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Sample records for concrete containment vessel

  1. Development of prestressed concrete containment vessels

    International Nuclear Information System (INIS)

    Yuji, Hideo; Kuniyoshi, Mutsumu; Nagata, Kaoru

    1983-01-01

    This paper presents a summary of evaluations for the selection of the structural and prestressing system type to be employed for the first domestic Prestressed Concrete Containment Vessel (PCCV) in Japan. This paper also discusses characteristic features in the design of the liner plate system provided on the PCCV inner surface to assure its leak-tight integrity. Prestressed concrete containment vessels so far constructed in foreign countries are to a considerable extent of different structural types, depending on differences in dome shapes, prestressing systems and number of buttresses. These differences are caused not only by differences in design philosophy and construction practices, but also by difference in the level of technology of the times when the individual containment vessels are being constructed. In the investigation reported herein, the most suitable types of PCCV and Prestressing Systems were determined as the results of an overall comparative evaluation of data and information obtained from PCCV's so far constructed from the design, construction and cost aspects, taking into consideration the seismic criteria, available technology, construction practices, regulations and technical standards in Japan. The function of the liner plate system requires the liner to have enough deformability so that the liner deformation can be consistent with the PCCV concrete deformation. Therefore, in the design of the liner plate system a method for evaluating liner deformability was employed, instead of the stress evaluation method which is widely used in the design of ordinary structures. (author)

  2. Concrete containment vessels (CCV) for nuclear power plants, (1)

    International Nuclear Information System (INIS)

    Ibe, Yukimi; Kitajima, Masatake

    1977-01-01

    Containment vessels (CV) and the construction of concrete containment vessels (CCV) for nuclear power plants are described generally, and their use and techniques in foreign countries are illustrated, in connection with the introduction of CCV to Japanese nuclear power plants. The introduction deals with the construction plan of Japanese nuclear power plants, and with the difficulties in the steel CV for large scale construction. The investigations, tests and researches are not yet sufficient. The prompt establishment of safety supported by technical criteria, analytical methods and experiments is desired. The second part deals with the consideration for aseismatic design, construction, function and characteristics of CCV. The classification and currently employed CCV, which is mainly reinforced concrete containment vessels (RCCV), are described, and the typical CCV employed for BWR is illustrated. Further, the typical arrangement of reinforcing steels at the cylindrical portion and the dome portion of RCCV is illustrated. The third part deals with the present state of CCV abroad. A prestressed concrete containment vessel (PCCV) of Turkey Point power plant is illustrated as a typical example of CCV. The tests reported in the international meeting for the design, construction and operation of concrete pressure vessels and concrete containment vessels at York University in England in 1975 are reviewed. Typical examples of the design conditions, the size and form, and the construction procedure for PCCV and RCCV abroad are reviewed. (Iwakiri, K.)

  3. Instrumentation and testing of a prestressed concrete containment vessel model

    International Nuclear Information System (INIS)

    Hessheimer, M.F.; Pace, D.W.; Klamerus, E.W.

    1997-01-01

    Static overpressurization tests of two scale models of nuclear containment structures - a steel containment vessel (SCV) representative of an improved, boiling water reactor (BWR) Mark II design and a prestressed concrete containment vessel (PCCV) for pressurized water reactors (PWR) - are being conducted by Sandia National Laboratories for the Nuclear Power Engineering Corporation of Japan and the U.S. Nuclear Regulatory Commission. This paper discusses plans for instrumentation and testing of the PCCV model. 6 refs., 2 figs., 2 tabs

  4. Mark III Containment vessel/annulus concrete design

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  5. Capacity of Prestressed Concrete Containment Vessels with Prestressing Loss

    International Nuclear Information System (INIS)

    SMITH, JEFFREY A.

    2001-01-01

    Reduced prestressing and degradation of prestressing tendons in concrete containment vessels were investigated using finite element analysis of a typical prestressed containment vessel. The containment was analyzed during a loss of coolant accident (LOCA) with varying levels of prestress loss and with reduced tendon area. It was found that when selected hoop prestressing tendons were completely removed (as if broken) or when the area of selected hoop tendons was reduced, there was a significant impact on the ultimate capacity of the containment vessel. However, when selected hoop prestressing tendons remained, but with complete loss of prestressing, the predicted ultimate capacity was not significantly affected for this specific loss of coolant accident. Concrete cracking occurred at much lower levels for all cases. For cases where selected vertical tendons were analyzed with reduced prestressing or degradation of the tendons, there also was not a significant impact on the ultimate load carrying capacity for the specific accident analyzed. For other loading scenarios (such as seismic loading) the loss of hoop prestressing with the tendons remaining could be more significant on the ultimate capacity of the containment vessel than found for the accident analyzed. A combination of loss of prestressing and degradation of the vertical tendons could also be more critical during other loading scenarios

  6. Instrumentation of a prestressed concrete containment vessel model

    International Nuclear Information System (INIS)

    Hessheimer, M.F.; Rightley, M.J.; Matsumoto, T.

    1995-01-01

    A series of static overpressurization tests of scale models of nuclear containment structures is being conducted by Sandia National Laboratories for the Nuclear Power Engineering Corporation of Japan and the U.S. Nuclear Regulatory Commission. At present, two tests are being planned: a test of a model of a steel containment vessel (SCV) that is representative of an improved, boiling water reactor (BWR) Mark II design; and a test of a model of a prestressed concrete containment vessel (PCCV). This paper discusses plans and the results of a preliminary investigation of the instrumentation of the PCCV model. The instrumentation suite for this model will consist of approximately 2000 channels of data to record displacements, strains in the reinforcing steel, prestressing tendons, concrete, steel liner and liner anchors, as well as pressure and temperature. The instrumentation is being designed to monitor the response of the model during prestressing operations, during Structural Integrity and Integrated Leak Rate testing, and during test to failure of the model. Particular emphasis has been placed on instrumentation of the prestressing system in order to understand the behavior of the prestressing strands at design and beyond design pressure levels. Current plans are to place load cells at both ends of one third of the tendons in addition to placing strain measurement devices along the length of selected tendons. Strain measurements will be made using conventional bonded foil resistance gages and a wire resistance gage, known as a open-quotes Tensmegclose quotes reg-sign gage, specifically designed for use with seven-wire strand. The results of preliminary tests of both types of gages, in the laboratory and in a simulated model configuration, are reported and plans for instrumentation of the model are discussed

  7. Behavior of cracked concrete nuclear containment vessels during earthquakes

    International Nuclear Information System (INIS)

    Gergely, P.; Stanton, J.F.; White, R.N.

    1975-01-01

    When pressure builds up in a reinforced concrete nuclear containment shell, its cylindrical wall cracks vertically and horizontally at intervals of about five feet. If an earthquake occurs simultaneously with this pressurization, inertia forces are transmitted across the horizontal crack planes. The forces and deformations must be small enough to maintain the integrity of the steel liner. A typical containment shell has a radius of about 65 ft. and a wall thickness of about 4 ft. It is heavily reinforced with vertical, horizontal, and sometimes diagonal bars. A steel shell of about 3 / 8 in. thickness is attached to the concrete with anchors. The seismic shear forces are transmitted across the horizontal cracks by interface shear transfer (combination of shear friction and aggregate interlocking), by dowel action of the bars, and by diagonal bars if they are used. One important question in the design of such vessels is whether the diagonal bars are necessary. In the experimental portion of the current investigation several types of tests were conducted to study the load-slip characteristics of interface shear transfer under high intensity cyclic loading. In some cases external bars provided the clamping action of reinforcement, in more recent tests large diameter embedded bars were used. This presentation summarizes the analytical part of the investigation. A representative load-slip curve has been used in the analyses to assess the intensity of the stresses and deformations, and to study the importance of the variables as an aid in planning future tests

  8. Seismic analysis of a reinforced concrete containment vessel model

    International Nuclear Information System (INIS)

    Randy, James J.; Cherry, Jeffery L.; Rashid, Yusef R.; Chokshi, Nilesh

    2000-01-01

    Pre-and post-test analytical predictions of the dynamic behavior of a 1:10 scale model Reinforced Concrete Containment Vessel are presented. This model, designed and constructed by the Nuclear Power Engineering Corp., was subjected to seismic simulation tests using the high-performance shaking table at the Tadotsu Engineering Laboratory in Japan. A group of tests representing design-level and beyond-design-level ground motions were first conducted to verify design safety margins. These were followed by a series of tests in which progressively larger base motions were applied until structural failure was induced. The analysis was performed by ANATECH Corp. and Sandia National Laboratories for the US Nuclear Regulatory Commission, employing state-of-the-art finite-element software specifically developed for concrete structures. Three-dimensional time-history analyses were performed, first as pre-test blind predictions to evaluate the general capabilities of the analytical methods, and second as post-test validation of the methods and interpretation of the test result. The input data consisted of acceleration time histories for the horizontal, vertical and rotational (rocking) components, as measured by accelerometers mounted on the structure's basemat. The response data consisted of acceleration and displacement records for various points on the structure, as well as time-history records of strain gages mounted on the reinforcement. This paper reports on work in progress and presents pre-test predictions and post-test comparisons to measured data for tests simulating maximum design basis and extreme design basis earthquakes. The pre-test analyses predict the failure earthquake of the test structure to have an energy level in the range of four to five times the energy level of the safe shutdown earthquake. The post-test calculations completed so far show good agreement with measured data

  9. Containment performance evaluation of prestressed concrete containment vessels with fiber reinforcement

    Energy Technology Data Exchange (ETDEWEB)

    Choun, Young Sun; Park, Hyung Kui [Integrated Safety Assessment Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-12-15

    Fibers in concrete resist the growth of cracks and enhance the postcracking behavior of structures. The addition of fibers into a conventional reinforced concrete can improve the structural and functional performance of safety-related concrete structures in nuclear power plants. The influence of fibers on the ultimate internal pressure capacity of a prestressed concrete containment vessel (PCCV) was investigated through a comparison of the ultimate pressure capacities between conventional and fiber-reinforced PCCVs. Steel and polyamide fibers were used. The tension behaviors of conventional concrete and fiber-reinforced concrete specimens were investigated through uniaxial tension tests and their tension-stiffening models were obtained. For a PCCV reinforced with 1% volume hooked-end steel fiber, the ultimate pressure capacity increased by approximately 12% in comparison with that for a conventional PCCV. For a PCCV reinforced with 1.5% volume polyamide fiber, an increase of approximately 3% was estimated for the ultimate pressure capacity. The ultimate pressure capacity can be greatly improved by introducing steel and polyamide fibers in a conventional reinforced concrete. Steel fibers are more effective at enhancing the containment performance of a PCCV than polyamide fibers. The fiber reinforcement was shown to be more effective at a high pressure loading and a low prestress level.

  10. Containment performance evaluation of prestressed concrete containment vessels with fiber reinforcement

    International Nuclear Information System (INIS)

    Choun, Young Sun; Park, Hyung Kui

    2015-01-01

    Fibers in concrete resist the growth of cracks and enhance the postcracking behavior of structures. The addition of fibers into a conventional reinforced concrete can improve the structural and functional performance of safety-related concrete structures in nuclear power plants. The influence of fibers on the ultimate internal pressure capacity of a prestressed concrete containment vessel (PCCV) was investigated through a comparison of the ultimate pressure capacities between conventional and fiber-reinforced PCCVs. Steel and polyamide fibers were used. The tension behaviors of conventional concrete and fiber-reinforced concrete specimens were investigated through uniaxial tension tests and their tension-stiffening models were obtained. For a PCCV reinforced with 1% volume hooked-end steel fiber, the ultimate pressure capacity increased by approximately 12% in comparison with that for a conventional PCCV. For a PCCV reinforced with 1.5% volume polyamide fiber, an increase of approximately 3% was estimated for the ultimate pressure capacity. The ultimate pressure capacity can be greatly improved by introducing steel and polyamide fibers in a conventional reinforced concrete. Steel fibers are more effective at enhancing the containment performance of a PCCV than polyamide fibers. The fiber reinforcement was shown to be more effective at a high pressure loading and a low prestress level

  11. Capacity assessment of concrete containment vessels subjected to aircraft impact

    International Nuclear Information System (INIS)

    Andonov, Anton; Kostov, Marin; Iliev, Alexander

    2015-01-01

    Highlights: • An approach to assess the containment capacity to aircraft impact via fragility curves is proposed. • Momentum over Area was defined as most suitable reference parameter to describe the aircraft load. • The effect of the impact induced damages on the containment pressure capacity has been studied. • The studied containment shows no reduction of the pressure capacity for the investigated scenarios. • The effectiveness of innovative protective structure against aircraft impact has been evaluated. - Abstract: The paper describes the procedure and the results from the assessment of the vulnerability of a generic pre-stressed containment structure subjected to a large commercial aircraft impact. Impacts of Boeing 737, Boeing 767 and Boeing 747 have been considered. The containment vulnerability is expressed by fragility curves based on the results of a number of nonlinear dynamic analyses. Three reference parameters have been considered as impact intensity measure in the fragility curve definition: peak impact force (PIF), peak impact pressure (PIP) and Momentum over Area (MoA). Conclusions on the most suitable reference parameter as well on the vulnerability of such containment vessels are drawn. The influence of the aircraft impact induced damages on the containment ultimate pressure capacity is also assessed and some preliminary conclusions on this are drawn. The paper also addresses a conceptual design of a protective structure able to decrease the containment vulnerability and provide a preliminary assessment of the applicability of such concept.

  12. Capacity assessment of concrete containment vessels subjected to aircraft impact

    Energy Technology Data Exchange (ETDEWEB)

    Andonov, Anton, E-mail: anton.andonov@mottmac.com; Kostov, Marin; Iliev, Alexander

    2015-12-15

    Highlights: • An approach to assess the containment capacity to aircraft impact via fragility curves is proposed. • Momentum over Area was defined as most suitable reference parameter to describe the aircraft load. • The effect of the impact induced damages on the containment pressure capacity has been studied. • The studied containment shows no reduction of the pressure capacity for the investigated scenarios. • The effectiveness of innovative protective structure against aircraft impact has been evaluated. - Abstract: The paper describes the procedure and the results from the assessment of the vulnerability of a generic pre-stressed containment structure subjected to a large commercial aircraft impact. Impacts of Boeing 737, Boeing 767 and Boeing 747 have been considered. The containment vulnerability is expressed by fragility curves based on the results of a number of nonlinear dynamic analyses. Three reference parameters have been considered as impact intensity measure in the fragility curve definition: peak impact force (PIF), peak impact pressure (PIP) and Momentum over Area (MoA). Conclusions on the most suitable reference parameter as well on the vulnerability of such containment vessels are drawn. The influence of the aircraft impact induced damages on the containment ultimate pressure capacity is also assessed and some preliminary conclusions on this are drawn. The paper also addresses a conceptual design of a protective structure able to decrease the containment vulnerability and provide a preliminary assessment of the applicability of such concept.

  13. A three-dimensional rupture analysis of steel liners anchored to concrete pressure and containment vessels

    International Nuclear Information System (INIS)

    Bangash, Y.

    1987-01-01

    Steel liners or plates are anchored to concrete pressure and containment vessels for nuclear and offshore facilities. Due to extreme loading conditions a liner may buckle due to the pull-out or shearing of anchors from the base metal and concrete. Under certain conditions attributed to loadings, liner metal deterioration and cracking of concrete behind the liner, the liner may fail by rupture. This paper presents a three-dimensional analysis of steel-concrete elements, using finite elements analysis in which a provision is made for liner instability, anchor strength and stiffness, concrete cracking and finally liner rupture. The analysis is tested first on an octagonal slab with and without an anchored steel liner. It is then extended to concrete pressure and containment vessels. The analytical results obtained are compared well with those available from the experimental tests and other sources. (author)

  14. Containers, particularly prestressed concrete pressure vessels for nuclear reactor plants

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.; Mitterbacher, P.

    1986-01-01

    Pressure and temperature changes act on the liner, which cause differential expansion between the liner and the prestressed concrete. So that there will be no overload or damage to the liner, its anchoring or the concrete structure, cutouts are provided in the concrete at deflection positions of the steel cladding, connections and penetrations. These cut-outs are filled with inserts made of elastic or plastic material. (DG) [de

  15. Aseismic safety analysis of a prestressed concrete containment vessel for CPR1000 nuclear power plant

    Science.gov (United States)

    Yi, Ping; Wang, Qingkang; Kong, Xianjing

    2017-01-01

    The containment vessel of a nuclear power plant is the last barrier to prevent nuclear reactor radiation. Aseismic safety analysis is the key to appropriate containment vessel design. A prestressed concrete containment vessel (PCCV) model with a semi-infinite elastic foundation and practical arrangement of tendons has been established to analyze the aseismic ability of the CPR1000 PCCV structure under seismic loads and internal pressure. A method to model the prestressing tendon and its interaction with concrete was proposed and the axial force of the prestressing tendons showed that the simulation was reasonable and accurate. The numerical results show that for the concrete structure, the location of the cylinder wall bottom around the equipment hatch and near the ring beam are critical locations with large principal stress. The concrete cracks occurred at the bottom of the PCCV cylinder wall under the peak earthquake motion of 0.50 g, however the PCCV was still basically in an elastic state. Furthermore, the concrete cracks occurred around the equipment hatch under the design internal pressure of 0.4MPa, but the steel liner was still in the elastic stage and its leak-proof function soundness was verified. The results provide the basis for analysis and design of containment vessels.

  16. The permeability of concrete for reactor containment vessels

    International Nuclear Information System (INIS)

    Mills, R.H.

    1983-07-01

    Review of the literature pertaining to water, water vapour and gas transmission through concrete revealed conflicting views on the mechanisms involved and the influence of mix design parameters such as initial porosities and water/cement ratio. Consideration of the effects of ageing and of construction defects in field concrete were totally neglected in published work. Permeability data from three published papers were compared with permeability calculated according to Powers. The ratio of calculated to observed permeability varied from 40 x 10 -3 to 860 x 10 -3 for one group: from 0.17 x 10 3 to 8.6 x 10 3 in the second; and from 24 x 10 3 to 142 x 10 3 for the third. There were therefore wide discrepancies within each group of data and between groups. A bibliography was prepared and an exploratory experimental programme was mounted to determine the relative importance of key parameters such as cement type, porosity and water/cement ratio. Contrary to frequently cited references it was found that permeability of concrete was not significantly influenced by water/cement ratio when the starting porosity was constant. If water/cement ratio was held constant, however, the permeability was strongly influenced by starting porosity. It was also found that with constant water/cement ratio permeability increased with cement content. The value of fly ash and blast furnace slag in partial substitution for Portland cement is neglected in the literature but it is important since such substitutions alleviate alkali-silicate reactions. Permeability of concrete was significantly decreased by partial substitution of Portland cement with fly ash but there was no benefit in the use of blast furnace slag

  17. Three dimensional non-linear cracking analysis of prestressed concrete containment vessel

    International Nuclear Information System (INIS)

    Al-Obaid, Y.F.

    2001-01-01

    The paper gives full development of three-dimensional cracking matrices. These matrices are simulated in three-dimensional non-linear finite element analysis adopted for concrete containment vessels. The analysis includes a combination of conventional steel, the steel line r and prestressing tendons and the anisotropic stress-relations for concrete and concrete aggregate interlocking. The analysis is then extended and is linked to cracking analysis within the global finite element program OBAID. The analytical results compare well with those available from a model test. (author)

  18. Containment vessel

    International Nuclear Information System (INIS)

    Zbirohowski-Koscia, K.F.; Roberts, A.C.

    1980-01-01

    A concrete containment vessel for nuclear reactors is disclosed that is spherical and that has prestressing tendons disposed in first, second and third sets, the tendons of each set being all substantially concentric and centred around a respective one of the three orthogonal axes of the sphere; the tendons of the first set being anchored at each end at a first anchor rib running around a circumference of the vessel, the tendons of the second set being anchored at each end at a second anchor rib running around a circumference of the sphere and disposed at 90 0 to the first rib, and the tendons of the third set being anchored some to the first rib and the remainder to the second rib. (author)

  19. Evaluation of seismic shear capacity of prestressed concrete containment vessels with fiber reinforcement

    Energy Technology Data Exchange (ETDEWEB)

    Choun, Young Sun; Park, Jun Hee [Integrated Safety Assessment Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Fibers have been used in cement mixture to improve its toughness, ductility, and tensile strength, and to enhance the cracking and deformation characteristics of concrete structural members. The addition of fibers into conventional reinforced concrete can enhance the structural and functional performances of safety-related concrete structures in nuclear power plants. The effects of steel and polyamide fibers on the shear resisting capacity of a prestressed concrete containment vessel (PCCV) were investigated in this study. For a comparative evaluation between the shear performances of structural walls constructed with conventional concrete, steel fiber reinforced concrete, and polyamide fiber reinforced concrete, cyclic tests for wall specimens were conducted and hysteretic models were derived. The shear resisting capacity of a PCCV constructed with fiber reinforced concrete can be improved considerably. When steel fiber reinforced concrete contains hooked steel fibers in a volume fraction of 1.0%, the maximum lateral displacement of a PCCV can be improved by > 50%, in comparison with that of a conventional PCCV. When polyamide fiber reinforced concrete contains polyamide fibers in a volume fraction of 1.5%, the maximum lateral displacement of a PCCV can be enhanced by ∼40%. In particular, the energy dissipation capacity in a fiber reinforced PCCV can be enhanced by > 200%. The addition of fibers into conventional concrete increases the ductility and energy dissipation of wall structures significantly. Fibers can be effectively used to improve the structural performance of a PCCV subjected to strong ground motions. Steel fibers are more effective in enhancing the shear performance of a PCCV than polyamide fibers.

  20. Evaluation of seismic shear capacity of prestressed concrete containment vessels with fiber reinforcement

    International Nuclear Information System (INIS)

    Choun, Young Sun; Park, Jun Hee

    2015-01-01

    Fibers have been used in cement mixture to improve its toughness, ductility, and tensile strength, and to enhance the cracking and deformation characteristics of concrete structural members. The addition of fibers into conventional reinforced concrete can enhance the structural and functional performances of safety-related concrete structures in nuclear power plants. The effects of steel and polyamide fibers on the shear resisting capacity of a prestressed concrete containment vessel (PCCV) were investigated in this study. For a comparative evaluation between the shear performances of structural walls constructed with conventional concrete, steel fiber reinforced concrete, and polyamide fiber reinforced concrete, cyclic tests for wall specimens were conducted and hysteretic models were derived. The shear resisting capacity of a PCCV constructed with fiber reinforced concrete can be improved considerably. When steel fiber reinforced concrete contains hooked steel fibers in a volume fraction of 1.0%, the maximum lateral displacement of a PCCV can be improved by > 50%, in comparison with that of a conventional PCCV. When polyamide fiber reinforced concrete contains polyamide fibers in a volume fraction of 1.5%, the maximum lateral displacement of a PCCV can be enhanced by ∼40%. In particular, the energy dissipation capacity in a fiber reinforced PCCV can be enhanced by > 200%. The addition of fibers into conventional concrete increases the ductility and energy dissipation of wall structures significantly. Fibers can be effectively used to improve the structural performance of a PCCV subjected to strong ground motions. Steel fibers are more effective in enhancing the shear performance of a PCCV than polyamide fibers

  1. Ultimate Pressure Capacity of Prestressed Concrete Containment Vessels with Steel Fibers

    Energy Technology Data Exchange (ETDEWEB)

    Hahm, Dae Gi; Choun, Young Sun; Choi, In Kil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    The ultimate pressure capacity (UPC) of the prestressed concrete containment vessel (PCCV) is very important since the PCCV are final protection to prevent the massive leakage of a radioactive contaminant caused by the severe accident of nuclear power plants (NPPs). The tensile behavior of a concrete is an important factor which influence to the UPC of PCCVs. Hence, nowadays, it is interested that the application of the steel fiber to the PCCVs since that the concrete with steel fiber shows an improved performance in the tensile behavior compared to reinforced concrete (RC). In this study, we performed the UPC analysis of PCCVs with steel fibers corresponding to the different volume ratio of fibers to verify the effectiveness of steel fibers on PCCVs

  2. Experimental study of the structural behavior of the reinforced concrete containment vessel beyond design pressure

    International Nuclear Information System (INIS)

    Oyamada, O.; Saito, H.; Muramatsu, Y.; Hasegawa, T.; Tanaka, N.

