme in the super epithermal neutron spectrum reactor. This study can make nuclear fuel as one way to solve the energy problems of mankind in future. Methods: Firstly, SCALE program is used for molten salt reactor graphite channel, molten salt core structure, control rods, graphite reflector and layer cladding structure. And the SMART modules are used to record the important actinides isotopes and their related reaction values of each reaction channel. Secondly, the thorium-uranium conversion rate is calculated. Finally, the better molten salt reactor core optimum layout scheme is studied comparing with various beading rates. Results: Breading zone layout scheme has an important influence on the breading rate of MSR. Central graphite channels in the core can get higher neutron flux irradiation. And more 233 Th can convert to 233 Pa, which then undergoes beta decay to become 233 U. The graphite in the breading zone gets much lower neutron flux irradiation, so the life span of this graphite can be much longer than that of others. Because neutron flux irradiation in the uranium molten salt graphite has nearly 10 times higher than the graphite in the breading zone, it has great impact on the thorium-uranium conversion rates. For the super epithermal neutron spectrum molten salt reactors, double salt design cannot get higher thorium-uranium conversion rates. The single molten salt can get the same thorium-uranium conversion rate, meanwhile it can greatly extend the life of graphite in the core. Conclusions: From the analysis of calculation results, Blanket breeding area in different locations in the core can change the breeding rates of thorium

  18. Mechanical structure and problem of thorium molten salt reactor

    International Nuclear Information System (INIS)

    Kamei, Takashi

    2011-01-01

    After Fukushima Daiichi accident, there became great interest in Thorium Molten Salt Reactor (MSR) for the safety as station blackout leading to auto drainage of molten salts with freeze valve. This article described mechanical structure of MSR and problems of materials and pipes. Material corrosion problem by molten salts would be solved using modified Hastelloy N with Ti and Nb added, which should be confirmed by operation of an experimental reactor. Trends in international activities of MSR were also referred including China declaring MSR development in January 2011 to solve thorium contamination issues at rare earth production and India rich in thorium resources. (T. Tanaka)

  19. An Analysis of Reactor Structural Response to Fuel Sodium Interaction in a Hypothetical Core Disruptive Accident

    International Nuclear Information System (INIS)

    Suzuki, K.; Tashiro, M.; Sasanuma, K.; Nagashima, K.

    1976-01-01

    This study shows the effect of constraints around FSI zone on FSI phenomena and deformations of reactor structures. SUGAR-PISCES code system has been developed to evaluate the phenomena of FSI and the response of reactor structure. SUGAR calculates the phenomena of FSI. PISCES, developed by Physics International Company in U.S.A, calculates the dynamic response of reactor structure in two-dimensional, time-dependent finite-difference Lagrangian model. The results show that the peak pressure and energy by FSI and the deformation of reactor structures are about twice in case of FSI zone surrounding by blanket than by coolant. The FSI phenomena highly depend on the reactor structure and the realistic configuration around core must be considered for analyzing hypothetical core disruptive accident. In conclusion: FSI phenomena depend highly on constraints around FSI zone, so that the constraints must be dealt with realistically in analytical models. Although a two-dimensional model is superior to a quasi-two-dimensional model. The former needs long calculation time, so it is very expensive using in parametric study. Therefore, it is desirable that the two-dimensional model is used in the final study of reactor design and the quasi-two-dimensional model is used in parametric study. The blanket affects on the acoustic pressure and the deformations of radial structures, but affects scarcely on the upper vessel deformation. The blanket also affects on the mechanical work largely. The core barrel gives scarcely the effects on pressure in single phase but gives highly the effects on pressure in two-phase and deformation of reactor structures in this study. For studying the more realistic phenomena of FSI in the reactor design, the following works should be needed. (i) Spatial Distribution of FSI Region Spatial and time-dependent distribution of fuel temperature and molten fuel fraction must be taken in realistic simulation of accident condition. To this purpose, the code will

  20. Recent developments for fast reactor structural design standard (FDS)

    International Nuclear Information System (INIS)

    Kasahara, N.; Nakamuria, K.; Morishita, M.; Shibamoto, H.; Nagashima, H.; Inoue, K.

    2005-01-01

    For realization of reliable and economical fast reactor (FR) plants, Japan Nuclear Cycle Development Institute(JNC) and Japan Atomic Power Company(JAPC) are cooperating on 'Feasibility Study on Commercialized FR Cycle Systems'. To certify the design concepts through evaluation of their structural integrity, the research and development of 'Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)' is recognized as an essential theme. FDS focuses on particular failure modes of FRs such as ratchet deformation and creep fatigue damages due to cyclic thermal loads. To evaluate these modes, three main developments are in progress. One is 'Refinement of Failure Criteria' for particular modes of FRs. Next is development of 'Guidelines for Inelastic Design Analysis' in order to predict elastic plastic and creep behaviors. Furthermore, efforts are being made toward preparing 'Guidelines for Thermal Load Modeling' for FR component design where thermal loads are dominant. These studies were performed under the sponsorship of the Ministry of Economy, Trade and Industry of Japanese government. (authors)