    1990-01-01

    The first Advanced Boiling Water Reactor (ABWR) including a reinforced concrete containment vessel (RCCV) is scheduled to be constructed in the 1990s, in Japan. As the RCCV is new to Japan, we performed a trial design, several series of fundamental experiments and partial/total model experiments. This paper presents a summary of the 'TOP SLAB EXPERIMENT' carried out as one of partial model experiments, in which the structural behavior of the RCCV was examined under internal pressure. (orig.)

  3. Pretest round robin analysis of 1:4-scale prestressed concrete containment vessel model

    International Nuclear Information System (INIS)

    Hessheimer, M.F.; Luk, V.K.; Klamerus, E.W.; Shibata, S.; Mitsugi, S.; Costello, J.F.

    2001-01-01

    The work reported herein represents, arguably, the state of the art in the numerical simulation of the response of a prestressed concrete containment vessel (PCCV) model to pressure loads up to failure. A significant expenditure of time and money on the part of the sponsors, contractors, and Round Robin participants was required to meet the objectives. While it is difficult to summarize the results of this extraordinary effort in a few paragraphs, the following observations are offered for the reader's consideration: almost half the participants used ABAQUS as the primary computational tool for performing the pretest analyses. The other participants used a variety of codes, most of which were developed ''in house''. (author)

  4. Preliminary analysis of a 1:4 scale prestressed concrete containment vessel model

    International Nuclear Information System (INIS)

    Dameron, R.A.; Rashid, Y.R.; Luk, V.K.; Hessheimer, M.F.

    1997-01-01

    Sandia National Laboratories is conducting a research program to investigate the integrity of nuclear containment structures. As part of the program Sandia will construct an instrumented 1:4 scale model of a prestressed concrete containment vessel (PCCV) for pressurized water reactors (PWR), which will be pressure tested up to its ultimate capacity. One of the key program objectives is to develop validated methods to predict the structural performance of containment vessels when subjected to beyond design basis loadings. Analytical prediction of structural performance requires a stepwise, systematic approach that addresses all potential failure modes. The analysis effort includes two and three-dimensional nonlinear finite element analyses of the PCCV test model to evaluate its structural performance under very high internal pressurization. Such analyses have been performed using the nonlinear concrete constitutive model, ANACAP-U, in conjunction with the ABAQUS general purpose finite element code. The analysis effort is carried out in three phases: preliminary analysis; pretest prediction; and post-test data interpretation and analysis evaluation. The preliminary analysis phase serves to provide instrumentation support and identify candidate failure modes. The associated tasks include the preliminary prediction of failure pressure and probable failure locations and the development of models to be used in the detailed failure analyses. This paper describes the modeling approaches and some of the results obtained in the first phase of the analysis effort

  5. Plan on test to failure of a prestressed concrete containment vessel model

    International Nuclear Information System (INIS)

    Takumi, K.; Nonaka, A.; Umeki, K.; Nagata, K.; Soejima, M.; Yamaura, Y.; Costello, J.F.; Riesemann, W.A. von.; Parks, M.B.; Horschel, D.S.

    1992-01-01

    A summary of the plans to test a prestressed concrete containment vessel (PCCV) model to failure is provided in this paper. The test will be conducted as a part of a joint research program between the Nuclear Power Engineering Corporation (NUPEC), the United States Nuclear Regulatory Commission (NRC), and Sandia National Laboratories (SNL). The containment model will be a scaled representation of a PCCV for a pressurized water reactor (PWR). During the test, the model will be slowly pressurized internally until failure of the containment pressure boundary occurs. The objectives of the test are to measure the failure pressure, to observe the mode of failure, and to record the containment structural response up to failure. Pre- and posttest analyses will be conducted to forecast and evaluate the test results. Based on these results, a validated method for evaluating the structural behavior of an actual PWR PCCV will be developed. The concepts to design the PCCV model are also described in the paper

  6. Measured Prestress Loss of over 20-Year-Old Prestressed Concrete Containment Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Hahm, Dae Gi; Choun, Young Sun; Choi, In Kil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Most nuclear reactors, both in Korea and worldwide, are enclosed by a prestressed concrete containment vessels(PCCVs). The containment wall is approximately 1 m thick and is prestressed in two directions by large prestressing tendons. The main purpose of the containment is to maintain the structural integrity of the containment in the event of a major internal accident. The main accidental scenario, which the containment is designed to withstand, is a so-called loss of coolant accident (LOCA). A LOCA is initiated by a pipe rupture in the cooling system, discharging hot steam into the containment. The escape of steam increases both the temperature and pressure inside the containment. The increased internal pressure arising from a LOCA is referred to as the design pressure. The prestressing system is designed to counterbalance the tensile forces arising from the design pressure. The status of the containment is gradually changed due to environmental factors and by alterations in the micro structure of the material. The prestress will be reduced due to shrinkage and creep in the concrete and relaxation in the tendons. The corrosion protection of tendons are for Korean containments arranged in two different ways, either by cement grouting (bonded tendons) or e.g. by grease injection (unbonded tendons). The major advantage using unbonded tendons is the possibilities of assessing their status (e.g. prestress losses or corrosion damages) which is not possible using bonded tendons. Both bonded and unbonded tendons are used worldwide. For example in the U.S. almost all tendons are unbonded, whereas in France almost all tendons are bonded. For Korean reactor containments with unbonded tendons (14 containments) the tendon force is monitored at regular in-service inspections. The power plant Wolsung in Korea has bonded tendons and several prestressed concrete beams were constructed with the single purpose to follow up the prestress losses. The remaining tendon forces in some

  7. Measured Prestress Loss of over 20-Year-Old Prestressed Concrete Containment Vessels

    International Nuclear Information System (INIS)

    Hahm, Dae Gi; Choun, Young Sun; Choi, In Kil

    2010-01-01

    Most nuclear reactors, both in Korea and worldwide, are enclosed by a prestressed concrete containment vessels(PCCVs). The containment wall is approximately 1 m thick and is prestressed in two directions by large prestressing tendons. The main purpose of the containment is to maintain the structural integrity of the containment in the event of a major internal accident. The main accidental scenario, which the containment is designed to withstand, is a so-called loss of coolant accident (LOCA). A LOCA is initiated by a pipe rupture in the cooling system, discharging hot steam into the containment. The escape of steam increases both the temperature and pressure inside the containment. The increased internal pressure arising from a LOCA is referred to as the design pressure. The prestressing system is designed to counterbalance the tensile forces arising from the design pressure. The status of the containment is gradually changed due to environmental factors and by alterations in the micro structure of the material. The prestress will be reduced due to shrinkage and creep in the concrete and relaxation in the tendons. The corrosion protection of tendons are for Korean containments arranged in two different ways, either by cement grouting (bonded tendons) or e.g. by grease injection (unbonded tendons). The major advantage using unbonded tendons is the possibilities of assessing their status (e.g. prestress losses or corrosion damages) which is not possible using bonded tendons. Both bonded and unbonded tendons are used worldwide. For example in the U.S. almost all tendons are unbonded, whereas in France almost all tendons are bonded. For Korean reactor containments with unbonded tendons (14 containments) the tendon force is monitored at regular in-service inspections. The power plant Wolsung in Korea has bonded tendons and several prestressed concrete beams were constructed with the single purpose to follow up the prestress losses. The remaining tendon forces in some

  8. Nonlinear analysis of pre-stressed concrete containment vessel (PCCV) using the damage plasticity model

    Energy Technology Data Exchange (ETDEWEB)

    Shokoohfar, Ahmad; Rahai, Alireza, E-mail: rahai@aut.ac.ir

    2016-03-15

    Highlights: • This paper describes nonlinear analyses of a 1:4 scale model of a (PCCV). • Coupled temp-disp. analysis and concrete damage plasticity are considered. • Temperature has limited effects on correct failure mode estimation. • Higher pre-stressing forces have limited effects on ultimate radial displacements. • Anchorage details of liner plates leads to prediction of correct failure mode. - Abstract: This paper describes the nonlinear analyses of a 1:4 scale model of a pre-stressed concrete containment vessel (PCCV). The analyses are performed under pressure and high temperature effects with considering anchorage details of liner plate. The temperature-time history of the model test is considered as an input boundary condition in the coupled temp-displacement analysis. The constitutive model developed by Chang and Mander (1994) is adopted in the model as the basis for the concrete stress–strain relation. To trace the crack pattern of the PCCV concrete faces, the concrete damage plasticity model is applied. This study includes the results of the thermal and mechanical behaviors of the PCCV subject to temperature loading and internal pressure at the same time. The test results are compared with the analysis results. The analysis results show that the temperature has little impact on the ultimate pressure capacity of the PCCV. To simulate the exact failure mode of the PCCV, the anchorage details of the liner plates around openings should be maintained in the analytical models. Also the failure mode of the PCCV structure hasn’t influenced by hoop tendons pre-stressing force variations.

  9. Effects of no stiffness inside unbonded tendon ducts on the behavior of prestressd concrete containment vessels

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Sang Hoon; Kwak, Hyo Gyong [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Jung, Rae Young; Noh, Sang Hoon [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-06-15

    The numerical simulation methodologies to evaluate the structural behaviors of prestressed concrete containment vessels (PCCVs) have been substantially developed in recent decades. However, there remain several issues to be investigated more closely to narrow the gap between test results and numerical simulations. As one of those issues, the effects of no stiffness inside unbonded tendon ducts on the behavior of PCCVs are investigated in this study. Duct holes for prestressing cables' passing are provided inside the containment wall and dome in one to three directions for general PCCVs. The specific stress distribution along the periphery of the prestressing duct hole and the loss of stiffness inside the hole, especially in an unbonded tendon system, are usually neglected in the analysis of PCCVs with the assumption that the duct hole is filled with concrete. However, duct holes are not small enough to be neglected. In this study, the effects of no stiffness inside the unbonded tendon system on the behaviors of PCCVs are evaluated using both analytical and numerical approaches. From the results, the effects of no stiffness in unbonded tendons need to be considered in numerical simulations for PCCVs, especially under internal pressure loading.

  10. Effects of no stiffness inside unbonded tendon ducts on the behavior of prestressd concrete containment vessels

    International Nuclear Information System (INIS)

    Noh, Sang Hoon; Kwak, Hyo Gyong; Jung, Rae Young; Noh, Sang Hoon

    2016-01-01

    The numerical simulation methodologies to evaluate the structural behaviors of prestressed concrete containment vessels (PCCVs) have been substantially developed in recent decades. However, there remain several issues to be investigated more closely to narrow the gap between test results and numerical simulations. As one of those issues, the effects of no stiffness inside unbonded tendon ducts on the behavior of PCCVs are investigated in this study. Duct holes for prestressing cables' passing are provided inside the containment wall and dome in one to three directions for general PCCVs. The specific stress distribution along the periphery of the prestressing duct hole and the loss of stiffness inside the hole, especially in an unbonded tendon system, are usually neglected in the analysis of PCCVs with the assumption that the duct hole is filled with concrete. However, duct holes are not small enough to be neglected. In this study, the effects of no stiffness inside the unbonded tendon system on the behaviors of PCCVs are evaluated using both analytical and numerical approaches. From the results, the effects of no stiffness in unbonded tendons need to be considered in numerical simulations for PCCVs, especially under internal pressure loading

  11. Parametric Study on Important Variables of Aircraft Impact to Prestressed Concrete Containment Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Sangshup; Hahm, Daegi; Choi, Inkil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    In this paper, to find the damage parameter, it is necessary to use many analysis cases and the time reduction. Thus, this paper uses a revised version of Riera's method. Using this method, the response has been found a Prestressed Concrete Containments Vessels (PCCVs) subject to impact loading, and the results of the velocity and mass of the important parameters have been analyzed. To find the response of the PCCVs subjected to aircraft impact load, it is made that a variable forcing functions depending on the velocity and fuel in the paper. The velocity variation affects more than fuel percentage, and we expect that the severe damage of the PCCVs with the same material properties is subject to aircraft impact load (more than 200m/s and 70%)

  12. Safety margin evaluation of pre-stressed concrete nuclear containment vessel model with BARC code ULCA

    International Nuclear Information System (INIS)

    Basha, S.M.; Patnaik, R.; Ramanujam, S.; Singh, R.K.; Kushwaha, H.S.; Venkat Raj, V.

    2002-01-01

    Full text: Ultimate load capacity assessment of nuclear containments has been a thrust research area for Indian pressurised heavy water reactor (PHWR) power programme. For containment safety assessment of Indian PHWRs a finite element code ULCA was developed at BARC, Trombay. This code has been extensively benchmarked with experimental results and for prediction of safety margins of Indian PHWRs. The present paper highlights the analysis results for prestressed concrete containment vessel (PCCV) tested at Sandia National Labs, USA in a round robin analysis activity co-sponsored by Nuclear Power Engineering Corporation (NUPEC), Japan and the U.S Nuclear Regulatory Commission (NRC). Three levels of failure pressure predictions namely the upper bound, the most probable and the lower bound (all with 90% confidence) were made as per the requirements of the round robin analysis activity. The most likely failure pressure is predicted to be in the range of 2.95 Pd to 3.15 Pd (Pd = design pressure of 0.39 MPa for the PCCV model) depending on the type of liners used in the construction of the PCCV model. The lower bound value of the ultimate pressure of 2.80 Pd and the upper bound of the ultimate pressure of 3.45 Pd are also predicted from the analysis. These limiting values depend on the assumptions of the analysis for simulating the concrete tendon interaction and the strain hardening characteristics of the steel members. The experimental test has been recently concluded at Sandia Laboratory and the peak pressure reached during the test is 3.3 Pd that is enveloped by our upper bound prediction of 3.45 Pd and is close to the predicted most likely pressure of 3.15 Pd

  13. Posttest analysis of a 1:4-scale prestressed concrete containment vessel model

    International Nuclear Information System (INIS)

    Dameron, R.A.; Rashid, Y.R.; Hessheimer, M.F.

    2003-01-01

    The Nuclear Power Engineering Corporation (NUPEC) of Japan and the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research, co-sponsored a Cooperative Containment Research Program at Sandia National Laboratories (SNL) in Albuquerque, New Mexico. As part of the program, a prestressed concrete containment vessel (PCCV) model was subjected to a series of overpressurization tests at SNL beginning in July 2000 and culminating in a functional failure mode or Limit State Test (LST) in September 2000 and a Structural Failure Mode Test (SFMT) in November 2001. The PCCV model, uniformly scaled at 1:4, is representative of the containment structure of an actual Pressurized Water Reactor (PWR) plant (OHI-3) in Japan. The objectives of the pressurization tests were to obtain measurement of the structural response to pressure loading beyond design basis accident in order to validate analytical modeling, to find pressure capacity of the model, and to observe its failure mechanisms. This paper compares results of pretest analytical studies of the PCCV model to the PCCV high pressure test measurements and describes results of post-test analytical studies. These analyses have been performed by ANATECH Corp. under contract with Sandia National Laboratories. The post-test analysis represents the third phase of a comprehensive PCCV analysis effort. The first phase consisted of preliminary analyses to determine what finite element models would be necessary for the pretest prediction analyses, and the second phase consisted of the pretest prediction analyses. The principal objectives of the post-test analyses were: (1) to provide insights to improve the analytical methods for predicting the structural response and failure modes of a prestressed concrete containment, and (2) to evaluate by analysis any phenomena or failure mode observed during the test that had not been explicitly predicted by analysis. In addition to summarizing comparisons between measured

  14. An international survey of in-service inspection experience with prestressed concrete pressure vessels and containments for nuclear reactors

    International Nuclear Information System (INIS)

    1982-04-01

    An international survey is presented of experience obtained from the in-service surveillance of prestressed concrete pressure vessels and containments for nuclear reactors. Some information on other prestressed concrete structures is also given. Experience has been gained during the working life of such structures in Western Europe and the USA over the years since 1967. For each country a summary is given of the nuclear programme, national standards and Codes of Practice, and the detailed in-service inspection programme. Reports are then given of the actual experience obtained from the inspection programme and the methods of measurement, examination and reporting employed in each country. A comprehensive bibliography of over 100 references is included. The appendices contain information on nuclear power stations which are operating, under construction or planned worldwide and which employ either prestressed concrete pressure vessels or containments. (U.K.)

  15. Analysis and application of prestressed concrete reactor vessels for LMFBR containment

    International Nuclear Information System (INIS)

    Marchertas, A.H.; Fistedis, S.H.; Bazant, Z.P.; Belytschko, T.B.

    1978-01-01

    An analytical model of a prestressed concrete reactor vessel (PCRV) for LMFBR and the associated finite element computer code, involving an explicit time integration procedure, is described. The model is axisymmetric and includes simulations of the tensile cracking of concrete, the reinforcement, and a prestressing capability. The tensile cracking of concrete and the steel reinforcement are both modeled as continuously distributed within the finite element. The stresses in the reinforcement and concrete are computed separately and combined to give an overall stress state of the composite material. Attention is given to the fact that cracks do not form instantaneously, but develop gradually. Thus, after crack initiation the normal stress is reduced to zero gradually as a function of time. Residual shear resistance of cracks due to aggregate interlock is also taken into account. Prestressing of the PCRV is modeled by special structural members which represent an averaged prestressing layer equivalent to an axisymmetric shell. The internal prestressing members are superimposed over the reinforced concrete body of the PCRV; they are permitted to stretch and slide in a predetermined path, simulating the actual tendons. The validity of the code is examined by comparison with experimental data. (Auth.)

  16. BBRV post-tensioning systems as applied to reactor containments and prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Thorpe, W.; Speck, F.E.

    1976-01-01

    Nuclear containments and pressure vessels can be post-tensioned by using two basically different methods: tendons and winding. The fundamental differences between the two concepts are shown by introductory examples. A discussion of tendon units, usually lying in the range 4000 to 10,000 kN, is followed by a detailed presentation of the BBRV winding system. After giving a short comment to factors influencing the choice of a post-tensioning system the authors discuss specific aspects of some application groups: cable layout with containments and pressure vessels, conditions for a wrapped design, corrosion protection. (author)

  17. Durability and safety of concrete structures in the nuclear context. The case of the containment vessel

    Energy Technology Data Exchange (ETDEWEB)

    Torrenti, J.M. [Universite Paris Est, LCPC (France); Nahas, G. [IRSN/DSR (France)

    2011-07-01

    The durability of structures, because of its economic and environmental implications, is one of the actual hot topics in civil engineering. In the field of nuclear energy, we are facing very challenging problems like: how could we prolong the service life of actual nuclear containments and how can we assure the durability of a radioactive storage on the very long term (several centuries)? These already difficult questions in a classical civil engineering view are even more complicated in the field of nuclear energy where the structures are massive and the safety of the installations has to be considered. For the containment of nuclear power plants, these stakes will be lit with some examples of research concerning the mechanical behaviour of concrete and concrete structures (at early age, in service on long scales of time and in the event of an accident), the durability of the concrete structures (leaching, swelling due to delayed ettringite formation - DEF -) and the couplings between mechanics and durability. Finally, the importance of probabilistic aspects and the inherent difficulties will be shown. (authors)

  18. Durability and safety of concrete structures in the nuclear context. The case of the containment vessel

    International Nuclear Information System (INIS)

    Torrenti, J.M.; Nahas, G.

    2011-01-01

    The durability of structures, because of its economic and environmental implications, is one of the actual hot topics in civil engineering. In the field of nuclear energy, we are facing very challenging problems like: how could we prolong the service life of actual nuclear containments and how can we assure the durability of a radioactive storage on the very long term (several centuries)? These already difficult questions in a classical civil engineering view are even more complicated in the field of nuclear energy where the structures are massive and the safety of the installations has to be considered. For the containment of nuclear power plants, these stakes will be lit with some examples of research concerning the mechanical behaviour of concrete and concrete structures (at early age, in service on long scales of time and in the event of an accident), the durability of the concrete structures (leaching, swelling due to delayed ettringite formation - DEF -) and the couplings between mechanics and durability. Finally, the importance of probabilistic aspects and the inherent difficulties will be shown. (authors)

  19. Nuclear Power Plant Prestressed Concrete Containment Vessel Structure Monitoring during Integrated Leakage Rate Testing Using Fiber Bragg Grating Sensors

    Directory of Open Access Journals (Sweden)

    Jinke Li

    2017-04-01

    Full Text Available As the last barrier of nuclear reactor, prestressed concrete containment vessels (PCCVs play an important role in nuclear power plants (NPPs. To test the mechanical property of PCCV during the integrated leakage rate testing (ILRT, a fiber Bragg grating (FBG sensor was used to monitor concrete strain. In addition, a finite element method (FEM model was built to simulate the progress of the ILRT. The results showed that the strain monitored by FBG had the same trend compared to the inner pressure variation. The calculation results showed a similar trend compared with the monitoring results and provided much information about the locations in which the strain sensors should be installed. Therefore, it is confirmed that FBG sensors and FEM simulation are very useful in PCCV structure monitoring.

  20. Study of the concrete tensile creep: application for the containment vessel of the nuclear power plants (PWR)

    International Nuclear Information System (INIS)

    Reviron, Nanthilde

    2009-01-01

    The aim of this work is to study experimentally and to conduct numerical simulations on the creep of concrete subjected to tensile stresses. The main purpose is to predict the behaviour of containment vessels of nuclear power plants (PWR) in the case of decennial test or accident. In order to satisfy to these industrial needs, it is necessary to characterize the behaviour of concrete under uniaxial tension. Thus, an important experimental study of tensile creep in concrete has been performed for different loading levels (50%, 70% and 90% of the tensile strength). In these tests, load was kept constant during 3 days. Several tests were performed: measurements of elastic properties and strength (in tension and in compression), monitoring of drying, shrinkage, basic creep and drying creep strains. Moreover, compressive creep tests were also performed and showed a difference with tensile creep. Furthermore, decrease of tensile strength and failure under tensile creep for large loading levels were observed. A numerical model has been proposed and developed in Cast3m finite element code. (author)

  1. Viscoelastic and thermal behavior of structural concrete with reference to containment vessels

    International Nuclear Information System (INIS)

    Stefanou, G.D.

    1981-01-01

    A method of numerical viscoelastic stress analysis is described suitable for concrete structures operating at elevated temperatures. The paper describes how approximate numerical methods of elastic analysis of the finite element type can be extended to incorporate the viscoelastic behavior of structural concrete of the quasi-static type. A new eight parameter viscoelastic model is proposed to represent concrete behavior in the loaded and unloaded stage. The deformational expressions for the proposed viscoelastic analogue are also developed. Finally, as a result of courve-fitting procedures, the evaluation of the creep law coefficients are obtained for creep laws appropriate to a test regime. The proposed method is of general application providing that the properties of concrete are assessed reasonably well. The analytical predictions are compared with experimental results obtained on concrete model specimens loaded for 3 1/2 months, at a temperature of 80 0 C. (author)

  2. Process for producing curved surface of membrane rings for large containers, particulary for prestressed concrete pressure vessels of nuclear reactors

    International Nuclear Information System (INIS)

    Kumpf, H.

    1977-01-01

    Membrane rings for large pressure vessels, particularly for prestressed-concrete pressure vessels, often have curved surfaces. The invention describes a process of producing these at site, which is particularly advantageous as the forming and installation of the vessel component coincide. According to the invention, the originally flat membrane ring is set in a predetermined position, is then pressed in sections by a forming tool (with a preformed support ring as opposite tool), and shaped. After this, the shaped parts are welded to the ring-shaped wall parts of the large vessel. The manufacture of single and double membrane rings arrangements is described. (HP) [de

  3. A new model for anisotropic damage in concrete and its application to the prediction of failure of some containment vessel

    International Nuclear Information System (INIS)

    Badel, P.-B.; Godard, V.; Leblond, J.-B.

    2005-01-01

    The aim of this paper is to propose a new model for damage in concrete structures which incorporates such complex features as damage anisotropy and asymmetry between tension and compression, while being expressed in a format well suited for numerical applications and involving a limited number of material parameters which can be determined from standard experiments. A crude version of the model involving a single tonsorial internal variable representing damage in tension, and a single material parameter, is presented first. The predictions of this simple model are satisfactory in simple tension, but not so in simple compression. As a remedy, various refinements are then introduced in a second version of the model involving an additional tonsorial or scalar internal variable representing damage in compression, and five additional material parameters. An example of determination of the model parameters using experimental stress-strain curves in simple tension and compression, plus failure envelopes in biaxial tension/compression, is presented next. The model is finally applied to the numerical prediction of the failure of some containment vessel subjected to some large internal pressure, with a comparison with calculations based on a simpler isotropic variant of the model using a single scalar damage variable. The results illustrate the relevance of models incorporating both asymmetry between tension and compression and anisotropy of damage for simulations of industrial concrete structures. (authors)

  4. Assessment of Ultimate Load Capacity for Pre-Stressed Concrete Containment Vessel Model of PWR Design With BARC Code ULCA

    International Nuclear Information System (INIS)

    Basha, S.M.; Singh, R.K.; Patnaik, R.; Ramanujam, S.; Kushwaha, H.S.; Venkat Raj, V.

    2002-01-01

    Ultimate load capacity assessment of nuclear containments has been a thrust research area for Indian Pressurised Heavy Water Reactor (PHWR) power programme. For containment safety assessment of Indian PHWRs a finite element code ULCA was developed at BARC, Trombay. This code has been extensively benchmarked with experimental results. The present paper highlights the analysis results for Prestressed Concrete Containment Vessel (PCCV) tested at Sandia National Labs, USA in a Round Robin analysis activity co-sponsored by Nuclear Power Engineering Corporation (NUPEC), Japan and the U.S Nuclear Regulatory Commission (NRC). Three levels of failure pressure predictions namely the upper bound, the most probable and the lower bound (all with 90% confidence) were made as per the requirements of the round robin analysis activity. The most likely failure pressure is predicted to be in the range of 2.95 Pd to 3.15 Pd (Pd= design pressure of 0.39 MPa for the PCCV model) depending on the type of liners used in the construction of the PCCV model. The lower bound value of the ultimate pressure of 2.80 Pd and the upper bound of the ultimate pressure of 3.45 Pd are also predicted from the analysis. These limiting values depend on the assumptions of the analysis for simulating the concrete-tendon interaction and the strain hardening characteristics of the steel members. The experimental test has been recently concluded at Sandia Laboratory and the peak pressure reached during the test is 3.3 Pd that is enveloped by our upper bound prediction of 3.45 Pd and is close to the predicted most likely pressure of 3.15 Pd. (authors)

  5. A prediction method for long-term behavior of prestressed concrete containment vessels

    International Nuclear Information System (INIS)

    Ozaki, M.; Abe, T.; Watanabe, Y.; Kato, A.; Yamaguchi, T.; Yamamoto, M.

    1995-01-01

    This paper presents results of studies on the long-term behavior of PCCVs at Taruga Unit No 2 and Ohi Unit No 3/4 power stations. The objective of this study is to evaluate the measured strain in the concrete and reduction force in the tendons, and to establish the prediction methods for long-term PCCVs behavior. Comparing the measured strains with those calculated due to creep and shrinkage of the concrete, those in contrast were investigated. Furthermore, the reduced tendon forces are calculated considering losses in elasticity, relaxation, creep and shrinkage. The measured reduction in the tendon forces is compared with the calculated. Considering changes in temperature and humidity, the measured strains and tendon forces were in good agreement with those calculated. From the above results, it was confirmed that the residual pre stresses in the PCCVs maintain the predicted values at the design stage, and that the prediction method of long-term behaviors has sufficient reliability. (author). 10 refs., 8 figs., 3 tabs

  6. Development of ultrasonic testing technique with the large transducer to inspect the containment vessel plates of nuclear power plant embedded in concrete

    International Nuclear Information System (INIS)

    Ishida, Hitoshi; Kurozumi, Yasuo; Kaneshima, Yoshiari

    2004-01-01

    The containment vessel plates embedded in concrete on Pressurized Water Reactors are inaccessible to inspect directly. Therefore, it is advisable to prepare inspection technology to detect existence and a location of corrosion on the embedded plates indirectly. In order to establish ultrasonic testing technique to be able to inspect the containment vessel plates embedded in concrete widely at the accessible point, experiments to detect artificial hollows simulating corrosion on a surface of a carbon steel plate mock-up covered with concrete simulating the embedded containment vessel plates were carried out with newly made ultrasonic transducers. We made newly low frequency (0.3 MHz and 0.5 MHz) surface shear horizontal (SH) wave transducers combined with three large active elements, which were equivalent to a 120mm width element. As a result of the experiments, the surface SH transducers could detect clearly the echo from the hollows with a depth of 9.5 mm and 19 mm at a distance of 1500mm from the transducers on the surface of the mock-up covered with concrete. Therefore, we evaluate that it is possible to detect the defects such as corrosion on the plates embedded in concrete with the newly made low frequency surface SH transducers with large elements. (author)

  7. Structural integrity test of prestressed concrete containment vessel for Tsuruga Unit No. 2 Nuclear Power Station

    International Nuclear Information System (INIS)

    Tamura, S.; Nagata, K.; Takeda, T.; Yamaguchi, T.; Nakayama, T.

    1987-01-01

    In introducing the PCCV to Japan, various verification tests were carried out to understand the structural performance of the PCCV and confirm the reliability of its design. In addition to those tests, a Structural Integrity Test (SIT) was conducted in Feb. 1986 as a final acceptance test. This report discusses the results of the SIT on the PCCV. The test was carried out simultaneously with an Integrated Leak Rate Test (ILRT) under the same pressure sequence. 1) Pressure-displacement relationships and pressure-strain relationships were more or less linear. 2) The measured displacement values at the maximum pressure (4.5 kgf/cm 2 G) corresponded well with calculated values. Correspondence with converted displacement obtained from strain and measured displacement was also good. 3) The residual displacement when 24 hours had elapsed after completion of depressurization was not more than 10% of the displacement at the maximum pressure. 4) The variation in tendon force at the maximum pressure is smaller than the calculated value in proportion to the elongation of the PCCV. 5) Although fine surface cracks due to shrinkage of concrete were seen, new structural cracks due to pressure were not observed. The leakage rate was evaluated at 0.016% of volume per day. It is much smaller than the design value of 0.1% of volume per day. (orig./HP)

  8. Analysis study on change of tendon behavior during pressurization process of Pre-stressed Concrete Containment Vessel

    International Nuclear Information System (INIS)

    Kashiwase, Takako; Nagasaka, Hideo

    1999-01-01

    NUPEC has been planning the ultimate strength test of Pre-stressed Concrete Containment Vessel (PCCV). The test model is 1/4 uniform scale model of Japan actual PCCV. It involves an equipment hatch, several penetrations and liner with T-anchors. The ancillary test for the PCCV test was conducted, in which friction coefficient of hoop tendon was evaluated by tensile force distribution using the same tendon as that of 1/4 PCCV model. Tendon will be in plastic region under internal pressure above 3.5 times design pressure (Pd) and surface characteristic of tendon and the resultant friction coefficient will be changed. In the present paper, tendon friction coefficient in the plastic region was obtained by evaluating plastic region data of tendon in the ancillary test. The validity of the obtained friction coefficient was confirmed by the tendon elongation data. In addition to the formally developed elastic region friction coefficient, the obtained plastic region correlation was incorporated into ABAQUS Ver. 5.6. The effect of tendon tensile force distribution change on structural behavior up to 3.8 Pd was evaluated. (author)

  9. Fibre-concrete container

    International Nuclear Information System (INIS)

    2000-01-01

    In this leaflet the fibre-concrete container for radioactive wastes is described. The fibre container is made of fibre-concrete that contains cement, aggregate, sand, filter, flame-silica, super-plastificator, water and scattered metal fibres. The fibre-concrete container has a dice shape with outer dimension 1.7 x 1.7 x 1.7 m. It is mounted of a container body, a container cover and two caps. Total weight of container is 4,240 kg, maximum weight of loaded container do not must exceed 15,000 kg. The physical and mechanical properties of the fibre-concrete container are described in detail. The fibre-concrete container manufactured for storing of low and intermediate radioactive wastes. A fibre-concrete container utilization to store of radioactive wastes solves these problems: increase of stability of stored packages of radioactive waste; watertightness within 300 years at least; static stability of bearing space; better utilization of bearing spaces; insulation of radioactive waste in a case of seismic and geological event; increase of fire resistance; and transport of radioactive waste

  10. Containment vessel drain system

    Science.gov (United States)

    Harris, Scott G.

    2018-01-30

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

  11. Recent investigations and tests with the BBR winding system for circumferential prestressing of concrete vessels and containments

    International Nuclear Information System (INIS)

    Schuett, K.; Speck, F.E.

    1993-01-01

    Prestressed concrete pressure vessels for nuclear power stations need post-tensioning systems of large capacity. For the circumferential prestressing, the continuous winding of prestressing steel has several advantages when compared to the use of large numbers of single tendons. About 15 years ago Bureau BBR Ltd (Zuerich) developed the winding system SW 8500. The further development work interrupted at that time for lack of immediate applications was resumed 4 years ago by Bureau BBR together with SUSPA on the ground of new projects being evaluated

  12. Containment vessel design and practice

    International Nuclear Information System (INIS)

    Bangash, Y.

    1983-01-01

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

  13. Method of producing the arched surfaces of diaphragm rings for large containers, especially for prestressed-concrete pressure vessels of nuclear reactors

    International Nuclear Information System (INIS)

    Kumpf, H.

    1976-01-01

    In producing arched surfaces of diaphragm rings for large containers, especially for prestressed-concrete pressure vessels for nuclear power plants, it is of advantage to manufacture these directly on the construction site. According to the invention the, at first level, diaphragm ring is put on the predetermined place, sectionally pressed against and shaped by a shaping tool - with a profiled supporting ring as a counter-acting tool - and afterwards welded together with the annular wall sections of the large container along the shaped parts. The manufacture of single and double configurations of diaphragm rings is described. It is of advantage if shaping and mounting position coincide. (UWI) [de

  14. Conceptual design for Japan Sodium-Cooled Fast Reactor. (4) Developmental study of steel plate reinforced concrete containment vessel for JSFR

    International Nuclear Information System (INIS)

    Hosoya, Takusaburo; Negishi, Kazuo; Satoh, Kenichiro; Somaki, Takahiro; Matsuo, Ippei; Shimizu, Katsusuke

    2009-01-01

    An innovative containment vessel, namely Steel plate reinforced Concrete Containment Vessel (SCCV) is developed for Japan Sodium-Cooled Fast Reactor (JSFR). Reducing plant construction cost is one of the most important issues for commercialization of fast reactors. This study investigated construction issues including the building structure and the construction method as well as design issues in terms of the applicability of SCCV to fast reactors. An experimental study including loading and/or heating tests has been carried out to investigate the fundamental structural features, which would be provided to develop methodology to evaluate the feasibility of SCCV under the severe conditions. In this paper, the test plan is described as well as the first test results. (author)

  15. Radioactive liquid containing vessel

    International Nuclear Information System (INIS)

    Sakurada, Tetsuo; Kawamura, Hironobu.

    1993-01-01

    Cooling jackets are coiled around the outer circumference of a container vessel, and the outer circumference thereof is covered with a surrounding plate. A liquid of good conductivity (for example, water) is filled between the cooling jackets and the surrounding plate. A radioactive liquid is supplied to the container vessel passing through a supply pipe and discharged passing through a discharge pipe. Cooling water at high pressure is passed through the cooling water jackets in order to remove the heat generated from the radioactive liquid. Since cooling water at high pressure is thus passed through the coiled pipes, the wall thickness of the container vessel and the cooling water jackets can be reduced, thereby enabling to reduce the cost. Further, even if the radioactive liquid is leaked, there is no worry of contaminating cooling water, to prevent contamination. (I.N.)

  16. CONCRETE REACTOR CONTAINMENT

    Energy Technology Data Exchange (ETDEWEB)

    Lumb, Ralph F.; Hall, William F.; Fruchtbaum, Jacob

    1963-06-15

    The results of various leak-rate tests demonstrate the practicality of concrete as primary containment for the maximum credible accident for a research reactor employing plate-type fuel and having a power in excess of one megawatt. Leak-test time was shortened substantially by measuring the relaxation time for overpressure decay, which is a function of leak rate. (auth)

  17. Automatic design of prestressed concrete vessels

    International Nuclear Information System (INIS)

    Sotomura, Kentaro; Murazumi, Yasuyuki

    1984-01-01

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

  18. Development of ultrasonic testing technique with a large transducer to inspect the containment vessel plates embedded in concrete for corrosion on nuclear power plant (2)

    International Nuclear Information System (INIS)

    Ishida, Hitoshi

    2005-01-01

    The containment vessel plates embedded in concrete on Pressurized Water Reactors are inaccessible to inspect directly. Therefore, it is advisable to prepare inspection technology to detect existence and a location of corrosion on the embedded plates indirectly. The purpose of this study is establishment of ultrasonic testing technique to be able to inspect the containment vessel plates embedded in concrete widely from the accessible point. Experiments to detect artificial hollows simulating corrosion and stud bolts which hold the mold of concrete on a surface of a carbon steel plate mock-up covered with concrete were carried out with newly made low frequency (0.3MHz and 0.5MHz) 90 degrees refraction angle shear horizontal (SH) wave transducers combined with three active elements, which were equivalent to a 120 mm width element. As the results: (1) The echoes from the artificial hollows with a depth of 19 mm and 9.5mm at a distance of 1.5 m and the stud bolts with a diameter of 8mm at a distance of 0.7 - 1.7m could be discriminated clearly. (2) The multiple echoes bouncing three times between the front side and the back side of the plate, which was equivalent to a distance of about 12m, could be discriminated. (3) A divergence angle and a -6dB divergence angle of the large element (combined three elements) transducer were about 7 degrees and about 3 degrees. (4) The echoes from the hollows with a depth of 9.5m could be detected at a distance of 3.6 m with a reflection at the side wall of the mock-up. (5) It was estimated that the maximum distance of detection of the echo from the stud bolt with a diameter of 8mm was about 2.9 ∼ 3.6 m. Therefore we evaluate that the large element transducer can propagate the SH wave to about a half of a distance to the bottom of the embedded containment vessel and it is possible to detect the defects such as corrosion to a distance of 3.6 m. (author)

  19. Investigation of radial shear in the wall-base juncture of a 1:4 scale prestressed concrete containment vessel model

    Energy Technology Data Exchange (ETDEWEB)

    Dameron, R.A.; Rashid, Y.R. [ANATECH Corp., San Diego, CA (United States); Luk, V.K.; Hessheimer, M.F. [Sandia National Labs., Albuquerque, NM (United States)

    1998-04-01

    Construction of a prestressed concrete containment vessel (PCCV) model is underway as part of a cooperative containment research program at Sandia National Laboratories. The work is co-sponsored by the Nuclear Power Engineering Corporation (NUPEC) of Japan and US Nuclear Regulatory Commission (NRC). Preliminary analyses of the Sandia 1:4 Scale PCCV Model have determined axisymmetric global behavior and have estimated the potential for failure in several areas, including the wall-base juncture and near penetrations. Though the liner tearing failure mode has been emphasized, the assumption of a liner tearing failure mode is largely based on experience with reinforced concrete containments. For the PCCV, the potential for shear failure at or near the liner tearing pressure may be considerable and requires detailed investigation. This paper examines the behavior of the PCCV in the region most susceptible to a radial shear failure, the wall-basemat juncture region. Prediction of shear failure in concrete structures is a difficult goal, both experimentally and analytically. As a structure begins to deform under an applied system of forces that produce shear, other deformation modes such as bending and tension/compression begin to influence the response. Analytically, difficulties lie in characterizing the decrease in shear stiffness and shear stress and in predicting the associated transfer of stress to reinforcement as cracks become wider and more extensive. This paper examines existing methods for representing concrete shear response and existing criteria for predicting shear failure, and it discusses application of these methods and criteria to the study of the 1:4 scale PCCV.

  20. Investigation of radial shear in the wall-base juncture of a 1:4 scale prestressed concrete containment vessel model

    International Nuclear Information System (INIS)

    Dameron, R.A.; Rashid, Y.R.; Luk, V.K.; Hessheimer, M.F.

    1998-04-01

    Construction of a prestressed concrete containment vessel (PCCV) model is underway as part of a cooperative containment research program at Sandia National Laboratories. The work is co-sponsored by the Nuclear Power Engineering Corporation (NUPEC) of Japan and US Nuclear Regulatory Commission (NRC). Preliminary analyses of the Sandia 1:4 Scale PCCV Model have determined axisymmetric global behavior and have estimated the potential for failure in several areas, including the wall-base juncture and near penetrations. Though the liner tearing failure mode has been emphasized, the assumption of a liner tearing failure mode is largely based on experience with reinforced concrete containments. For the PCCV, the potential for shear failure at or near the liner tearing pressure may be considerable and requires detailed investigation. This paper examines the behavior of the PCCV in the region most susceptible to a radial shear failure, the wall-basemat juncture region. Prediction of shear failure in concrete structures is a difficult goal, both experimentally and analytically. As a structure begins to deform under an applied system of forces that produce shear, other deformation modes such as bending and tension/compression begin to influence the response. Analytically, difficulties lie in characterizing the decrease in shear stiffness and shear stress and in predicting the associated transfer of stress to reinforcement as cracks become wider and more extensive. This paper examines existing methods for representing concrete shear response and existing criteria for predicting shear failure, and it discusses application of these methods and criteria to the study of the 1:4 scale PCCV

  1. Model tests for prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Stoever, R.

    1975-01-01

    Investigations with models of reactor pressure vessels are used to check results of three dimensional calculation methods and to predict the behaviour of the prototype. Model tests with 1:50 elastic pressure vessel models and with a 1:5 prestressed concrete pressure vessel are described and experimental results are presented. (orig.) [de

  2. Prestressed concrete pressure vessels for nuclear reactors - 1973

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

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

  3. Containment vessel stability analysis

    International Nuclear Information System (INIS)

    Harstead, G.A.; Morris, N.F.; Unsal, A.I.

    1983-01-01

    The stability analysis for a steel containment shell is presented herein. The containment is a freestanding shell consisting of a vertical cylinder with a hemispherical dome. It is stiffened by large ring stiffeners and relatively small longitudinal stiffeners. The containment vessel is subjected to both static and dynamic loads which can cause buckling. These loads must be combined prior to their use in a stability analysis. The buckling loads were computed with the aid of the ASME Code case N-284 used in conjunction with general purpose computer codes and in-house programs. The equations contained in the Code case were used to compute the knockdown factors due to shell imperfections. After these knockdown factors were applied to the critical stress states determined by freezing the maximum dynamic stresses and combining them with other static stresses, a linear bifurcation analysis was carried out with the aid of the BOSOR4 program. Since the containment shell contained large penetrations, the Code case had to be supplemented by a local buckling analysis of the shell area surrounding the largest penetration. This analysis was carried out with the aid of the NASTRAN program. Although the factor of safety against buckling obtained in this analysis was satisfactory, it is claimed that the use of the Code case knockdown factors are unduly conservative when applied to the analysis of buckling around penetrations. (orig.)

  4. Mobile nuclear reactor containment vessel

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  5. Reactor containment vessel

    International Nuclear Information System (INIS)

    Ochiai, Kanehiro; Hayagumo, Sunao; Morikawa, Matsuo.

    1981-01-01

    Purpose: To safety and simplify the structure in a reactor containment vessel. Constitution: Steam flow channels with steam jetting ports communicating to coolants are provided between a communication channel and coolants in a pressure suppression chamber. Upon loss of coolant accidents, pressure in a dry well will increase, then force downwards water in an annulus portion and further flow out the water through steam jetting ports into a suppression pool. Thus, the steam flow channel is filled with steams or airs present in the dry well, which are released through the steam jetting ports into the pressure suppression chamber. Even though water is violently vibrated owing to the upward movement of air bubbles and condensation of steam bubbles, the annular portion and the steam jetting ports are filled with steams or the like, direct dynamic loads onto the structures such as communication channels can be avoided. (J.P.N.)

  6. Design and analysis of multicavity prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Goodpasture, D.W.; Burdette, E.G.; Callahan, J.P.

    1977-01-01

    During the past 25 years, a rather rapid evolution has taken place in the design and use of prestressed concrete reactor vessels (PCRVs). Initially the concrete vessel served as a one-to-one replacement for its steel counterpart. This was followed by the development of the integral design which led eventually to the more recent multicavity vessel concept. Although this evolution has seen problems in construction and operation, a state-of-the-art review which was recently conducted by the Oak Ridge National Laboratory indicated that the PCRV has proven to be a satisfactory and inherently safe type of vessel for containment of gas-cooled reactors from a purely functional standpoint. However, functionalism is not the only consideration in a demanding and highly competitive industry. A summary is presented of the important considerations in the design and analysis of multicavity PCRVs together with overall conclusions concerning the state of the art of these vessels

  7. Stress criteria for nuclear vessel concrete

    International Nuclear Information System (INIS)

    Costes, D.

    1975-01-01

    Concrete nuclear vessels are submitted to prestressing forces which limit tensile stresses in concrete when the vessel is under pressure with thermal gradients. Hence, the most severe conditions for concrete appear when the vessel is prestressed and not submitted to internal pressure. The triaxial states of stress in the concrete may be computed postulating elastic or other behavior and compared with safe limits obtained from rupture tests and fatigue tests. The first part of the paper, recalls experimental rupture results and the acceptability procedures currently used. Criteria founded on the lemniscoid surfaces are proposed, parameters for which are obtained by various tests and safety considerations. In the second part, rupture tests are reported on small, thick, cylindrical vessels submitted to external hydraulic pressure simulating prestressing forces. Materials used are plain concrete, microconcrete, marble and graphite. The strengths obtained are much higher than those which could be elastically computed, triaxial rupture states being provided by previous experiments. Such results may be due to a plastic stress redistribution before fracture and to stabilizing effects of stress gradients around the more stressed areas. Fatigue tests by external hydraulic loading are reported [fr

  8. Nuclear power plant prestressed concrete containment vessel structure monitoring during integrated leakage rate test using three kinds of fiber optic sensors

    Science.gov (United States)

    Liao, Kaixing; Li, Jinke; Kong, Xianglong; Sun, Changsen; Zhao, Xuefeng

    2017-04-01

    After years of operation, the safety of the prestressed concrete containment vessel (PCCV) structure of Nuclear Power Plant (NPP) is an important aspect. In order to detect the strength degradation and the structure deformation, several sensors such as vibrating wire strain gauge, invar wires and pendulums were installed in PCCV. However, the amounts of sensors above are limited due to the cost. Due to the well durability of fiber optic sensors, three kinds of fiber optic sensors were chosen to install on the surface of PCCV to monitor the deformation during Integrated Leakage Rate Test (ILRT). The three kinds of fiber optic sensors which had their own advantages and disadvantages are Fiber Bragg Grating (FBG), white light interferometry (WLI) and Brillouin Optical Time Domain Analysis (BOTDA). According to the measuring data, the three fiber optic sensors worked well during the ILRT. After the ILRT, the monitoring strain was recoverable thus the PCCV was still in the elastic stage. If these three kinds of fiber optic sensors are widely used in the PCCV, the unusual deformations are easier to detect. As a consequence, the three fiber optic sensors have good potential in the structure health monitoring of PCCV.

  9. Prestressed and reinforced concrete containments. Analysis - design - construction

    International Nuclear Information System (INIS)

    Schnellenbach, G.

    1975-01-01

    Nuclear reactors performing in the German Federal Republic to date were supplied with steel containments. The first reinforced concrete and prestressed concrete containments, respectively, are going to be used for the nuclear power plants Kalkar and Gundremmingen (KRB II) as well as for the HTR plant. Because of their function and nature of loading these structures, similarly to the prestressed concrete reactor pressure vessels, belong to the special structures of civil engineering. Yet, they are substantially different from the prestressed concrete reactor pressure vessels. The problems connected with analysis, design, and construction of these structures are new as well. (orig.) [de

  10. Depleted uranium concrete container feasibility study

    International Nuclear Information System (INIS)

    Haelsig, R.T.

    1994-09-01

    The purpose of this report is to consider the feasibility of using containers constructed of depleted uranium aggregate concrete (DUCRETE) to store and transport radioactive materials. The method for this study was to review the advantages and disadvantages of DUCRETE containers considering design requirements for potential applications. The author found that DUCRETE is a promising material for onsite storage containers, provided DUCRETE vessels can be certified for one-way transport to disposal sites. The author also found that DUCRETE multipurpose spent nuclear fuel storage/transport packages are technically viable, provided altered temperature acceptance limits can be developed for DUCRETE

  11. The evolution and structural design of prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Hannah, I.W.

    1978-01-01

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

  12. Results of steel containment vessel model test

    International Nuclear Information System (INIS)

    Luk, V.K.; Ludwigsen, J.S.; Hessheimer, M.F.; Komine, Kuniaki; Matsumoto, Tomoyuki; Costello, J.F.

    1998-05-01

    A series of static overpressurization tests of scale models of nuclear containment structures is being conducted by Sandia National Laboratories for the Nuclear Power Engineering Corporation of Japan and the US Nuclear Regulatory Commission. Two tests are being conducted: (1) a test of a model of a steel containment vessel (SCV) and (2) a test of a model of a prestressed concrete containment vessel (PCCV). This paper summarizes the conduct of the high pressure pneumatic test of the SCV model and the results of that test. Results of this test are summarized and are compared with pretest predictions performed by the sponsoring organizations and others who participated in a blind pretest prediction effort. Questions raised by this comparison are identified and plans for posttest analysis are discussed

  13. Finite element analysis of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Smith, P.D.; Cook, W.A.; Anderson, C.A.

    1977-01-01

    Several present and proposed gas-cooled reactors use concrete pressure vessels. In addition, concrete is almost universally used for the secondary containment structures of water-cooled reactors. Regulatory agencies must have means of assuring that these concrete structures perform their containment functions during normal operation and after extreme conditions of transient overpressure and high temperature. The NONSAP nonlinear structural analysis program has been extensively modified to provide one analytical means of assessing the safety of reinforced concrete pressure vessels and containments. Several structural analysis codes were studied to evaluate their ability to model the nonlinear static and dynamic behavior of three-dimensional structures. The NONSAP code was selected because of its availability and because of the ease with which it can be modified. In particular, the modular structure of this code allows ready addition of specialized material models. Major modifications have been the development of pre- and post-processors for mesh generation and graphics, the addition of an out-of-core solver, and the addition of constitutive models for reinforced concrete subject to either long-term or short-term loads. Emphasis was placed on development of a three-dimensional analysis capability

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

    International Nuclear Information System (INIS)

    1987-11-01

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

  15. Stowing the Right Containers on Container Vessels

    DEFF Research Database (Denmark)

    Jensen, Rune Møller

    2014-01-01

    ’s largest container vessels using standard mathematical programming techniques and off-the-shelf solvers. The presentation will provide basic insight into the domain, with pointers to further information that enable you to join in this promising new path of operations research and business....

  16. Molten material-containing vessel

    International Nuclear Information System (INIS)

    Akagawa, Katsuhiko

    1998-01-01

    The molten material-containing vessel of the present invention comprises a vessel main body having an entrance opened at the upper end, a lid for closing the entrance, an outer tube having an upper end disposed at the lower surface of the lid, extended downwardly and having an closed lower end and an inner tube disposed coaxially with the outer tube. When a molten material is charged from the entrance to the inside of the vessel main body of the molten material-containing vessel and the entrance is closed by the lid, the outer tube and the inner tube are buried in the molten material in the vessel main body, accordingly, a fluid having its temperature elevated by absorption of the heat of the molten material rises along the inner circumferential surface of the outer tube, abuts against the lower surface of the lid and cooled by exchanging heat with the lid and forms a circulating flow. Since the heat in the molten material is continuously absorbed by the fluid, transferred to the lid and released from the lid to the atmospheric air, heat releasing efficiency can be improved compared with conventional cases. (N.H.)

  17. Material problems in accident analysis of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Bazant, Z.P.

    1977-01-01

    Due to their very high energy absorption capability, as well as their inherent safety advantages, prestressed concrete reactor vessels are presently being keenly studied as the basic barrier to contain hypothetical core disruptive accidents in a fast breeder reactor. One problem investigated is the nonlinear constitutive behavior and failure criteria for concrete. Previously, a comprehensive theory, called endochronic theory, has been shown to satisfy all basic currently known features of test data. Nevertheless uncertainty still exists with regard to non-proportional loading paths, for which good test data are lacking at present. An extension of the endochronic theory which correlates best with general experimental evidence and includes fracturing terms is given, and a comparison with vertex-type hardening in plasticity is made. A second problem which must be analysed in accident situations is the high temperature shock on the concrete walls (due to liquid sodium, up to 850 0 C). Refining a previous crude formulation, a rational model for calculating moisture and heat transfer and pore pressures in concrete subjected to thermal shock is presented. In conclusion, a new design concept, in which the concrete vessel is completely dehydrated and kept hot throughout its service life in order to substantially improve its response to thermal shock as well as liquid sodium contact, is described. (Auth.)

  18. Rapid construction of concrete pressure vessels

    International Nuclear Information System (INIS)

    Limbert, D.; Weatherseed, D.C.

    1989-01-01

    This paper opens with a general description of the concrete pressure vessel followed by a more detailed examination of the critical elements of the construction, including choice of methods and plant which were selected to ensure its rapid construction. The pressure vessel construction cannot be treated in isolation, because it is very closely linked with its surrounding structures - namely the reactor hall which surrounds it and the charge hall which tops it, as will be seen in the context of this paper. Rate of progress of construction is not entirely in the civil contractor's hands because so many of the operations affecting the civil works are of a mechanical nature, hence a very close liaison and understanding amongst all contractors concerned was of the utmost importance. (author)

  19. Concrete and prestressing process, container made with this concrete

    International Nuclear Information System (INIS)

    Gerard, M.

    1992-01-01

    Shape memory alloy fibers or heat shrinking fibers are encapsulated in a standard concrete. Prestressed concrete is obtained by heat treatment. Application is made to the fabrication of radioactive waste containers

  20. Construction of reactor vessel bottom of prestressed reinforced concrete

    International Nuclear Information System (INIS)

    Sitnikov, M.I.; Metel'skij, V.P.

    1980-01-01

    Methods are described for building reactor vessel bottoms of prestressed reinforced concrete during NPPs construction in Great Britain, France, Germany (F.R.) and the USA. Schematic of operations performed in succession is presented. Considered are different versions of one of the methods for concreting a space under a facing by forcing concrete through a hole in the facing. The method provides tight sticking of the facing to the reactor vessel bottom concrete

  1. Method of detecting construction faults in concrete pressure vessels

    International Nuclear Information System (INIS)

    Robertson, S.A.; Duhoux, M.; Dawance, G.; Carrie, C.; Morel, D.

    1976-01-01

    A major problem in the design and construction of concrete pressure vessels for nuclear power stations is the risk of excessive air leaks through the concrete itself, due to faulty construction. The 'sonic coring' method of non-destructive concrete testing has been used successfully in pile and diaphragm wall construction control for several years, and the potential use of this method to control the presence of faults in concrete pressure vessels is here described. (author)

  2. Nonlinear analysis of end slabs in prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Abdulrahman, H.O.

    1978-01-01

    A procedure for the nonlinear analysis of end slabs is prestressed concrete reactor vessels (PCRVs), based on the finite element method, is presented. The applicability of the procedure to the ultimate load analysis of small-scale models of the primary containment of nuclear reactors is shown. Material nonlinearity only is considered. The procedure utilizes the four-node linear quadrilateral isoparametric element with the choice of incorporating the nonconforming modes. This element is used for modeling the vessel as an axisymmetric solid. Concrete is assumed to be an isotropic material in the elastic range. The compressive stresses are judged according to a special form of the Mohr-Coulomb criterion. The nonlinear problem was solved using a generalized Newton-Raphson procedure. A detailed example problem of a pressure vessel with penetrations is presented. This is followed by a summary of the other cases studied. The solutions obtained match very closely the measured response of the test vessels under increasing internal pressure up to failure. The procedure is thus adequate for the assessment of the ultimate load behavior and failure of actual pressure vessels with a moderate demand on human and computational resources

  3. HTGR Base Technology Program. Task 2: concrete properties in nuclear environment. A review of concrete material systems for application to prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Naus, D.J.

    1981-05-01

    Prestressed concrete pressure vessels (PCPVs) are designed to serve as primary pressure containment structures. The safety of these structures depends on a correct assessment of the loadings and proper design of the vessels to accept these loadings. Proper vessel design requires a knowledge of the component (material) properties. Because concrete is one of the primary constituents of PCPVs, knowledge of its behavior is required to produce optimum PCPV designs. Concrete material systems are reviewed with respect to constituents, mix design, placing, curing, and strength evaluations, and typical concrete property data are presented. Effects of extreme loadings (elevated temperature, multiaxial, irradiation) on concrete behavior are described. Finally, specialty concrete material systems (high strength, fibrous, polymer, lightweight, refractory) are reviewed. 235 references

  4. Proceedings of the CSNI workshop on International Standard Problem 48 - Analysis of 1:4-scale prestressed concrete containment vessel model under severe accident conditions

    International Nuclear Information System (INIS)

    2005-01-01

    At the CSNI meeting in June 2002, the proposal for an International Standard Problem on containment integrity (ISP 48) based on the NRC/NUPEC/Sandia test was approved. Objectives were to extend the understanding of capacities of actual containment structures based on results of the recent PCCV Model test and other previous research. The ISP was sponsored by the USNRC, and results had been made available thanks to NUPEC and to the USNRC. Sandia National Laboratory was contracted to manage the technical aspects of the ISP. At the end of the ISP48, a workshop was organized in Lyon, France on April 6-7, 2005 hosted by Electricite de France. Its overall objective was to present results obtained by participants in the ISP 48 and to assess the current practices and the state of the art with respect to the calculation of concrete structures under severe accident conditions. Experience from other areas in civil engineering related to the modelling of complex structures was greatly beneficial to all. Information obtained as a result of this assessment were utilized to develop a consensus on these calculations and identify issues or 'gaps' in the present knowledge for the primary purpose of formulating and prioritizing research needs on this topic. The ISP48 exercise was published in the report referenced NEA/CSNI/R(2005)5 in 3 volumes. Volume 1 contains the synthesis of the exercise; Volumes 2 and 3 contain individual contributions of participating organizations. The CSNI Working Group on the Integrity and Ageing and in particular its sub-group on the behaviour of concrete structures has produced extensive material over the last few years. The complete list of references is given in this document. These proceedings gather the papers and presentations given by the participants at the Lyon workshop

  5. Temperature field in the bottom of concrete reactor vessel; Temperaturno polje u podu betonskog reaktorskog suda

    Energy Technology Data Exchange (ETDEWEB)

    Jovasevic, V; Tosic, D; Zaric, S; Maksimovic, Lj [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1969-07-01

    This paper contains detailed scheme of reactor bottom vessel made of concrete and the results of calculated relevant temperature distribution. Method applied for calculation is described taking into account all relevant factors and assuming that thermal conductivity of concrete is homogeneous and independent of temperature.

  6. Limit analysis and design of containment vessels

    International Nuclear Information System (INIS)

    Save, M.

    1984-01-01

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

  7. Application of an automated patrol-type monitoring system in the reinforced concrete containment vessel at Kashiwazaki-Kariwa Nuclear Power Plant Units 6 and 7

    International Nuclear Information System (INIS)

    Kimoto, T.; Senoh, M.; Hirakawa, H.; Tanaka, K.; Takahashi, M.

    1996-01-01

    An automated patrol-type monitoring system was applied to the first Advanced Boiling Water Reactor (ABWR) plants in the world, Kashiwazaki-Kariwa Nuclear Power Plant Units 6 and 7. The system consists of two robots traveling along monorail and several ITV systems. This monitoring system is used to watch the operating incident of key components of ABWR, such as water leakage from packing in Fine Motion Control Rod Drive (FM-CRD), Reactor Internal Pump (RIP) etc. which are located in the reinforced concrete containment and usually could not be accessible under operation. The traveling robots are equipped with the three kinds of sensors, an infrared camera, a color TV camera and a microphone, and have a capability to inspect the pre-determined inspection points guided by the monorail. This monitoring system is automatically turned on to start inspection at a scheduled time. For Units 6 and 7, this monitoring system, two robots and several ITV systems, can be controlled from one common control desk and one handy-type remote control panel. Since this monitoring system was installed at Kashiwazaki-Kariwa Nuclear Power Plant Units 6 and 7, this system has been operating successfully. (The installation and testing of the system for Unit 6 was finished completely and the testing the system for Unit 7 has been carrying out.) (author)

  8. Radiation resistance of concrete of nuclear reactor vessel

    International Nuclear Information System (INIS)

    Belyakov, V.V.; Denisov, A.V.; Korenevskij, V.V.; Muzalevskij, L.P.; Dubrovskij, V.B.; Ivanov, D.A.; Nazarov, I.L.; Sashin, N.L.

    1992-01-01

    Results of calculational-experimental determination of radiation resistance for concrete bases on limestone gravel and quartz sand, which are the most perspective materials for manufacturing prestressed concrete of the VG-400 reactor vessel are considered. Material samples under investigation were irradiated in the channels of the IBR-2 research reactor for the purpose of the calcultional result verification

  9. Review of analysis methods for prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Dodge, W.G.; Bazant, Z.P.; Gallagher, R.H.

    1977-02-01

    Theoretical and practical aspects of analytical models and numerical procedures for detailed analysis of prestressed concrete reactor vessels are reviewed. Constitutive models and numerical algorithms for time-dependent and nonlinear response of concrete and various methods for modeling crack propagation are discussed. Published comparisons between experimental and theoretical results are used to assess the accuracy of these analytical methods

  10. High temperature helium test rig with prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Schmidl, H.

    1975-10-01

    The report gives a short description of the joint project prestressed concrete vessel-helium test station as there is the building up of the concrete structure, the system of instrumentation, the data processing, the development of the helium components as well as the testing programs. (author)

  11. Reactors with pressure vessel in pre-stressed concrete

    International Nuclear Information System (INIS)

    Devillers, Christian; Lafore, Pierre

    1964-12-01

    After having proposed a general description of the evolution of the general design of reactors with a vessel in pre-stressed concrete, this report outlines the interest of this technical solution of a vessel in pre-stressed concrete with integrated exchangers, which is to replace steel vessel. This solution is presented as much safer. The authors discuss the various issues related to protection: inner and outer biological protection of the vessel, material protection (against heating, steel irradiation, Wigner effect, and moderator radiolytic corrosion). They report the application of calculation methods: calculation of vessel concrete heating, study of the intermediate zone in integrated reactors, neutron spectrum and flows in the core of a graphite pile

  12. Corrosion of steel tendons used in prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Griess, J.C.; Naus, D.J.

    The purpose of this investigation was to determine the corrosion behavior of a high strength steel (ASTM A416-74 grade 270), typical of those used as tensioning tendons in prestressed concrete pressure vessels, in several corrosive environments and to demonstrate the protection afforded by coating the steel with either of two commercial petroleum-base greases or Portland Cement grout. In addition, the few reported incidents of prestressing steel failures in concrete pressure vessels used for containment of nuclear reactors are reviewed. The susceptibility of the steel to stress corrosion cracking and hydrogen embrittlement and its general corrosion rate were determined in several salt solutions. Wires coated with the greases and grout were soaked for long periods in the same solutions and changes in their mechanical properties were subsequently determined. All three coatings appeared to give essentially complete protection but small flaws in the grease coatings were detrimental; flaws or cracks less than 1 mm wide in the grout were without effect

  13. Cylindrical prestressed concrete pressure vessel for a nuclear power plant

    International Nuclear Information System (INIS)

    Horner, M.; Hodzic, A.; Haferkamp, D.

    1976-01-01

    A prestressed concrete pressure vessel for a HTGR is proposed which encloses, in addition to the reactor core, not only the heat-exchanging facilities but also the turbine unit. The reinforcement of the cylindrical concrete body is to be carried out with special care, it is provided for horizontal tendons, the prestressed concrete pressure vessel has a wire-winding device, while the longitudinal reinforcement is achieved by tendous guided in parallel to the vesses axes through the interspaces between the pods. (UWI) [de

  14. In-service supervision of a prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Zemann, H.; Mayer, N.; Amberg, C.

    1985-01-01

    On-line measurements of the physical state of a prestressed concrete pressure vessel and a comparison of the distribution of temperature, strain and stress within the concrete member to the optimized statical predictions and the criterions of layout yield to an efficient and economical method of operating the vessel with a high potential of safety. The requirements of instrumentation and the comparison with static calculations are discussed on the prototype vessel at Seibersdorf Research Center during the phase of construction and prestressing, the phase of the first thermal treatment (stabilization), the pressure tests and under the operating conditions of a high temperature reactor (150 0 C/50 bar). (Author)

  15. In-service supervision of a prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Zemann, H.; Weissbacher, L.; Mayer, N.; Amberge, C.

    1985-01-01

    On-line measurements of the physical state of a prestressed concrete pressure vessel, and comparison with the design predictions of the distribution of temperature, strain and stress within the concrete member and the criteria of layout, provide an efficient and economical method of operating the vessel with a high potential of safety. The requirements of instrumentation and the comparison with static calculations are discussed with reference to the prototype vessel at Seibersdorf Research Centre during the phase of construction and prestressing, the phase of the first thermal treatment (stabilization), the pressure tests and under the operating conditions of a high temperature reactor (150 0 C, 50 bar). (author)

  16. Behaviour of concrete containment under over-pressure conditions

    International Nuclear Information System (INIS)

    Atchison, R.J.; Asmis, G.J.K.; Campbell, F.R.

    1979-01-01

    The Atomic Energy Control Board of Canada initiated June, 1975, a major study of the behaviour of concrete containment under over-pressure conditions. Although extensive theoretical and experimental work has been carried out for thick-walled Prestressed Concrete Reactor Vessels (PCRV's), there is a want of information on the non-linear response of thin-walled structures typical of the CANDU, 600 MW(e) cylindrical/spherical, post-tensioned containment shells. The purpose of this paper is to provide an overview of the total program, to present the reasons behind the research contract, and the specification and implementation of the work. The results of the theoretical and experimental work and their implications with respect to Canadian Concrete Containment practice are discussed. This study is unique, and, as far as is known, has no world-wide precedence. (orig.)

  17. Ageing management of CANDUtm concrete containment buildings

    International Nuclear Information System (INIS)

    Philipose, K.E.; Gregor, F.E.

    2009-01-01

    The containment system in a Nuclear Power Plant (NPP) provides the final physical barrier against release of radioactive materials to the external environment. Even though there are different physical configurations to meet this fundamental safety function in various reactor types, a common feature is the use of a thick-walled concrete structure as part of the containment system commonly referred to as 'Concrete Containment Building'. In order for the concrete containment buildings to continue to provide the required safety function, it has to maintain its structural integrity. As well, its leak rates under test pressures must be maintained below acceptable limits. As some of the containment buildings of the CANDU nuclear power plants are approaching their fourth decade of successful operation, questions regarding the impact of ageing on their ultimate useful service life emerge. Ageing Management has become the tool for addressing those questions. In this paper, the ageing and ageing management of the CANDU concrete containments are discussed, including the specific programs being implemented to monitor and trend the ageing conditions. Specifically, the usefulness of the embedded strain gauges as a tool for the assessment of the condition of the containment concrete structure is discussed. Some of the operational and test data accumulated over the last 30 years have been evaluated and trended to provide some results and conclusions regarding the satisfactory long-term behaviour of the concrete containment buildings. (authors)

  18. Shear strength of end slabs of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Cheung, K.C.; Gotschall, H.L.; Liu, T.C.

    1975-01-01

    Prestressed concrete reactor vessels (PCRV's) have been adopted for primary containments in most large high-temperature gas-cooled reactor installations. The most common configuration for PCRVs is a right-vertical cylinder with thick end slabs. In order to assess the integrity of a PCRV it is necessary to predict the ultimate strength of the end slabs. The complexity of the basic mechanism of shear failure in the PCRV end slabs has thus far prohibited the development of a completely analytical solution. However, many experimental investigations of PCRV end slabs have been conducted over the past decade. This information makes it possible to establish empirical formulae for the ultimate strength of PCRV end slabs. The basis and development of an empirical shear-flexure interaction expression is presented. (Auth.)

  19. Testing of plain and fibrous concrete single cavity prestressed concrete reactor vessel models

    International Nuclear Information System (INIS)

    Oland, C.B.

    1985-01-01

    Two single-cavity prestressed concrete reactor vessel (PCRV) models were fabricated and tested to failure to demonstrate the structural response and ultimate pressure capacity of models cast from high-strength concretes. Concretes with design compressive strengths in excess of 70 MPa (10,000 psi) were developed for this investigation. One model was cast from plain concrete and failed in shear at the head region. The second model was cast from fiber reinforced concrete and failed by rupturing the circumferential prestressing at the sidewall of the structure. The tests also demonstrated the capabilities of the liner system to maintain a leak-tight pressure boundary. 3 refs., 4 figs

  20. A thermal insulation system intended for a prestressed concrete vessel

    International Nuclear Information System (INIS)

    Aubert, Gilles; Petit, Guy.

    1975-01-01

    The description is given of a thermal insulation system withstanding the pressure of a vaporisable fluid for a prestressed concrete vessel, particularly the vessel of a boiling water nuclear reactor. The ring in the lower part of the vessel has, between the fluid inlet pipes and the bottom of the vessel, an annular opening of which the bottom edge is integral with an annular part rising inside the ring and parallel to it. This ring is hermetically connected to the bottom of the vessel and is coated with a metal lagging, at least facing the annular opening. This annular opening is made in the ring half-way up between the fluid inlet pipes and the bottom of the vessel. It is connected to the bottom of the vessel through the internal structure enveloping the reactor core [fr

  1. Completely integrated prestressed-concrete reactor pressure vessel, type 'Star'

    International Nuclear Information System (INIS)

    Neunert, B.; Jueptner, G.; Kumpf, H.

    1975-01-01

    The star support vessel is suitable for the connection to all primary circuit systems consisting of a main vessel and a number of satellite vessels around and connected to it, i.e. for LWR, HTR and process reactor. It must be made clear, however, that the PWR in particular with its components does not appear to be suited for the optimum incorporation in a prestressed-concrete pressure vessel system, no matter what kind. There are clear concepts about modifications which, however, require considerable development expenditure. (orig./LH) [de

  2. Prestressed concrete reactor vessels: review of design and failure criteria

    International Nuclear Information System (INIS)

    Endebrock, E.G.

    1975-03-01

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

  3. Preliminary investigation on the suitablity of using fiber reinforced concrete in the construction of a hazardous waste disposal vessel

    International Nuclear Information System (INIS)

    Ramey, M.R.; Daie-e, G.

    1988-07-01

    There are certain hazardous wastes that must be contained in an extremely secure vessel for transportation and disposal. The vessel, among other things, must be able to withstand relatively large impacts without rupturing. Such containment vessels therefore must be able to absorb substantial amounts of energy during an impact and still perform their function. One of the impacts that the vessel must withstand is a 30-foot fall onto an unyielding surface. For some disposal scenarios it is proposed to encase the waste in a steel enclosure which is to be surrounded by a thick layer of concrete which, in turn, is encased by a relatively thin steel shell. Tests on concrete in compression and flexure, including static, dynamic and impact tests, have shown that low modulus concretes tend to behave in a less brittle manner than higher modulus concretes. Tests also show that fiber reinforced concretes have significantly greater ductility, crack propagation resistance and toughness than conventional concretes. Since it is known that concrete is a reasonably brittle material, it is necessary to do impact tests on sample containment structures consisting of thin-walled metal containers having closed ends which are filled with concrete, grout, or fiber reinforced concrete. This report presents the results of simple tests aimed at observing the behavior of sample containment structures subjected to impacts due to a fall from 30 feet. 8 figs., 4 tabs

  4. Heat and mass transfer in a concrete pressure vessel

    International Nuclear Information System (INIS)

    Zangle, K.; Sadouki, H.; Wittmann, F.H.

    1989-01-01

    Pressure vessels of prestressed concrete for high temperature reactors are subjected to high mechanical and thermal stresses during the reactors normal working conditions and in particular accidental conditions. According to a large temperature gradient between the inner liner and the outer side of the thickwalled vessel, physical as well as chemical processes take place in concrete. Temperature and moisture content of concrete have a big influence on these processes. During the last years different investigations have been conducted in order to determine characteristic values of concrete under these conditions. At present the authors conduct a series of experiments on model vessels of prestressed concrete and a large number of small specimens. The aims of these tests can be briefly summarized as follows: experimental determination of transport coefficients for a numerical analysis; determination of chemical reactions under hydrothermal conditions and their significance for the risk of corrosion; determination of temperature and moisture distribution as a function of time; and determination of the strength development in the zones subjected to elevated temperatures

  5. Prestressed concrete pressure vessels for boiling water reactors

    International Nuclear Information System (INIS)

    Menon, S.

    1979-12-01

    Following a general description of the Scandinavian cooperative project on prestressed concrete pressure vessels for boiling water reactors, detailed discussion is given in four appendices of the following aspects: the verification programme of tests and studies, the development and testing of a liner venting system, a preliminary safety philosophy and comparative assessment of cold and hot liners. Vessel failure probability is briefly discussed and some figures presented. The pressure gradients in the vessel wall resulting from various stipulated linear cracks, with a liner venting system are presented graphically. (JIW)

  6. Tension tests of concrete containment wall elements

    International Nuclear Information System (INIS)

    Schultz, D.M.; Julien, J.T.; Russel, H.G.

    1984-01-01

    Tension tests of concrete containment wall elements were conducted as part of a three-phase research program sponsored by the Electric Power Research Institute (EPRI). The objective of the EPRI experimental/analytical program is twofold. The first objective is to provide the utility industry with a test-verified analytical method for making realistic estimates of actual capacities of reinforced and prestressed concrete containments under internal over-pressurization from postulated degraded core accidents. The second objective is to determine qualitative and quantitative leak rate characteristics of typical containment cross-sections with and without penetrations. This paper covers the experimental portion to the EPRI program. The testing program for Phase 1 included eight large-scale specimens representing elements from the wall of a containment. Each specimen was 60-in (1525-mm) square, 24-in (610-mm) thick, and had full-size reinforcing bars. Six specimens were representative of prototypical reinforced concrete containment designs. The remaining two specimens represented prototypical prestressed containment designs. Various reinforcement configurations and loading arrangements resulted in data that permit comparisons of the effects of controlled variables on cracking and subsequent concrete/reinforcement/liner interaction in containment elements. Subtle differences, due to variations in reinforcement patterns and load applications among the eight specimens, are being used to benchmark the codes being developed in the analytical portion of the EPRI program. Phases 2 and 3 of the test program will examine leak rate characteristics and failure mechanisms at penetrations and structural discontinuities. (orig.)

  7. Requirements for thermal insulation on prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  8. Design criteria for prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Schimmelpfennig, K.

    1989-01-01

    The work concerned with the PCRVs has been focussed on topics which are not sufficiently covered by the usual codes with respect to the special structure of PCRVs and the special demands on it, and different investigations yielding a basis for such specific design criteria have been carried out. Only a couple of subjects being in the fore under the aspect of defining quality enlarging design criteria for PCRVs are outlined. The materials for the concrete to be used for the PCRVs are carefully selected. (DG)

  9. Prestressed concrete vessels suitable for helium high temperature reactors

    International Nuclear Information System (INIS)

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

    1967-02-01

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

  10. Method of burying vessel containing radioactive waste

    International Nuclear Information System (INIS)

    Koga, Yoshihito.

    1989-01-01

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

  11. Reliability analysis of prestressed concrete containment structures

    International Nuclear Information System (INIS)

    Jiang, J.; Zhao, Y.; Sun, J.

    1993-01-01

    The reliability analysis of prestressed concrete containment structures subjected to combinations of static and dynamic loads with consideration of uncertainties of structural and load parameters is presented. Limit state probabilities for given parameters are calculated using the procedure developed at BNL, while that with consideration of parameter uncertainties are calculated by a fast integration for time variant structural reliability. The limit state surface of the prestressed concrete containment is constructed directly incorporating the prestress. The sensitivities of the Choleskey decomposition matrix and the natural vibration character are calculated by simplified procedures. (author)

  12. The need to pressure test prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Forgie, J.H.; Holland, J.A.

    1983-01-01

    In the period when PCRV were relatively unproven, proof pressure testing provided a useful demonstration of vessel integritiy and a confirmation of model testing and of analysis. No failures have occurred during concrete vessel tests in the UK or in the subsequent operational life of the vessels and much has been learned of their behaviour in service. The paper examines the advantages and disadvantages of proof testing PCRV in the light of the above increased knowledge of vessel performance. The paper draws attention to certain hypothetical loading cases that could be more onerous than the proof test and suggests that pressure testing could itself cause unnecessarily high loading to parts of the vessel. Always recognising the safety considerations and demonstrations of such are of prime importance, the authors suggest that a lower pressure level could be adopted without loss of original intent. In addition some ground rules are suggested as to cases where proof testing could be omitted. (orig./HP)

  13. Transient analysis of LMFBR reinforced/prestressed concrete containment

    International Nuclear Information System (INIS)

    Marchertas, A.H.; Belytschko, T.B.; Bazant, Z.P.

    1979-01-01

    The use of prestressed concrete reactor vessels (PCRVs) for LMFBR containment creates a need for analytical methods for treating the transient response of such structures, for LMFBR containments must be capable of sustaining the dynamic effects which arise in a hypothetical core disruptive accident (HCDA). These analyses require several unique features: a model of concrete which includes tensile cracking, a methodology for representing the prestressing tendons and for simulating the prestressing operation, and an efficient computational tool for treating the transient response. Furthermore, for the sake of convenience, all of these features should be available in a single computer code. For the purpose of treating the transient response, a finite element program with explicit time integration was chosen. The use of explicit time integration has the advantage that it can easily treat the complicated constitutive model which arises from the considerations of concrete cracking and it can handle the slip between reinforcing tendons and the concrete through the use of the well known sliding interface options. However, explicit time integration programs are usually not well suited to the simulation of static processes such as prestressing. Nevertheless, explicit time integration programs can handle static processes through the introduction of damping by what is known as a dynamic relaxation procedure. For this reason, the dynamic relaxation procedure was refined through the introduction of lumped mass, viscous damping. This provision made the prestressing operation of the concrete structures by means of the explicit formulation rather convenient. (orig.)

  14. Nonlinear analysis of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

    Berg, S.; Loeseth, S.; Holand, I.

    1977-01-01

    A computational model for circular symmetric reinforced concrete shell problems is described. The model is based on the Finite Element Method. Non-linear stress-strain constitutive relations are used for the concrete, the reinforcement and for the liner. The reinforcement layers may be of different steel qualities. Each layer may be given a specified prestressing. This can be done at the beginning of the computations or the specific reinforcement layer can be considered inactive until a specified level of loading is reached. Thus, the prestressing procedure may also be analyzed in detail. Bond-slip effects are not accounted for. However, no bond may be assumed for prestressing cables by inserting special reinforcement elements. Several models of prestressed concrete reactor pressure vessels which have been tested up to rupture have been analysed. Analytical (numerical) models for reinforced concrete are also discussed on a more general basis. (Auth.)

  15. Containment vessel construction for nuclear power reactors

    International Nuclear Information System (INIS)

    Sulzer, H.D.; Coletti, J.L.

    1975-01-01

    A nuclear containment vessel houses an inner reactor housing structure whose outer wall is closely spaced from the inner wall of the containment vessel. The inner reactor housing structure is divided by an intermediate floor providing an upper chamber for housing the reactor and associated steam generators and a lower chamber directly therebeneath containing a pressure suppression pool. Communication between the upper chamber and the pressure suppression pool is established by conduits extending through the intermediate floor which terminate beneath the level of the pressure suppression pool and by inlet openings in the reactor housing wall beneath the level of the pressure suppression pool which communicate with the annulus formed between the outer wall of the reactor housing structure and the inner wall of the containment vessel. (Official Gazette)

  16. Testing of a steel containment vessel model

    International Nuclear Information System (INIS)

    Luk, V.K.; Hessheimer, M.F.; Matsumoto, T.; Komine, K.; Costello, J.F.

    1997-01-01

    A mixed-scale containment vessel model, with 1:10 in containment geometry and 1:4 in shell thickness, was fabricated to represent an improved, boiling water reactor (BWR) Mark II containment vessel. A contact structure, installed over the model and separated at a nominally uniform distance from it, provided a simplified representation of a reactor shield building in the actual plant. This paper describes the pretest preparations and the conduct of the high pressure test of the model performed on December 11-12, 1996. 4 refs., 2 figs

  17. Innovations in prestressed concrete pressure vessel design

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  18. A prestressed concrete pressure vessel for helium high temperature reactor system

    International Nuclear Information System (INIS)

    Horner, R.M.W.; Hodzic, A.

    1976-01-01

    A novel prestressed concrete pressure vessel has been developed to provide the primary containment for a fully integrated system comprising a high temperature nuclear reactor, three horizontally mounted helium turbines, associated heat exchangers and inter-connecting ducts. The design and analysis of the pressure vessel is described. Factors affecting the final choice of layout are discussed, and earlier development work seeking to resolve the conflicting requirements of the structural, mechanical, and system engineers outlined. Proposals to increase the present output of about 1000 MW of electrical power to over 3000 MW, by incorporating four turbines in a single pressure vessel are presented. (author)

  19. Thermal effects in concrete containment analysis

    International Nuclear Information System (INIS)

    Pfeiffer, P.A.; Kennedy, J.M.; Marchertas, A.H.

    1988-01-01

    Analyses of the thermo-mechanical response of the 1:6-scale reinforced concrete containment are presented. Three temperature- pressure scenarios are analyzed to complete loss of the pressure integrity. These results are compared to the analysis of pressure alone, to assess the importance of thermal effects. 19 refs., 9 figs., 8 tabs

  20. Storage vessel for radiation contaminated container

    International Nuclear Information System (INIS)

    Sakatani, Tadatsugu.

    1996-01-01

    In a storage vessel of the present invention, a plurality of radiation contaminated material containing bodies are vertically stacked in a cell chamber. Then, the storage vessel comprises a containing tube for containing a plurality of the containing bodies, cooling coils wound around the containing tube, a cooling medium circulating system connected to the cooling coils and circulating cooling medium, and a heat exchanger interposed to the cooling medium circulating system for removing heat of the cooling medium. Heat of the radioactive material containing bodies is transferred to cooling air and cooling coils by way of the container tube, thereby cooling the containing bodies. By the operation of circulating pumps in a cooling medium circulation system, the cooling medium circulates through a circulation channel comprising a cooling medium transfer pipes, cooling medium branching tubes, cooling coils and the heat exchanger, then heat of the cooling medium is transferred to a heat utilizing system by way of the heat exchanger to attain effective utilization of the heat. In this case, heat can be taken out stably even when the storage amount fluctuates and heat releasing amount is reduced, and improvement of heat transfer promotes the cooling of the containing bodies, which enables minimization of the size of the storage vessel. (T.M.)

  1. Proof testing of an explosion containment vessel

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

  2. Evaluation of buckling on containment metallic vessels

    International Nuclear Information System (INIS)

    Silveira, Renato Campos da; Mattar Neto, Miguel

    2000-01-01

    The buckling analysis represents one of the most important aspects of the structural projects of nuclear power plants containment metallic vessels and in this work the Case N-284-1 ASME Code is used for evaluation of those structures submitted to this failure mode. From the stress analysis, performed by using finite element method on discrete structures with shell elements, the procedure of the Code Case are applied to the evaluation of the containment metallic vessel of the Angra 2 nuclear power plant submitted to the own weight, seismic loads and uplift in case of accident. A study of pressure vessel reinforced by rings submit ed to the external pressure. Conclusions and commentaries are established based on the obtained results

  3. Water content monitoring for Flamanville 3 EPR trademark prestressed concrete containment. An application for TDR techniques

    Energy Technology Data Exchange (ETDEWEB)

    Courtois, Alexis; Clauzon, Timothee [EDF DPIH DTG, Lyon (France); Taillade, Frederic [EDF R and D, Chatou (France); Martin, Gregoire [EDF CNEN, Montrouge (France)

    2015-07-01

    Long term operation of nuclear power plant requires an appropriate monitoring of containment structures. For prestressed concrete containment vessels, a key parameter for ageing analysis is the evolution of the amount of water remaining within the concrete pores. EDF decides to launch a development program, in order to determine what sensor technologies are able to achieve such kind of monitoring on large concrete structures. One of the main parts of this program is to determine the maximum allowable uncertainty for the measurement. Another stake is the calibration process of sensors dedicated to water content measurement in concrete structures and the management of the parameters which have the largest influence on the measurement process.

  4. Inelastic analysis of prestressed concrete secondary containments

    International Nuclear Information System (INIS)

    Murray, D.W.; Chitnuyanondh, L.; Wong, C.; Rijub-Agha, K.Y.

    1978-07-01

    An elastic-plastic constitutive model for the simulation of stress-strain response of concrete under any biaxial combination of compressive and/or tensile stresses is developed. An effective tensile stress-strain curve is obtained indirectly from experimental results of a test on a large scale prestressed concrete wall segment. These concrete properties are then utilized in predicting the response of a second test and the results compared with the experiment. Modificications to the BOSOR5 program, in order to incorporate the new constitutive relation into it, are described. Techniques of modelling structures in order to perform inelastic analysis of thin shell axisymmetric prestressed concrete secondary containments are investigated. The results of inelastic BOSOR5 analyses of two different models of the University of Alberta Test Structure are presented. The predicted deterioration of the structure and the limit states associated with its behaviour are determined and discussed. It is concluded that the technique is a practical one which can be used for the inelastic analysis of Gentilly-type containment structures. (author)

  5. TEMP-STRESS analysis of a reinforced concrete vessel under internal pressure

    International Nuclear Information System (INIS)

    Marchertas, A.H.; Kennedy, J.M.; Pfeiffer, P.A.

    1987-01-01

    The TEMP-STRESS FEM represents an axisymmetric simulation of the reinforced concrete vessel to internal pressurization. The information shows the global deformation, the state of strain/stress within the containment vessel with respect to the imposed pressures. Thus, the location and progress of concrete cracking, the stretching of the liner and the reinforcing bars and final failure are indicated through the entire loading range. Equilibrium of the entire system is assured at definite loading increments. With the progress of concrete cracking, the resisting load is continuously transferred to the reinforcing bars and the liner. Thus, after the tensile strength is exceeded and the concrete stress is set to zero, the internal pressures are entirely resisted by the liner and the reserve strength of the reinforcing bars. The reinforcing bars are mechanically connected to each other by splices, the ultimate strength of which is less than that of the rebars themselves. The corresponding strain at this limiting stress is lower than the ultimate strain of the liner. Therefore, the specified ultimate strength of the splices limits the pressurization of the vessel. Furthermore, once any of the splices fail, then load is transferred to the adjacent members, causing their failure and general failure of the vessel. (orig./HP)

  6. Concrete benchmark experiment: ex-vessel LWR surveillance dosimetry

    International Nuclear Information System (INIS)

    Ait Abderrahim, H.; D'Hondt, P.; Oeyen, J.; Risch, P.; Bioux, P.

    1993-09-01

    The analysis of DOEL-1 in-vessel and ex-vessel neutron dosimetry, using the DOT 3.5 Sn code coupled with the VITAMIN-C cross-section library, showed the same C/E values for different detectors at the surveillance capsule and the ex-vessel cavity positions. These results seem to be in contradiction with those obtained in several Benchmark experiments (PCA, PSF, VENUS...) when using the same computational tools. Indeed a strong decreasing radial trend of the C/E was observed, partly explained by the overestimation of the iron inelastic scattering. The flat trend seen in DOEL-1 could be explained by compensating errors in the calculation such as the backscattering due to the concrete walls outside the cavity. The 'Concrete Benchmark' experiment has been designed to judge the ability of this calculation methods to treat the backscattering. This paper describes the 'Concrete Benchmark' experiment, the measured and computed neutron dosimetry results and their comparison. This preliminary analysis seems to indicate an overestimation of the backscattering effect in the calculations. (authors). 5 figs., 1 tab., 7 refs

  7. Concrete containment modeling and management, Conmod

    International Nuclear Information System (INIS)

    Jovall, O.; Larsson, J.-A.; Shaw, P.; Touret, J.-P.; Karlberg, G.

    2003-01-01

    The CONMOD project aims to create a system which will ensure that safety requirements for concrete containment structures will be up-held during the entire planned lifetime of plants and possibly during an extended lifetime. An important part of the project is to develop the application and understanding of Non-Destructive Testing (NDT) techniques for the assessment of conformity and condition of concrete reactor containments and to integrate this with state-of-the-art and developed Finite Element (FE) modelling techniques and analysis of structural behaviour. The objective being to create a diagnostic method for evaluation of ageing and degradation of concrete containments. This method, the C ONMOD-methodology , will help in the planning and execution of actions that will improve safety in a manner which is optimal both in terms of economy and safety. The knowledge gained during the project will be presented in a handbook of best practice. The decommissioned Barsebaeck unit 1 reactor containment will be accessible for non-destructive examination throughout the duration of the project. Intrusive investigations will also be made including coring and material tests as a valuable complement to NDT. (author)

  8. Method for the construction of a nuclear reactor with a prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.

    1981-01-01

    Method for the construction of nuclear reactors with prestressed concrete pressure vessel, providing during the initial stage of construction of the prestressed concrete pressure vessel a support structure around the liner. This enables an early mounting of core components in clean conditions as well as load reductions for final concreting in layers of the prestressed concrete pressure vessel. By applying the support structure, the overall assembly time of these nuclear power plant is considerably reduced without extra cost. (orig.) [de

  9. Storage vessel for containing radiation contaminated material

    International Nuclear Information System (INIS)

    Ogawa, Kazuya.

    1995-01-01

    A container pipe and an outer pipe are coaxially assembled integrally in a state where securing spacers are disposed between the container pipe and the outer pipe, and an annular flow channel is formed around the container pipe. Radiation contaminated material-containing body (glass solidified package) is contained in the container pipe. The container pipe and the outer pipe in an integrated state are suspended from a ceiling plug of a cell chamber of a storage vessel, and supporting devices are assembled between the pipes and a support structure. A shear/lug mechanism is used for the supporting devices. The combination of the shear/lug allows radial and vertical movement but restrict horizontal movement of the outer tube. The supporting devices are assembled while visually recognizing the state of the shear/lug mechanism between the outer pipe and the support mechanism. Accordingly, operationability upon assembling the container pipe and the outer pipe is improved. (I.N.)

  10. Fast Generation of Container Vessel Stowage Plans

    DEFF Research Database (Denmark)

    Pacino, Dario

    that the vessel is stable and seaworthy, and at the same time arrange the cargo such that the time at port is minimized. Moreover, stowage coordinators only have a limited amount of time to produce the plan. This thesis addresses the question of whether it is possible to automatically generate stowage plans...... test instances provided by a major liner shipping company. Improvements to the modeling of vessel stability and an analysis of its accuracy together with an analysis of the computational complexity of the container stowage problem are also included in the thesis, resulting in an overall in...

  11. Closure system of a vessel made of prestressed concrete

    International Nuclear Information System (INIS)

    Audibert, Alain

    1974-01-01

    The present invention relates to removable plugs of prestressed concrete which can be fitted to every type of closed high pressure vessels and especially to the cylindrical vessels of nuclear reactors. The method involved permits the plug to be fitted to the vessel through both radial and axial prestress. In this purpose, said invention proposes removable prestress ribs fitted inside sheaths in the plug and extending throughout the upper part of the bearing surfaces of the plug, said ribs being regularly arranged along the generators of an hyperboloid of one sheet. Owing to this important feature, that is to say said inclination of the ribs in accordance with the generators of said hyperboloid, said rib inclination can be changed on requirement for each realization [fr

  12. Proof testing of CANDU concrete containment structures

    International Nuclear Information System (INIS)

    Pandey, M.D.

    1996-05-01

    Prior to commissioning of a CANDU reactor, a proof pressure test is required to demonstrate the structural integrity of the containment envelope. The test pressure specified by AECB Regulatory Document R-7 (1991) was selected without a rigorous consideration of uncertainties associated with estimates of accident pressure and conatinment resistance. This study was undertaken to develop a reliability-based philosophy for defining proof testing requirements that are consistent with the current limit states design code for concrete containments (CSA N287.3).It was shown that the upodated probability of failure after a successful test is always less than the original estimate

  13. Containment vessel for a nuclear reactor

    International Nuclear Information System (INIS)

    Yamanari, Sh.; Horiuchi, T.; Sugisaki, T.; Tominaga, K.

    1985-01-01

    A containment vessel for a nuclear reactor having a dry well for mounting therein a pressure vessel for containing the nuclear reactor, a pressure suppressing chamber having a pool of coolant therein, and a vent pipe device for releasing therethrough into the pool of coolant within the pressure suppressing chamber steam which will be produced as a result of the occurrence of an accident and escape into the dry well. The vent pipe device includes a plurality of vent pipe members inserted in the pool of coolant within the pressure suppressing chamber and each having at least one exhaust port opening in the coolant. The vent pipe members are divided into a plurality of groups in such a manner that the vent pipe members of different groups differ from one another in the length of submerged portions of the vent pipe members interposed between the liquid of the coolant within the pressure suppressing chamber and the exhaust ports of the vent pipe members

  14. Containing method for spent fuel and spent fuel containing vessel

    International Nuclear Information System (INIS)

    Maekawa, Hiromichi; Hanada, Yoshine.

    1996-01-01

    Upon containing spent fuels, a metal vessel main body and a support spacer having fuel containing holes are provided. The support spacer is disposed in the inside of the metal vessel main body, and spent fuel assemblies are loaded in the fuel containing holes. Then, a lid is welded at the opening of the metal vessel main body to provide a sealing state. In this state, heat released from the spent fuel assemblies is transferred to the wall of the metal vessel main body via the support spacer. Since the support spacer has a greater heat conductivity than gases, heat of the spent fuel assemblies tends to be released to the outside, thereby capable of removing heat of the spent fuel assemblies effectively. In addition, since the surfaces of the spent fuel assemblies are in contact with the inner surface of the fuel containing holes of the support spacer, impact-resistance and earthquake-resistance are ensured, and radiation from the spent fuel assemblies is decayed by passing through the layer of the support spacer. (T.M.)

  15. Concrete containment integrity program at EPRI

    International Nuclear Information System (INIS)

    Winkleblack, R.K.; Tang, Y.K.

    1984-01-01

    Many in the nuclear power plant business believe that the catastrophic failure mode for reactor containment structures is unrealistic. One of the goals of the EPRI containment integrity program is to demonstrate that this is true. The objective of the program is to provide the utility industry with an experimental data base and a test-validated analytical method for realistically evaluating the actual over-pressure capability of concrete containment buildings and to predict leakage behavior if higher pressures were to occur. The ultimate goal of this research effort is to characterize the containment leakage mode and rate as a function of internal pressure and time so that the risk can be realistically assessed for hypothetical degraded core accidents. Progress in the first and second phases of the three-phase analytical and testing efforts is discussed

  16. Research and development of the prestressed concrete reactor vessel

    International Nuclear Information System (INIS)

    Shiozawa, Shoji; Omata, Ippei; Nakamura, Norio

    1975-01-01

    Compared with the steel reactor vessel, the prestressed concrete reactor vessel (PCRV) is said to be superior in safety and economy. One of the characteristics of the high temperature gas cooled reactor (HTGR) is the adoption of the PCRV instead of the steel reactor vessel to ensure safety. In order to improve safety characteristics, it is necessary for the PCRV to be provided with more reliable functions. When the multi-purpose HTGR or the gas cooled fast breeder reactor (GCFR) are realized in future, more severe conditions of technology will be imposed on the PCRV, and accordingly, technical developments are now increasingly required. IHI is now proceeding with the technical research and development on the PCRV, in which a basic study of its liner cooling system has already been completed. In this study applying a large cylindrical PCRV model, comparison was made between experimental data and analyses concerning the liner cooling system, and the results of analytical technique have been evaluated. The analytical technique established this time is applicable to the estimation of temperature distribution in the concrete of a large PCRV and also to the evaluation of the liner cooling system. (auth.)

  17. Nonlinear analysis of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

    Connor, J.J.

    1975-01-01

    The numerical procedures for predicting the nonlinear behavior of a prestressed concrete reactor vessel over its design life are discussed. The numerical models are constructed by combining three-dimensional isoparametric finite elements which simulate the concrete, thin shell elements which simulate steel linear plates, and layers of reinforcement steel, and axial elements for discrete prestressing cables. Nonlinearity under compressive stress, multi-dimensional cracking, shrinkage and stress/temperature induced creep of concrete are considered in addition to the elasti-plastic behavior of the liner and reinforcing steel. Various failure theories for concrete have been proposed recently. Also, there are alternative strategies for solving the discrete system equations over the design life, accounting for test loads, pressure and temperature operational loads, creep unloading and abnormal loads. The proposed methods are reviewed, and a new formulation developed by the authors is described. A number of comparisons with experimental tests results and other numerical schemes are presented. These examples demonstrate the validity of the formulation and also provide valuable information concerning the cost and accuracy of the various solution strategies i.e., total vs. incremental loading and initial vs. tangent stiffness. Finally, the analysis of an actual PCRV is described. Stress contours and cracking patterns in the region of cutouts corresponding to operational pressure and temperature loads are illustrated. The effects of creep, unloading, and creep recovery are then shown. Lastly, a strategy for assessing the performance over its design life is discussed

  18. Advances in the analysis and design of concrete structures, metal containments and liner plate for extreme loads

    International Nuclear Information System (INIS)

    Stevenson, J.D.; Eibl, J.; Curbach, M.; Johnson, T.E.; Daye, M.A.; Riera, J.D.; Nemet, J.; Iyengar, K.T.S.

    1992-01-01

    The material presented in this paper summarizes the progress that has been made in the analysis, design, and testing of concrete structures. The material is summarized in the following documents: Part I: Containment Design Criteria and Loading Combinations; Part II: Reinforced and Prestressed Concrete Behavior; Part III: Concrete Containment Analysis, Design and Related Testing; Part IV: Impact and Impulse Loading and Response Prediction; Part V: Metal Containments and Liner Plate Systems; Part VI: Prestressed Reactor Vessel Design, Testing and Analysis. (orig.)

  19. Analysis of failures in concrete containments

    International Nuclear Information System (INIS)

    Moreno-Gonzalez, A.

    1989-09-01

    The function of Containment, in an accident event, is to avoid the release of radioactive substances into the surroundings. Containment failure, therefore, is defined as the appearance of leak paths to the external environment. These leak paths may appear either as a result of loss of leaktightness due to degradation of design conditions or structural failure with containment material break. This document is a survey of the state of the art of Containment Failure Analysis. It gives a detailed description of all failure mechanisms, indicating all the possible failure modes and their causes, right from failure resulting from degradation of the materials to structural failure and linear breake failure. Following the description of failure modes, possible failure criteria are identified, with special emphasis on structural failure criteria. These criteria have been obtained not only from existing codes but also from the latest experimental results. A chapter has been dedicated exclusively to failure criteria in conventional structures, for the purpose of evaluating the possibility of application to the case of containment. As the structural behaviour of the containment building is very complex, it is not possible to define failure through a single parameter. It is therefore advisable to define a methodology for containment failure analysis which could be applied to a particular containment. This methodology should include prevailing load and material conditions together with the behaviour of complex conditions such as the liner-anchorage-cracked concrete interaction

  20. Long-term properties of concrete in nuclear containment structures

    International Nuclear Information System (INIS)

    Field, S.N.; Bamforth, P.B.

    1991-01-01

    Over the last thirty years a large volume of testing has been carried out on concretes used in prestressed concrete pressure vessels and similar structures. The main aim of the work has been to provide the designers with a prediction method for elastic moduli and creep deformation which takes into account temperature and age at loading. This paper summarises and reviews the results from the six concretes tested by Taywood Engineering Ltd (T.E.L.), comparing mixes with and without PFA. (author)

  1. Mechanical properties of concrete containing recycled concrete aggregate (RCA) and ceramic waste as coarse aggregate replacement

    Science.gov (United States)

    Khalid, Faisal Sheikh; Azmi, Nurul Bazilah; Sumandi, Khairul Azwa Syafiq Mohd; Mazenan, Puteri Natasya

    2017-10-01

    Many construction and development activities today consume large amounts of concrete. The amount of construction waste is also increasing because of the demolition process. Much of this waste can be recycled to produce new products and increase the sustainability of construction projects. As recyclable construction wastes, concrete and ceramic can replace the natural aggregate in concrete because of their hard and strong physical properties. This research used 25%, 35%, and 45% recycled concrete aggregate (RCA) and ceramic waste as coarse aggregate in producing concrete. Several tests, such as concrete cube compression and splitting tensile tests, were also performed to determine and compare the mechanical properties of the recycled concrete with those of the normal concrete that contains 100% natural aggregate. The concrete containing 35% RCA and 35% ceramic waste showed the best properties compared with the normal concrete.

  2. Roles of concrete technology for containment of radioactive contaminants

    International Nuclear Information System (INIS)

    Kitsutaka, Yoshinori; Imamoto, Keiichi

    2014-01-01

    A large amount of radioactive materials was emitted in the environment by the reactor accident at Fukushima Daiichi Nuclear Power Plant. Nuclear debris still remains in the reactor container. An investigative committee was organized in Japan Concrete Institute to study on the containment of radioactive materials and the safe utilization of concrete materials. We have investigated the effect of the hydrogen explosion upon the property of concrete and the transfer of materials into the concrete. We also present the outline of the advice made by Japan Concrete Institute about technologies on the concrete materials for the waterproofing in buildings and for water-shielding walls. (J.P.N.)

  3. Seismic analysis of a containment vessel

    International Nuclear Information System (INIS)

    Toledo, E.M.; Jospin, R.J.; Loula, A.F.D.

    1987-01-01

    A seismic analysis of a nuclear power plant containment vessel is presented. Usual loads in this kind of analysis like SSE, DBE and SSB loadings are considered. With the response spectra, previously obtained, for the above mentioned loadings one uses the response spectrum techniques in order to obtain estimatives for the maximum values of the stresses. Some considerations about the problem and the approcah used herein, are initially described. Next, the analysed structure geometry and some results, compared with those obtained by using computer code ANSYS are shown. (Author) [pt

  4. Crack analysis of multicavity prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Gallix, R.; Liu, T.C.; Lu, S.C.H.

    1975-01-01

    A new method to perform the crack analysis of non-axisymmetric, multicavity prestressed concrete reactor vessels (PCRV's) subjected to hypothetical overpressure by using an axisymmetric two-dimensional finite element computer code is presented. Concrete, steel liner, bonded reinforcing steel and prestressing steel elements are modeled. The limiting tensile strain criterion is adopted for concrete cracking. The steel elements are assumed to be elastic/perfectly plastic. Von Mises yield criterion and Prandtl-Reuss flow equations define the behavior of the liner in the range of plastic deformations. An orthotropic stress-strain constitutive law is utilized for cracked concrete elements. To account for the presence of penetrations and secondary cavities in the PCRV, a modified finite element model based on the concept of effective moduli is adopted. The pressure in these cavities is simulated by equivalent axisymmetric pressure distributions. In the analysis, the pressure is applied incrementally. For a given pressure, the displacements, strains, and stresses are computed. The state of strains or stresses is then examined against the cracking or yield criteria. If cracking or yield is indicated, the stiffness and load matrices for the cracked and yielding elements are recomputed and a new equilibrium is sought. This procedure is repeated until the desired convergence of the solution is achieved. The validity of the adopted approach utilizing the two-dimensional finite element method for overpressure analyses of non-axisymmetric PCRV's is demonstrated through comparisons with two multicavity PCRV scale models. A reliable and conservative estimate of PCRV behavior under overpressure is obtained

  5. Finite element analysis of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Smith, P.D.; Cook, W.A.; Anderson, C.A.

    1977-01-01

    This paper discusses the development of a finite element code suitable for the safety analysis of prestressed concrete reactor vessels. The project has involved modification of a general purpose computer code to handle reinforced concrete structures as well as comparison of results obtained with the code against published experimental data. The NONSAP nonlinear structural analysis program was selected for the ease with which it can be modified to encompass problems peculiar to nuclear reactors. Pre- and post-processors have been developed for mesh generation and for graphical display of response variables. An out-of-core assembler and solver have been developed for the analysis of large three dimensional problems. The constitutive model for short term loads forms an orthotropic stress-strain relationship in which the concrete and the reinforcing steel are treated as a composite. The variation of stiffness and strength of concrete under multiaxial stress states is accounted for. Cracks are allowed to form at element integration points based on a three dimensional failure envelope in stress space. Composite tensile and shear properties across a crack are modified to account for bond degradation and for dowel action of the reinforcement. The constitutive law for creep is base on the expansion of the usual creep compliance function in the form of a Dirichlet exponential series. Empirical creep data are then fit to the Dirichlet series approximation by means of a least squares procedure. The incremental deformation process is subsequently reduced to a series of variable stiffness elasticity problems in which the past stress history is represented by a finite number of hidden material variables

  6. Cylindrical reinforced-concrete pressure vessel for nuclear reactors

    International Nuclear Information System (INIS)

    Vaessen, F.

    1975-01-01

    The cylindrical pressure vessel has got a wall and an isolating layer composed of blocks of heat-resistant concrete or of ceramic material. The side of the isolating layer facing the interior of the presssure vessel is coated by a liner made of metallic material. In cold state and without internal pressure, the radius of this liner is smaller by a differential amount than that of the isolating layer. By means of radially displaceable fixing elements consisting of an anchoring tube and a holding tube inserted in it, the liner can be made to rest against the isolating layer. This occurs if the pressure vessel is brought to operational temperature. The anchoring tube is attached to the isolating layer whereas the displaceable holding tube is connected with the liner. The possible relative travelling distance of these two elements is equal to the difference of length of the two radii. In addition, the liner may consist of single parts connected with each other through compensating flanges. There may also be additional springs arranged between the isolating layer and the liner. (DG/PB) [de

  7. Analyses of a steel containment vessel with an outer contact structure under severe internal overpressurization conditions

    International Nuclear Information System (INIS)

    Porter, V.L.

    1993-01-01

    Many Mark-I and Mark-II BWR plants are designed with a steel vessel as the primary containment. Typically, the steel containment vessel (SCV) is enclosed within a reinforced concrete shield building with only a small gap (50--90mm) separating the two structures. This paper describes finite element analyses performed to evaluate the effects of contact and friction between a steel containment vessel and an outer contact structure when the containment vessel is subjected to large internal pressures. These computations were motivated by a joint program on containment integrity involving the Nuclear Power Engineering Corporation (NUPEC) of Japan, the US Nuclear Regulatory Commission (NRC), and Sandia National Laboratories for testing model containments

  8. Properties of concrete containing foamed concrete block waste as fine aggregate replacement

    Science.gov (United States)

    Muthusamy, K.; Budiea, A. M. A.; Zaidan, A. L. F.; Rasid, M. H.; Hazimmah, D. S.

    2017-11-01

    Environmental degradation due to excessive sand mining dumping at certain places and disposal of foamed concrete block waste from lightweight concrete producing industry are issues that should be resolved for a better and cleaner environment of the community. Thus, the main intention of this study is to investigate the potential of foamed concrete block waste as partial sand replacement in concrete production. The foamed concrete waste (FCW) used in this research that were supplied by a local lightweight concrete producing industry. The workability and compressive strength of concrete containing various percentage of foamed concrete waste as partial sand replacement has been investigated. Prior to the use, the foamed concrete waste were crushed to produce finer particles. Six concrete mixes containing various content of crushed foamed concrete waste that are 0%, 10%, 20%, 30%, 40% and 50% were used in this experimental work. Then the prepared specimens were placed in water curing until the testing age. Compressive strength test and flexural strength tests were conducted at 7, 14 and 28 days. The result shows that integration of crushed foamed concrete waste as partial sand replacement in concrete reduces the mix workability. It is interesting to note that both compressive strength and flexural strength of concrete improves when 30% crushed foamed concrete waste is added as partial sand replacement.

  9. Safety analysis of nuclear containment vessels subjected to strong earthquakes and subsequent tsunamis

    Directory of Open Access Journals (Sweden)

    Feng Lin

    2017-08-01

    Full Text Available Nuclear power plants under expansion and under construction in China are mostly located in coastal areas, which means they are at risk of suffering strong earthquakes and subsequent tsunamis. This paper presents a safety analysis for a new reinforced concrete containment vessel in such events. A finite element method-based model was built, verified, and first used to understand the seismic performance of the containment vessel under earthquakes with increased intensities. Then, the model was used to assess the safety performance of the containment vessel subject to an earthquake with peak ground acceleration (PGA of 0.56g and subsequent tsunamis with increased inundation depths, similar to the 2011 Great East earthquake and tsunami in Japan. Results indicated that the containment vessel reached Limit State I (concrete cracking and Limit State II (concrete crushing when the PGAs were in a range of 0.8–1.1g and 1.2–1.7g, respectively. The containment vessel reached Limit State I with a tsunami inundation depth of 10 m after suffering an earthquake with a PGA of 0.56g. A site-specific hazard assessment was conducted to consider the likelihood of tsunami sources.

  10. Safety analysis of nuclear containment vessels subjected to strong earthquakes and subsequent tsunamis

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Feng; Li, Hong Zhi [Dept. Structural Engineering, Tongji University, Shanghai (China)

    2017-08-15

    Nuclear power plants under expansion and under construction in China are mostly located in coastal areas, which means they are at risk of suffering strong earthquakes and subsequent tsunamis. This paper presents a safety analysis for a new reinforced concrete containment vessel in such events. A finite element method-based model was built, verified, and first used to understand the seismic performance of the containment vessel under earthquakes with increased intensities. Then, the model was used to assess the safety performance of the containment vessel subject to an earthquake with peak ground acceleration (PGA) of 0.56g and subsequent tsunamis with increased inundation depths, similar to the 2011 Great East earthquake and tsunami in Japan. Results indicated that the containment vessel reached Limit State I (concrete cracking) and Limit State II (concrete crushing) when the PGAs were in a range of 0.8–1.1g and 1.2–1.7g, respectively. The containment vessel reached Limit State I with a tsunami inundation depth of 10 m after suffering an earthquake with a PGA of 0.56g. A site-specific hazard assessment was conducted to consider the likelihood of tsunami sources.

  11. Towards Better Understanding of Concrete Containing Recycled Concrete Aggregate

    Directory of Open Access Journals (Sweden)

    Hisham Qasrawi

    2013-01-01

    Full Text Available The effect of using recycled concrete aggregates (RCA on the basic properties of normal concrete is studied. First, recycled aggregate properties have been determined and compared to those of normal aggregates. Except for absorption, there was not a significant difference between the two. Later, recycled aggregates were introduced in concrete mixes. In these mixes, natural coarse aggregate was partly or totally replaced by recycled aggregates. Results show that the use of recycled aggregates has an adverse effect on the workability and air content of fresh concrete. Depending on the water/cement ratio and on the percent of the normal aggregate replaced by RCA, the concrete strength is reduced by 5% to 25%, while the tensile strength is reduced by 4% to 14%. All results are compared with previous research. As new in this research, the paper introduces a simple formula for the prediction of the modulus of elasticity of RCA concrete. Furthermore, the paper shows the variation of the air content of RAC.

  12. Behaviour of prestressed concrete containment structures

    International Nuclear Information System (INIS)

    MacGregor, J.G.; Murray, D.W.; Simmonds, S.H.

    1980-05-01

    The most significant finds from a study to assess the response of prestressed concrete secondary containment structures for nuclear reactors under the influence of high internal overpressures are presented. A method of analysis is described for determining the strains and deflections including effects of inelastic behaviour at various points in the structure resulting from increasing internal pressures. Experimentally derived relationships between the strains and crack spacing, crack width and leakage rate are given. These procedures were applied to the Gentilly-2 containment building to obtain the following results: (1) The first through-the-wall cracks would occur in the dome at 48 psi or 2.3 times the proof test pressure. (2) At this pressure leakage would begin and would increase exponentially as the pressure increases such that at 93% of the predicted failure load the calculated leakage rate would be approximately equal to the volume of the containment each second. (3) Assuming the pressurizing medium could be supplied sufficiently rapidly, failure would occur due to rupture of the horizontal tendons at approximately 77 psi. (author)

  13. Biaxial Loading Tests for steel containment vessel

    Energy Technology Data Exchange (ETDEWEB)

    Miyagawa, T. [Nuclear Power Engineering Corp., Tokyo (Japan); Wright, D.J.; Arai, S.

    1999-07-01

    The Nuclear Power Engineering Corporation (NUPEC) has conducted a 1/10 scale of the steel containment vessel (SCV) test for the understanding of ultimate structural behavior beyond the design pressure condition. Biaxial Loading Tests were supporting tests for the 1/10 scale SCV model to evaluate the method of estimating failure conditions of thin steel plates under biaxial loading conditions. The tentative material models of SGV480 and SPV490 were obtained. And the behavior of SGV480 and SPV490 thin steel plates under biaxial loading conditions could be well simulated by FE-Analyses with the tentative material models and Mises constitutive law. This paper describes the results and the evaluations of these tests. (author)

  14. Biaxial Loading Tests for steel containment vessel

    International Nuclear Information System (INIS)

    Miyagawa, T.; Wright, D.J.; Arai, S.

    1999-01-01

    The Nuclear Power Engineering Corporation (NUPEC) has conducted a 1/10 scale of the steel containment vessel (SCV) test for the understanding of ultimate structural behavior beyond the design pressure condition. Biaxial Loading Tests were supporting tests for the 1/10 scale SCV model to evaluate the method of estimating failure conditions of thin steel plates under biaxial loading conditions. The tentative material models of SGV480 and SPV490 were obtained. And the behavior of SGV480 and SPV490 thin steel plates under biaxial loading conditions could be well simulated by FE-Analyses with the tentative material models and Mises constitutive law. This paper describes the results and the evaluations of these tests. (author)

  15. Welding the AT-400A Containment Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Brandon, E.

    1998-11-01

    Early in 1994, the Department of Energy assigned Sandia National Laboratories the responsibility for designing and providing the welding system for the girth weld for the AT-400A containment vessel. (The AT-400A container is employed for the shipment and long-term storage of the nuclear weapon pits being returned from the nation's nuclear arsenal.) Mason Hanger Corporation's Pantex Plant was chosen to be the production facility. The project was successfully completed by providing and implementing a turnkey welding system and qualified welding procedure at the Pantex Plant. The welding system was transferred to Pantex and a pilot lot of 20 AT-400A containers with W48 pits was welded in August 1997. This document is intended to bring together the AT-400A welding system and product (girth weld) requirements and the activities conducted to meet those requirements. This document alone is not a complete compilation of the welding development activities but is meant to be a summary to be used with the applicable references.

  16. Prestressed concrete reactor vessel thermal cylinder model study

    International Nuclear Information System (INIS)

    Callahan, J.P.; Canonico, D.A.; Richardson, M.; Corum, J.M.; Dodge, W.G.; Robinson, G.C.; Whitman, G.D.

    1977-01-01

    The thermal cylinder experiment was designed both to provide information for evaluating the capability of analytical methods to predict the time-dependent stress-strain behavior of a 1 / 6 -scale model of the barrel section of a single-cavity prestressed concrete reactor vessel and to demonstrate the structural behavior under design and off-design thermal conditions. The model was a thick-walled cylinder having a height of 1.22 m, a thickness of 0.46 m, and an outer diameter of 2.06 m. It was prestressed both axially and circumferentially and subjected to 4.83 MPa internal pressure together with a thermal crossfall imposed by heating the inner surface to 338.8 K and cooling the outer surface to 297.1 K. The initial 460 days of testing were divided into time periods that simulated prestressing, heatup, reactor operation, and shutdown. At the conclusion of the simulated operating period, the model was repressurized and subjected to localized heating at 505.4 K for 84 days to produce an off-design hot-spot condition. Comparisons of experimental data with calculated values obtained using the SAFE-CRACK finite-element computer program showed that the program was capable of predicting time-dependent behavior in a vessel subjected to normal operating conditions, but that it was unable to accurately predict the behavior during off-design hot-spot heating. Readings made using a neutron and gamma-ray backscattering moisture probe showed little, if any, migration of moisture in the concrete cross section. Destructive examination indicated that the model maintained its basic structural integrity during localized hot-spot heating

  17. Minimum weight design of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

    Boes, R.

    1975-01-01

    A method of non-linear programming for the minimization of the volume of rotationally symmetric prestressed concrete reactor pressure vessels is presented. It is assumed that the inner shape, the loads and the degree of prestressing are prescribed, whereas the outer shape is to be detemined. Prestressing includes rotational and vertical tension. The objective function minimizes the weight of the PCRV. The constrained minimization problem is converted into an unconstrained problem by the addition of interior penalty functions to the objective function. The minimum is determined by the variable metric method (Davidson-Fletcher-Powell), using both values and derivatives of the modified objective function. The one-dimensional search is approximated by a method of Kund. Optimization variables are scaled. The method is applied to a pressure vessel like for THTR. It is found that the thickness of the cylindrical wall may be reduced considerably for the load cases considered in the optimization. The thickness of the cover is reduced slightly. The largest reduction in wall thickness occurs at the junction of wall and cover. (Auth.)

  18. Manufacturing method for radioactive material containing vessel

    International Nuclear Information System (INIS)

    Kamino, Yoshikazu; Nishioka, Eiji; Toyota, Michinori.

    1997-01-01

    A containing vessel for radioactive materials (for example, spent fuels) comprises an inner cylinder made of stainless steel having a space for containing radioactive materials at the inside and an outer cylinder made of stainless steel disposed at the outer side of the inner cylinder. Lead homogenization is applied to a space between the inner and the outer cylinders to deposit a lead layer. Then, molten lead heated to a predetermined temperature is cast into the space between the inner and the outer cylinders. A valve is opened to discharge the molten lead in the space from a molten lead discharge pipe, and heated molten lead is injected from a molten lead supply pipe. Then, the discharge of the molten lead and the injection of the molten lead are stopped, and the lead in the space is coagulated. With such procedures, gaps are not formed between the lead of the homogenized portion and the lead of cast portion even when the thickness of the inner and the outer cylinders is great. (I.N.)

  19. Shear strength of end slabs of prestressed concrete nuclear reactor vessels

    International Nuclear Information System (INIS)

    Reins, J.D.; Quiros, J.L. Jr.; Schnobrich, W.C.; Sozen, M.A.

    1976-07-01

    The report summarizes the experimental and part of the analytical work carried out in connection with an investigation of the structural strength of prestressed concrete reactor vessels. The project is part of the Prestressed Concrete Reactor Vessel Program of the Oak Ridge National Laboratory sponsored by ERDA. The objective of the current phase of the work is to develop procedures to determine the shear strength of flat end slabs of reactor vessels with penetrations

  20. Improvement to reactor vessel

    International Nuclear Information System (INIS)

    1974-01-01

    The vessel described includes a prestressed concrete vessel containing a chamber and a removable cover closing this chamber. The cover is in concrete and is kept in its closed position by main and auxiliary retainers, comprising fittings integral with the concrete of the vessel. The auxiliary retainers pass through the concrete of the cover. This improvement may be applied to BWR, PWR and LMFBR type reactor vessel [fr

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

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

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

  2. Durability of heavyweight concrete containing barite

    International Nuclear Information System (INIS)

    Binici, Hanifi

    2010-01-01

    The supplementary waste barite aggregates deposit in Osmaniye, southern Turkey, has been estimated at around 500 000 000 tons based on 2007 records. The aim of the present study is to investigate the durability of concrete incorporating waste barite as coarse and river sand (RS), granule blast furnace slag (GBFS), granule basaltic pumice (GBP) and ≤ 4 mm granule barite (B) as fine aggregates. The properties of the fresh concrete determined included the air content, slump, slump loss and setting time. They also included the compressive strength, flexural and splitting tensile strengths and Young's modulus of elasticity, resistance to abrasion and sulphate resistance of hardened concrete. Besides these, control mortars were prepared with crushed limestone aggregates. The influence of waste barite as coarse aggregates and RS, GBFS, GBP and B as fine aggregates on the durability of the concretes was evaluated. The mass attenuation coefficients were calculated at photon energies of 1 keV to 100 GeV using XCOM and the obtained results were compared with the measurements at 0.66 and 1.25 MeV. The results showed the possibility of using these waste barite aggregates in the production of heavy concretes. In several cases, some of these properties have been improved. Durability of the concrete made with these waste aggregates was improved. Thus, these materials should be preferably used as aggregates in heavyweight concrete production. (orig.)

  3. Manufacturing method for radioactive material containing vessel

    International Nuclear Information System (INIS)

    Nishioka, Hideharu; Matsushita, Kazuo; Toyota, Michinori.

    1997-01-01

    Lead homogenization is applied on the inner surface of a space formed between an inner cylinder and an outer cylinder, and a molten lead heated to about 400 to 500degC is cast into a space formed between the inner cylinder and the outer cylinder in a state where the inner and the outer cylinders are heated to from 200 to 300degC. The space formed between the inner cylinder and the outer cylinder is heated to and kept at 330degC or higher for at least 2minutes after the casting of the molten lead, and then it is cooled. Thus, lowering of density of the molten lead due to excess elevation of temperature or dropping of the lead at the homogenization portion by heating the inner and the outer cylinders to an excessively high temperature are not caused. In addition, formation of gaps in the boundary between the inner cylinder and the outer cylinder or between the lead of the homogenized portion and that of the cast portion due to the melting of the lead of the homogenized portion in the space is prevented reliably thereby capable of forming a satisfactory shielding member. Then, even when the thickness of the inner cylinder and the outer cylinder is large, radioactive material containing vessel excellent in heat releasing property and radiation shielding property can be manufactured. (N.H.)

  4. Analyses of a steel containment vessel with an outer contact structure under severe internal overpressurization conditions

    International Nuclear Information System (INIS)

    Porter, V.L.

    1994-01-01

    Many Mark-I and Mark-II BWR plants are designed with a steel vessel as the primary containment. Typically, the steel containment vessel (SCV) is enclosed within a reinforced concrete shield building with only a small gap (74-90 mm) separating the two structures. This paper describes finite element analyses performed to evaluate the effects of contact and friction between a steel containment vessel and an outer contact structure when the containment vessel is subjected to large internal pressures. These computations were motivated by a joint program on containment integrity involving the Nuclear Power Engineering Corporation (NUPEC) of Japan, the US Nuclear Regulatory Commission (NRC), and Sandia National Laboratories for testing model containments. Under severe accident loading conditions, the steel containment vessel in a typical Mark-I or Mark-II plant may deform under internal pressurization such that it contacts the inner surface of a shield building wall. (Thermal expansion from increasing accident temperatures would also close the gap between the SCV and the shield building, but temperature effects are not considered in these analyses.) The amount and location of contact and the pressure at which it occurs all affect how the combined structure behaves. A preliminary finite element model has been developed to analyze a model of a typical steel containment vessel con-ling into contact with an outer structure. Both the steel containment vessel and the outer contact structure were modelled with axisymmetric shell finite elements. Of particular interest are the influence that the contact structure has on deformation and potential failure modes of the containment vessel. Furthermore, the coefficient of friction between the two structures was varied to study its effects on the behavior of the containment vessel and on the uplift loads transmitted to the contact structure. These analyses show that the material properties of an outer contact structure and the amount

  5. Design and analysis of reactor containment of steel-concrete composite laminated shell

    International Nuclear Information System (INIS)

    Ichikawa, K.

    1977-01-01

    Reinforced and prestressed concrete containments for reactors have been developed in order to avoid the difficulties of welding of steel containments encountered as their capacities have become large: growing thickness of steel shells gave rise to the requirement of stress relief at the construction sites. However, these concrete vessels also seem to face another difficulty: the lack of shearing resistance capacity. In order to improve the shearing resistance capacity of the containment vessel, while avoiding the difficulty of welding, a new scheme of containment consisting of steel-concrete laminated shell is being developed. In the main part of a cylindrical vessel, the shell consists of two layers of thin steel plates located at the inner and outer surfaces, and a layer of concrete core into which both the steel plates are anchored. In order to validate the feasibility and safety of this new design, the results of analysis on the basis of up-to-date design loads are presented. The results of model tests in 1:30 scale are also reported. (Auth.)

  6. Engineering Performance of High Strength Concrete Containing Steel Fibre Reinforcement

    Directory of Open Access Journals (Sweden)

    Md Azree Othuman Mydin

    2013-09-01

    Full Text Available The development and utilization of the high strength concrete in the construction industry have been increasing rapidly. Fiber reinforced concrete is introduced to overcome the weakness of the conventional concrete because concrete normally can crack under a low tensile force and it is known to be brittle. Steel fibre is proved to be the popular and best combination in the high strength concrete to result the best in the mechanical and durability properties of high strength concrete with consideration of curing time, steel fibre geometry, concrete grade and else more. The incorporation of steel fibre in the mortar mixture is known as steel fibre reinforced concrete have the potential to produce improvement in the workability, strength, ductility and the deformation of high strength concrete. Besides that, steel fibre also increases the tensile strength of concrete and improves the mechanical properties of the steel fibre reinforced concrete. The range for any high strength concrete is between 60MPa-100MPa. Steel fibre reinforced concrete which contains straight fibres has poorer physical properties than that containing hooked end stainless steel fibre due to the length and the hooked steel fibre provide a better effective aspects ratio. Normally, steel fibre tensile strength is in the range of 1100MPa-1700MPa. Addition of less steel fibre volumes in the range of 0.5% to 1.0% can produce better increase in the flexural fatigue strength. The strength can be increased with addition of steel fibre up to certain percentage. This paper will review and present some basic properties of steel fibre reinforced concrete such as mechanical, workability and durability properties.

  7. Analyses and testing of model prestressed concrete reactor vessels with built-in planes of weakness

    International Nuclear Information System (INIS)

    Dawson, P.; Paton, A.A.; Fleischer, C.C.

    1990-01-01

    This paper describes the design, construction, analyses and testing of two small scale, single cavity prestressed concrete reactor vessel models, one without planes of weakness and one with planes of weakness immediately behind the cavity liner. This work was carried out to extend a previous study which had suggested the likely feasibility of constructing regions of prestressed concrete reactor vessels and biological shields, which become activated, using easily removable blocks, separated by a suitable membrane. The paper describes the results obtained and concludes that the planes of weakness concept could offer a means of facilitating the dismantling of activated regions of prestressed concrete reactor vessels, biological shields and similar types of structure. (author)

  8. Models and Algorithms for Container Vessel Stowage Optimization

    DEFF Research Database (Denmark)

    Delgado-Ortegon, Alberto

    .g., selection of vessels to buy that satisfy specific demands), through to operational decisions (e.g., selection of containers that optimize revenue, and stowing those containers into a vessel). This thesis addresses the question of whether it is possible to formulate stowage optimization models...... container of those to be loaded in a port should be placed in a vessel, i.e., to generate stowage plans. This thesis explores two different approaches to solve this problem, both follow a 2-phase decomposition that assigns containers to vessel sections in the first phase, i.e., master planning...

  9. EDS V25 containment vessel explosive qualification test report.

    Energy Technology Data Exchange (ETDEWEB)

    Rudolphi, John Joseph

    2012-04-01

    The V25 containment vessel was procured by the Project Manager, Non-Stockpile Chemical Materiel (PMNSCM) as a replacement vessel for use on the P2 Explosive Destruction Systems. It is the first EDS vessel to be fabricated under Code Case 2564 of the ASME Boiler and Pressure Vessel Code, which provides rules for the design of impulsively loaded vessels. The explosive rating for the vessel based on the Code Case is nine (9) pounds TNT-equivalent for up to 637 detonations. This limit is an increase from the 4.8 pounds TNT-equivalency rating for previous vessels. This report describes the explosive qualification tests that were performed in the vessel as part of the process for qualifying the vessel for explosive use. The tests consisted of a 11.25 pound TNT equivalent bare charge detonation followed by a 9 pound TNT equivalent detonation.

  10. The 1500 MW fast breeder reactor the double envelope-vessel anchored in concrete

    International Nuclear Information System (INIS)

    Bolvin, M.

    1981-01-01

    This paper givers an account of EDF investigations to reduce the investment costs of the 1500 MW Fast Reactor (RNR 1500) without prejudice to the safety requirements. It deals with the double envelope-vessel, designed to minimize radiation consequences in the event of accidental leakage in the main vessel. In the Fast Reactors in operation (PHOENIX), under construction (CRYS-MALVILLE), and under project (NR 1500), the double envelope-steel vessel hangs down from the upper part of the reactor block, its weight being approximately 300 t. In the new design, the vessel is fixed into the concrete which supports the main vessel, by means of steel anchors. A thermal insulation isolates it from the main vessel. The installation of coils in the concrete, next to the lining, allows for water circulation to cool the concrete. (orig./GL)

  11. Sealing method and sealing device for radioactive waste containing vessel

    International Nuclear Information System (INIS)

    Ishiwatari, Koji; Otsuki, Akira

    1998-01-01

    A radioactive waste-containing body is hoisted down into a strong-material vessel opened upwardly, and a strong-material lid is hoisted down to the opening of the strong-material-vessel and welded. The strong material vessel is hoisted up and loaded on a corrosion resistant-material bottom plate placed horizontally. A corrosion resistant-material vessel having one opening end and having a corrosion resistant-material flange on the other end and previously agreed with the strong material-vessel main body is hoisted up by a hoisting device having an inserting device so that the opening of the corrosion resistant vessel is directed downwardly. The corrosion resistant vessel is press-fitted to the outside of the strong material-vessel by the inserting device while being heated by a preheater to shrink. Subsequently, the lower end of the corrosion resistant-material vessel and the corrosion resistant-material bottom plate are welded to constitute a corrosion resistant-material vessel. Then, the radioactive waste containing body can be sealed in a sealing vessel comprising the strong-material vessel and the corrosion resistant-material vessel. (N.H.)

  12. Radon emanation fractions from concretes containing fly ash and metakaolin

    International Nuclear Information System (INIS)

    Taylor-Lange, Sarah C.; Juenger, Maria C.G.; Siegel, Jeffrey A.

    2014-01-01

    Radon ( 222 Rn) and progenies emanate from soil and building components and can create an indoor air quality hazard. In this study, nine concrete constituents, including the supplementary cementitious materials (SCMs) fly ash and metakaolin, were used to create eleven different concrete mixtures. We investigated the effect of constituent radium specific activity, radon effective activity and emanation fraction on the concrete emanation fraction and the radon exhalation rate. Given the serious health effects associated with radionuclide exposure, experimental results were coupled with Monte Carlo simulations to demonstrate predictive differences in the indoor radon concentration due to concrete mixture design. The results from this study show that, on average, fly ash constituents possessed radium specific activities ranging from 100 Bq/kg to 200 Bq/kg and emanation fractions ranging from 1.1% to 2.5%. The lowest emitting concrete mixture containing fly ash resulted in a 3.4% reduction in the concrete emanation fraction, owing to the relatively low emanation that exists when fly ash is part of concrete. On average, the metakaolin constituents contained radium specific activities ranging from 67 Bq/kg to 600 Bq/kg and emanation fractions ranging from 8.4% to 15.5%, and changed the total concrete emanation fraction by roughly ± 5% relative to control samples. The results from this study suggest that SCMs can reduce indoor radon exposure from concrete, contingent upon SCM radionucleotide content and emanation fraction. Lastly, the experimental results provide SCM-specific concrete emanation fractions for indoor radon exposure modeling. - Highlights: • Fly ash or metakaolin SCMs can neutralize or reduce concrete emanation fractions. • The specific activity of constituents is a poor predictor of the concrete emanation fraction. • Exhalation from fly ash concretes represents a small fraction of the total indoor radon concentration

  13. Time varying stress in ligaments of perforated plates with reference to prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Stefanou, G.D.

    1978-01-01

    The work described herein relates to the prediction of stresses in materials which exhibit time varying strains with particular reference to the ligaments of perforated circular concrete slabs, subjected to long-term radial prestress and uniform elevated temperature. The perforations are reinforced with steel liners and arranged in a square central lattice symmetrical about two orthogonal axes. Special reference is made to the distribution of stress in the standpipe region of prestressed concrete cylindrical pressure or containment vessels for gas cooled reactors. In order to assess the stress distribution around the perforated zone of a circular slab, a method of analysis was developed by the author, based on the ''Equivalent Elastic Modulus'' of the perforated zone and the ''Effective Modulus Method'', utilizing experimental data obtained from tests performed on model specimens. The object of this paper is to extend the above method of analysis into the perforated region, and assess the long-term stresses in the ligaments. The proposed method is accomplished by an application of the Finite Element Method for the elastic plane stress case. Comparisons of experimental results and theoretical predictions by the proposed method, and other analytical methods are made for a series of perforated concrete slabs subjected to radial in-plane loading: 10,342 kN/m 2 (1,5000 psi), and uniform elevated temperature of 80 0 C. The investigation, though in general terms, could be applied to the perforated region of cylindrical pressure vessels for nuclear reactors. Finally the paper describes briefly in Appendix 3 a direct solution procedure for calculating time dependent stresses in concrete structures based on the principles of variational calculus. Analytical predictions obtained by the proposed method which is a step-by-step analysis, are compared with the variational principle method. (author)

  14. Analytical investigation of multicavity prestressed concrete pressure vessels for elastic loading conditions

    International Nuclear Information System (INIS)

    Fanning, D.N.

    1978-09-01

    A three-dimensional finite-element analysis of a commercial high-temperature gas-cooled reactor (HTGR) was made using the finite-element code STATIC-SAP. Four loading conditions were analyzed elastically to evaluate the behavior of the concentric core prestressed concrete reactor vessel (PCRV) of the HTGR. The results of the analysis were evaluated in accordance with Section III, Division 2, of the ASME Code for Reactor Vessels and Containments. The calculated maximum stresses were found to be well within the Code-allowable values. The analysis was preceded by an evaluation of candidate computer codes using comparisons of experimental data with analytical results for the Ohbayashi-Gumi multicavity PCRV model. This vessel was chosen as a basis for comparison because of its geometrical similarity to the large multicavity PCRV and the anticipated availability of a complete set of the original experimental data. The three-dimensional finite-element codes NONSAP and STATIC-SAP were used for the analysis of the Ohbayashi-Gumi vessel

  15. Fukushima Daiichi Unit 1 Ex-Vessel Prediction: Core Concrete Interaction

    International Nuclear Information System (INIS)

    Robb, Kevin R; Farmer, Mitchell; Francis, Matthew W

    2015-01-01

    Lower head failure and corium concrete interaction were predicted to occur at Fukushima Daiichi Unit 1 (1F1) by several different system-level code analyses, including MELCOR v2.1 and MAAP5. Although these codes capture a wide range of accident phenomena, they do not contain detailed models for ex-vessel core melt behavior. However, specialized codes exist for analysis of ex-vessel melt spreading (e.g., MELTSPREAD) and long-term debris coolability (e.g., CORQUENCH). On this basis, an analysis was carried out to further evaluate ex-vessel behavior for 1F1 using MELTSPREAD and CORQUENCH. Best-estimate melt pour conditions predicted by MELCOR v2.1 and MAAP5 were used as input. MELTSPREAD was then used to predict the spatially dependent melt conditions and extent of spreading during relocation from the vessel. The results of the MELTSPREAD analysis are reported in a companion paper. This information was used as input for the long-term debris coolability analysis with CORQUENCH.

  16. Material characteristics and construction methods for a typical research reactor concrete containment in Iran

    International Nuclear Information System (INIS)

    Ebrahimia, Mahsa; Suha, Kune Y.; Eghbalic, Rahman; Jahan, Farzaneh Asadi malek

    2012-01-01

    Generally selecting an appropriate material and also construction style for a concrete containment due to its function and special geometry play an important role in applicability and also construction cost and duration decrease in a research reactor (RR) project. The reactor containment enclosing the reactor vessel comprises physical barriers reflecting the safety design and construction codes, regulations and standards so as to prevent the community and the environment from uncontrolled release of radioactive materials. It is the third and the last barrier against radioactivity release. It protects the reactor vessel from such external events as earthquake and aircraft crash as well. Thus, it should be designed and constructed in such a manner as to withstand dead and live loads, ground and seismic loads, missiles and aircraft loads, and thermal and shrinkage loads. This study aims to present a construction method for concrete containment of a typical RR in Iran. The work also presents an acceptable characteristic for concrete and reinforcing re bar of a typical concrete containment. The current study has evaluated the various types of the RR containments. The most proper type was selected in accordance with the current knowledge and technology of Iran

  17. Material characteristics and construction methods for a typical research reactor concrete containment in Iran

    Energy Technology Data Exchange (ETDEWEB)

    Ebrahimia, Mahsa; Suha, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of); Eghbalic, Rahman; Jahan, Farzaneh Asadi malek [School of Architecture and Urbanism, Qazvin (Iran, Islamic Republic of)

    2012-10-15

    Generally selecting an appropriate material and also construction style for a concrete containment due to its function and special geometry play an important role in applicability and also construction cost and duration decrease in a research reactor (RR) project. The reactor containment enclosing the reactor vessel comprises physical barriers reflecting the safety design and construction codes, regulations and standards so as to prevent the community and the environment from uncontrolled release of radioactive materials. It is the third and the last barrier against radioactivity release. It protects the reactor vessel from such external events as earthquake and aircraft crash as well. Thus, it should be designed and constructed in such a manner as to withstand dead and live loads, ground and seismic loads, missiles and aircraft loads, and thermal and shrinkage loads. This study aims to present a construction method for concrete containment of a typical RR in Iran. The work also presents an acceptable characteristic for concrete and reinforcing re bar of a typical concrete containment. The current study has evaluated the various types of the RR containments. The most proper type was selected in accordance with the current knowledge and technology of Iran.

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

    International Nuclear Information System (INIS)

    Watson, P.S.

    1990-01-01

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

  19. Concrete containment integrity software: Procedure manual and guidelines

    International Nuclear Information System (INIS)

    Dameron, R.A.; Dunham, R.S.; Rashid, Y.R.

    1990-06-01

    This report is an executive summary describing the concrete containment analysis methodology and software that was developed in the EPRI-sponsored research to predict the overpressure behavior and leakage of concrete containments. A set of guidelines has been developed for performing reliable 2D axisymmetric concrete containment analysis with a cracking concrete constitutive model developed by ANATECH. The software package developed during this research phase is designed for use in conjunction with ABAQUS-EPGEN; it provides the concrete model and automates axisymmetric grid preparation, and rebar generation for 2D and 3D grids. The software offers the option of generating pre-programmed axisymmetric grids that can be tailored to a specific containment by input of a few geometry parameters. The goal of simplified axisymmetric analysis within the framework of the containment leakage prediction methodology is to compute global liner strain histories at various locations within the containment. A simplified approach for generating peak liner strains at structural discontinuities as function of the global liner strains has been presented in a separate leakage criteria document; the curves for strain magnification factors and liner stress triaxiality factors found in that document are intended to be applied to the global liner strain histories developed through global 2D analysis. This report summarizes the procedures for global 2D analysis and gives an overview of the constitutive model and the special purpose concrete containment analysis software developed in this research phase. 8 refs., 10 figs

  20. Properties of Pervious Concrete Containing Scrap Tyre Tubes

    Directory of Open Access Journals (Sweden)

    Boon Koh Heng

    2017-01-01

    Full Text Available There is a huge quantity of waste tyre tubes generated every year due to the increasing of motorcycle user. Therefore, recycling of the waste tyre tubes is become mandatory. The aim of this research was to study the properties of pervious concrete containing scrap tyre tube (STT rubber particles with percentages of 3%, 5% and 7% of the cement content. The properties studied are void content, compressive strength measured at 7, 14 and 28 days, flexural strength and flow rate which were determined at 28 day. The experimental results showed that, there were increased in void content and flow rate of pervious concrete containing STT. Both compressive strength and flexural strength of pervious concrete containing STT showed a lower value compared to the control mix without STT. The reductions of the mechanical strengths are likely due to the increase of void content. Overall, pervious concrete which contains 7% STT has shown an increment of mechanical strengths and flow rate compared to other STT pervious concrete. Nonetheless, the results indicate that there are potentials for use of STT in pervious concrete, especially for use in pervious concrete applications such as pavements, driveways and parking lots.

  1. Float level switch for a nuclear power plant containment vessel

    International Nuclear Information System (INIS)

    Powell, J.G.

    1993-01-01

    This invention is a float level switch used to sense rise or drop in water level in a containment vessel of a nuclear power plant during a loss of coolant accident. The essential components of the device are a guide tube, a reed switch inside the guide tube, a float containing a magnetic portion that activates a reed switch, and metal-sheathed, ceramic-insulated conductors connecting the reed switch to a monitoring system outside the containment vessel. Special materials and special sealing techniques prevent failure of components and allow the float level switch to be connected to a monitoring system outside the containment vessel. 1 figures

  2. Float level switch for a nuclear power plant containment vessel

    Science.gov (United States)

    Powell, James G.

    1993-01-01

    This invention is a float level switch used to sense rise or drop in water level in a containment vessel of a nuclear power plant during a loss of coolant accident. The essential components of the device are a guide tube, a reed switch inside the guide tube, a float containing a magnetic portion that activates a reed switch, and metal-sheathed, ceramic-insulated conductors connecting the reed switch to a monitoring system outside the containment vessel. Special materials and special sealing techniques prevent failure of components and allow the float level switch to be connected to a monitoring system outside the containment vessel.

  3. Development of polymer concrete radioactive waste management containers

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.; Lee, M. S.; Ahn, D. H.; Won, H. J.; Kang, H. S.; Lee, H. S.; Lim, S.P.; Kim, Y. E.; Lee, B. O.; Lee, K. P.; Min, B. Y.; Lee, J.K.; Jang, W. S.; Sim, W. B.; Lee, J. C.; Park, M. J.; Choi, Y. J.; Shin, H. E.; Park, H. Y.; Kim, C. Y

    1999-11-01

    A high-integrity radioactive waste container has been developed to immobilize the spent resin wastes from nuclear power plants, protect possible future, inadvertent intruders from damaging radiation. The polymer concrete container is designed to ensure safe and reliable disposal of the radioactive waste for a minimum period of 300 years. A built-in vent system for each container will permit the release of gas. An experimental evaluation of the mechanical, chemical, and biological tests of the container was carried out. The tests showed that the polymer concrete container is adequate for safe disposal of the radioactive wastes. (author)

  4. The dynamic relaxation method in the structural analysis of concrete pressure vessels

    International Nuclear Information System (INIS)

    Davidson, I.; Assis Bastos, M.R. de; Camargo, P.B. de.

    1977-01-01

    The dynamic relaxation method, applied to 3 dimensional concrete structures, especially pressure vessels, is demonstrated. It utilizes the finite difference method and allows the growth of cracks to be followed up to the point of vessel rupture. A FORTRAN IV program is developed, which can also be utilized, with the necessary modifications, for other structure calculations [pt

  5. Compressive strength of concrete and mortar containing fly ash

    Science.gov (United States)

    Liskowitz, John W.; Wecharatana, Methi; Jaturapitakkul, Chai; Cerkanowicz, deceased, Anthony E.

    1997-01-01

    The present invention relates to concrete, mortar and other hardenable mixtures comprising cement and fly ash for use in construction. The invention includes a method for predicting the compressive strength of such a hardenable mixture, which is very important for planning a project. The invention also relates to hardenable mixtures comprising cement and fly ash which can achieve greater compressive strength than hardenable mixtures containing only concrete over the time period relevant for construction. In a specific embodiment, a formula is provided that accurately predicts compressive strength of concrete containing fly ash out to 180 days. In other specific examples, concrete and mortar containing about 15% to 25% fly ash as a replacement for cement, which are capable of meeting design specifications required for building and highway construction, are provided. Such materials can thus significantly reduce construction costs.

  6. Effect of high temperature on integrity of concrete containment structures

    International Nuclear Information System (INIS)

    Bhat, P.D.

    1986-01-01

    The effect of high temperature on concrete material properties and structural behavior are studied in order to relate these effects to the performance of concrete containment structures. Salient data obtained from a test program undertaken to study the behavior of a restrained concrete structure under thermal gradient loads up to its ultimate limit are described. The preliminary results indicate that concrete material properties can be considered to remain unaltered up to temperatures of 100 0 C. The presence of thermal gradients did not significantly affect the structures ultimate mechanical load capacity. Relaxation of restraint forces due to creep was found to be an important factor. The test findings are compared with the observations made in available literature. The effect of test findings on the integrity analysis of a containment structure are discussed. The problem is studied from the viewpoint of a CANDU heavy water reactor containment

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  8. State-of-the-Art Report on Molten Corium Concrete Interaction and Ex-Vessel Molten Core Coolability

    International Nuclear Information System (INIS)

    Bonnet, Jean-Michel; Cranga, Michel; Vola, Didier; Marchetto, Cathy; Kissane, Martin; ); Robledo, Fernando; Farmer, Mitchel T.; Spengler, Claus; Basu, Sudhamay; Atkhen, Kresna; Fargette, Andre; Fisher, Manfred; Foit, Jerzi; Hotta, Akitoshi; Morita, Akinobu; Journeau, Christophe; Moiseenko, Evgeny; Polidoro, Franco; Zhou, Quan

    2017-01-01

    Activities carried out over the last three decades in relation to core-concrete interactions and melt coolability, as well as related containment failure modes, have significantly increased the level of understanding in this area. In a severe accident with little or no cooling of the reactor core, the residual decay heat in the fuel can cause the core materials to melt. One of the challenges in such cases is to determine the consequences of molten core materials causing a failure of the reactor pressure vessel. Molten corium will interact, for example, with structural concrete below the vessel. The reaction between corium and concrete, commonly referred to as MCCI (molten core concrete interaction), can be extensive and can release combustible gases. The cooling behaviour of ex-vessel melts through sprays or flooding is also complex. This report summarises the current state of the art on MCCI and melt coolability, and thus should be useful to specialists seeking to predict the consequences of severe accidents, to model developers for severe-accident computer codes and to designers of mitigation measures

  9. Experience of in-service surveillance and monitoring of prestressed concrete pressure vessels for nuclear reactors

    International Nuclear Information System (INIS)

    Irving, J.; Smith, J.R.; Eadie, D.McD.; Hornby, I.W.

    1976-01-01

    Details are given of the statutory requirements for the inspection of prestressed concrete pressure vessels in the United Kingdom, with particular emphasis on the prestressing system. The results of periodic examinations under the Licencing Conditions of the Oldbury and Wylfa vessels are presented and discussed in relation to design expectations and future requirements. Strain, moisture and temperature records obtained from the Oldbury PCPV's over a 10 year period are compared with prediction and new developments in vessel instrumentation are discussed. (author)

  10. Radon emanation fractions from concretes containing fly ash and metakaolin.

    Science.gov (United States)

    Taylor-Lange, Sarah C; Juenger, Maria C G; Siegel, Jeffrey A

    2014-01-01

    Radon ((222)Rn) and progenies emanate from soil and building components and can create an indoor air quality hazard. In this study, nine concrete constituents, including the supplementary cementitious materials (SCMs) fly ash and metakaolin, were used to create eleven different concrete mixtures. We investigated the effect of constituent radium specific activity, radon effective activity and emanation fraction on the concrete emanation fraction and the radon exhalation rate. Given the serious health effects associated with radionuclide exposure, experimental results were coupled with Monte Carlo simulations to demonstrate predictive differences in the indoor radon concentration due to concrete mixture design. The results from this study show that, on average, fly ash constituents possessed radium specific activities ranging from 100 Bq/kg to 200 Bq/kg and emanation fractions ranging from 1.1% to 2.5%. The lowest emitting concrete mixture containing fly ash resulted in a 3.4% reduction in the concrete emanation fraction, owing to the relatively low emanation that exists when fly ash is part of concrete. On average, the metakaolin constituents contained radium specific activities ranging from 67 Bq/kg to 600 Bq/kg and emanation fractions ranging from 8.4% to 15.5%, and changed the total concrete emanation fraction by roughly ±5% relative to control samples. The results from this study suggest that SCMs can reduce indoor radon exposure from concrete, contingent upon SCM radionucleotide content and emanation fraction. Lastly, the experimental results provide SCM-specific concrete emanation fractions for indoor radon exposure modeling. © 2013.

  11. Seismic transient analysis of a containment vessel with penetrations

    International Nuclear Information System (INIS)

    Dahlke, H.J.; Weiner, E.O.

    1979-12-01

    A linear transient analysis of the FFTF containment vessel was conducted with STAGS to justify the load levels used for the seismic qualification testing of the heating and ventiliation valve operators. The modeling consists of a thin axisymmetric shell for the containment vessel with four penetrations characterized by linear and rotational inertias as well as attachment characteristics to the shell. Motions considered are horizontal, rocking and vertical input to the base, and the solution is carried out by direct integration. Results show that the test levels and the approximate analyses considered are conservative. Response spectra for some containment vessel penetrations applicable to the model are presented

  12. Tests on concrete containing cork powder admixtures

    Directory of Open Access Journals (Sweden)

    Guerra, I.

    2007-06-01

    Full Text Available The present study aimed to determine the physical and mechanical properties of laboratory concrete made with different proportions of cork powder. While the resulting material lacked the mechanical strength characteristic of concrete, its properties may prove to be apt for certain hardscaping and agricultural uses, such as in the manufacture of pavement for playgrounds and parks, or certain kinds of structures used in livestock raising. These findings need to be analyzed and verified.Este trabajo de investigación tiene por objeto conocer algunas propiedades físicas y mecánicas de un hormigón elaborado en laboratorio, adicionándole diversas proporciones de polvo de corcho. Las propiedades del material resultante, si bien carecen de la resistencia mecánica que caracteriza al hormigón, parecen interesantes para su uso en ciertas aplicaciones de la ingeniería agronómica tales como en la fabricación de piezas para solados de parques infantiles y jardines, o en los cubículos de ciertas construcciones ganaderas, extremos que es preciso analizar y comprobar.

  13. Determining prestressing forces for inspection of prestressed concrete containments

    International Nuclear Information System (INIS)

    1990-07-01

    General Design Criterion 53, ''Provisions for Containment Testing and Inspection,'' of Appendix A, ''General Design Criteria for Nuclear Power Plants,'' to 10 CFR Part 50, ''Domestic Licensing of Production and Utilization Facilities,'' requires, in part, that the reactor containment be designed to permit (1) periodic inspection of all important areas and (2) an appropriate surveillance program. Regulatory Guide 1.35, ''Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containment Structures,'' describes a basis acceptable to the NRC staff for developing an appropriate inservice inspection and surveillance program for ungrouted tendons in prestressed concrete containment structures of light-water-cooled reactors. This guide expands and clarifies the NRC staff position on determining prestressing forces to be used for inservice inspections of prestressed concrete containment structures

  14. Concept study for a combined reinforced concrete containment

    International Nuclear Information System (INIS)

    Liersch, G.; Peter, U.; Danisch, R.; Freiman, M.; Hummer, M.; Roettinger, H.; Hansen, H.

    1994-01-01

    A variety of different steel and concrete containment types had been designed and constructed in the past. Most of the concrete containments had been prestressed offering the advantage of small displacements and certain leak tightness of the concrete itself. However, considerable stresses in concrete as well as in the tendons have to be maintained during the whole lifetime of the plant in order to guarantee the required prestressing. The long-time behaviour and the ductility in case of beyond design load cases must be verified. In contrary to a prestressed containment a reinforced containment will only significantly be loaded during test conditions or when needed in case of accidents. It offers additional margins which can be used especially for dynamic loads like impacts or for beyond design considerations. The aim of this paper is to show the feasibility of a so-called combined containment which means capable to resist both - severe internal accidents and external hazards mainly the aircraft crash impact as considered in the design of nuclear power plants in Germany. The concept is a lined reinforced containment without prestressing. The mechanical resistance function is provided by the reinforced concrete and the leak tightness function will be taken by a so called composite liner made of non-metallic materials. Some results of tests performed at SIEMENS laboratories and at the University of Karlsruhe which show the capability of a composite liner to bridge over cracks at the concrete surface will be presented in the paper. The study shows that the combined reinforced concrete containment with a composite liner offers a robust concept with high flexibility with respect to load requirements, beyond design considerations and geometrical shaping (arrangement of openings, integration with adjacent structures). The concept may be further optimized by partial prestressing at areas of high concentration of stresses such as at transition zones or at disturbances around

  15. Behaviour of a pre-stressed concrete pressure-vessel subjected to a high temperature gradient

    International Nuclear Information System (INIS)

    Dubois, F.

    1965-01-01

    After a review of the problems presented by pressure-vessels for atomic reactors (shape of the vessel, pressures, openings, foundations, etc.) the advantages of pre-stressed concrete vessels with respect to steel ones are given. The use of pre-stressed concrete vessels however presents many difficulties connected with the properties of concrete. Thus, because of the absence of an exact knowledge of the material, it is necessary to place a sealed layer of steel against the concrete, to have a thermal insulator or a cooling circuit for limiting the deformations and stresses, etc. It follows that the study of the behaviour of pre-stressed concrete and of the vessel subjected- to a high temperature gradient can yield useful information. A one-tenth scale model of a pre-stressed concrete cylindrical vessel without any side openings and without a base has been built. Before giving a description of the tests the authors consider some theoretical aspects concerning 'scale model-actual structure' similitude conditions and the calculation of the thermal and mechanical effects. The pre-stressed concrete model was heated internally by a 'pyrotenax' element and cooled externally by a very strong air current. The concrete was pre-stressed using horizontal and vertical cables held at 80 kg/cm 2 ; the thermal gradient was 160 deg. C. During the various tests, measurements were made of the overall and local deformations, the changes in water content, the elasticity modulus, the stress and creep of the cables and the depths of the cracks. The overall deformations observed are in line with thermal deformation theories and the creep of the cables attained 20 to 30 per cent according to their position relative to the internal surface. The dynamic elasticity modulus decreased by half but the concrete keeps its good mechanical properties. Finally, cracks 8 to 12 cm deep and 2 to 3 mms wide appeared in that part of the concrete which was not pre-stressed. The results obtained make it

  16. Influence of Silicon-Containing Additives on Concrete Waterproofness Property

    Science.gov (United States)

    Butakova, M. D.; Saribekyan, S. S.; Mikhaylov, A. V.

    2017-11-01

    The article studies the influence of silicon-containing additives on the property of the water resistance of concrete samples. It provides a review of the literature on common approaches and technologies improving concrete waterproofness and reinforced concrete structures. Normal hardening samples were obtained on the basis of concretes containing microsilica, aerosil or ash, or the combinations thereof. This research is aimed at the study of the complex modifier effect r on the basis of metakaolin, superplasticizer and silicon containing additives on the property of concrete water resistance. The need to use a superplasticizer to reduce the water-cement ratio and metakaolin as a hardening accelerator along with the set of strength is substantiated. This article describes a part of the results of the experiment conducted to find alternative options for colmatizing expensive additives used in the concreting foundations of private house-building. The implementation of the scientific work will not only clarify this area but will also broaden the knowledge of such additive as aerosol.

  17. Experimental verification of creep analyses for prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Aoyagi, Y.; Abe, H.; Ohnuma, H.

    1977-01-01

    The authors proposed a new method of creep analysis based on the theory of strain hardening, which assumes that accumulated creep at a given time influences the creep after that. This method was applied to calculate step-by-step the behaviors of uniaxial creep of concrete under variable temperatures and stresses, creep in reinforced concrete specimens and the behaviors of prestressed concrete beams under themal gradients. The experimental and calculated results agreed fairly well. Further, this method was incorporated in the finite element creep analysis for the prestressed concrete hollow cylinder and the full scale model. The calculated strain changes with time pursued closely those obtained by experiments. The above led to the conclusion that from the viewpoint of both accuracy and computation time the strain hardening method proposed by the authors may be judged advantageous for practical usages

  18. System for cooling the containment vessel of a nuclear reactor

    International Nuclear Information System (INIS)

    Costes, Didier.

    1982-01-01

    The invention concerns a post-accidental cooling system for a nuclear reactor containment vessel. This system includes in series a turbine fed by the moist air contained in the vessel, a condenser in which the air is dried and cooled, a compressor actuated by the turbine and a cooling exchanger. The cold water flowing through the condenser and in the exchanger is taken from a tank outside the vessel and injected by a pump actuated by the turbine. The application is for nuclear reactors under pressure [fr

  19. Fuel containing vessel for transporting nuclear fuel

    International Nuclear Information System (INIS)

    Yoshizawa, Hiroyasu; Shimizu, Fukuzo; Tanaka, Nobuyuki.

    1996-01-01

    A shock absorbing mechanism is disposed on an inner bottom of a vessel main body. The shock absorbing mechanism comprises a shock absorbing member disposed on the upper surface of a bottom wall, an annular metal plate disposed on the upper surface of the shock absorbing member and an annular spacer disposed on the upper surface of the metal plate. The shock absorbing member is made of a material such as of wood, lead, metal honeycomb or a metal mesh, which plastically deforms when applied with load higher than a predetermined level, and is formed in a square block-like form covering the upper surface of the bottom wall. The spacer is made of a thin soft material such as tetrafluoroethylene, and is formed in such a shape as capable of preventing direct contact of the lower end of the cylindrical member in a lower tie plate of nuclear fuels with the metal portion. This can ensure integrity of nuclear fuels even when they fall from a high place upon an assumed dropping accident. (I.N.)

  20. Transient thermal creep of nuclear reactor pressure vessel type concretes

    International Nuclear Information System (INIS)

    Khoury, G.A.

    1983-01-01

    The immediate aim of the research was to study the transient thermal strain behaviour of four AGR type nuclear reactor concretes during first time heating in an unsealed condition to 600 deg. C. The work being also relevant to applications of fire exposed concrete structures. The programme was, however, expanded to serve a second more theoretical purpose, namely the further investigation of the strain development of unsealed concrete under constant, transient and cyclic thermal states in particular and the effect of elevated temperatures on concrete in general. The range of materials investigated included seven different concretes and three types of cement paste. Limestone, basalt, gravel and lightweight aggregates were employed as well as OPC and SRC cements. Cement replacements included pfa and slag. Test variables comprised two rates of heating (0.2 and 1 deg. C/minute), three initial moisture contents (moist as cast, air-dry and oven dry at 105 deg. C), two curing regimes (bulk of tests represented mass cured concrete), five stress levels (0, 10, 20, 30 and a few tests at 60% of the cold strength), two thermal cycles and levels of test temperature up to 720 deg. C. Supplementary, dilatometry, TGA and DTA tests were performed at CERL on individual samples of aggregate and cement paste which helped towards explaining the observed trends in the concretes. A simple formula was developed which relates the elastic thermal stresses generated from radial temperature gradients to the solution obtained from the transient heat conduction equation. Thermal stresses can, therefore, be minimized by reductions in the radius of the specimen and the rate of heating The results were confirmed by finite element analysis which indicate( tensile stresses in the central region and compressive stresses near the surf ace during heating which are reversed during cooling. It is shown that the temperature gradients, pore pressures and tensile thermal stresses during both heating and

  1. Evaluation of temperature distribution in a containment vessel during operation

    International Nuclear Information System (INIS)

    Utanohara, Yoichi; Murase, Michio; Yanagi, Chihiro; Masui, Akihiro; Inomata, Ryo; Kamiya, Yuji

    2012-01-01

    For safety analysis of the containment vessel (CV) in a nuclear power plant, the average temperature of the gas phase in the CV during operation is used as an initial condition. An actual CV, however, has a temperature distribution, which makes the estimation of the average temperature difficult. Numerical simulation seems to be useful for the average temperature estimation, but it has several difficulties such as predictions of temperature distribution in a large and closed space that has several compartments, and modeling the heat generating components and the convection-diffusion of heat by ventilation air-conditioning systems. The main purpose of this study was to simulate the temperature distribution and evaluate the average temperature in the CV of a three-loop pressurized water reactor (PWR) during the reactor operation. The simulation considered the heat generation of equipment, flow due to the ventilation and air conditioning systems, heat loss to the CV exterior, and the solar heat. The predicted temperature distribution was significantly affected by the flow. Particularly, openings, which became flow paths, affected the temperature distribution. The temperature increased with a rise in height within the CV and the flow field seemed to transform from forced convection to natural convection. The volume-averaged temperature was different between gas and solid (concrete, CV wall) phases as well as between heights. The total volume-averaged temperature of the CV was nearly equal to the average gas phase temperature. It was found to be easy to evaluate the effect of openings on the temperature distribution and estimate the average temperature in CV by numerical simulation. (author)

  2. Random thermal stress in concrete containments

    International Nuclear Information System (INIS)

    Singh, M.P.; Heller, R.A.

    1980-01-01

    Currently, the overly conservative thermal design forces are obtained on the basis of simplified assumptions made about the temperature gradient across the containment wall. Using the method presented in this paper, a more rational and better estimate of the design forces can be obtained. Herein, the outside temperature is considered to consist of a constant mean on which yearly and daily harmonic changes plus a randomly varying part are superimposed. The random part is modeled as a stationary random process. To obtain the stresses due to random and harmonic temperatures, the complex frequency response function approach has been used. Numerical results obtained for a typical containment show that the higher frequency temperature variations, though of large magnitude, induce relatively small forces in a containment. Therefore, in a containment design, a rational separation of more effective, slowly varying temperatures, such as seasonal cycle from less effective but more frequently occuring daily and hourly changes, is desirable to obtain rational design forces. 7 refs

  3. Method of detecting water leakage in radioactive waste containing vessel

    International Nuclear Information System (INIS)

    Ishioka, Hitoshi; Takao, Yoshiaki; Hayakawa, Kiyoshige.

    1989-01-01

    Lower level radioactive wastes formed upon operation of nuclear facilities are processed by underground storage. In this case, a plurality of drum cans packed with radioactive wastes are contained in a vessel and a water soluble dye material is placed at the inside of the vessel. The method of placing the water soluble dye material at the inside of the vessel includes a method of coating the material on the inner surface of the vessel and a method of mixing the material in sands to be filled between each of the drum cans. Then, leakage of water soluble dye material is detected when water intruding from the outside into the vessel is again leached out of the vessel, to detect the water leakage from the inside of the vessel. In this way, it is possible to find a water-invaded vessel before corrosion of the drum can by water intruded into the vessel and leakage of nuclides in the drum can. Accordingly, it is possible to apply treatment such as repair before occurrence of accident and can maintain the safety of radioactive water processing facilities. (I.S.)

  4. Reinforced concrete containment structures in high seismic zones

    International Nuclear Information System (INIS)

    Aziz, T.S.

    1977-01-01

    A new structural concept for reinforced concrete containment structures at sites where earthquake ground motions in terms of the Safe Shutdown Earthquake (SSE) exceeds 0.3 g is presented. The structural concept is based on: (1) an inner steel-lined concrete shell which houses the reactor and provides shielding and containment in the event of loss of coolant accident; (2) an outer annular concrete shell structure which houses auxiliary reactor equipment and safeguards systems. These shell structures are supported on a common foundation mat which is embedded in the subgrade. Under stipulated earthquake conditions the two shell structures interact to resist lateral inertia forces. Thus the annular structure which is not a pressure boundary acts as a lateral support for the inner containment shell. The concept is practical, economically feasible and new to practice. (Auth.)

  5. Leakage of pressurized gases through unlined concrete containment structures

    International Nuclear Information System (INIS)

    Rizkalla, S.H.; Simmonds, S.H.

    1983-01-01

    Eight reinforced concrete specimens were fabricated and subjected to tensile membrane forces and air pressure to study the air leakage characteristics in cracked reinforced concrete members. A mathematical expression for the rate of pressurized air flowing through an idealized crack is presented. The mathematical expression is refined by using the experimental data to describe the air flow rate through any given crack pattern. Graphical charts are also presented for the calculation of the air leakage rate through concrete cracks. The concept of equivalent crack width for a given crack pattern is introduced. The mathematical expression and graphical charts are modified to include this equivalent crack width concept. The proposed technique is applicable for the prediction of the leakage from concrete containment structures or any similar structures due to high internal pressure sufficient to initiate cracking. (orig.)

  6. Failure/leakage predictions of concrete structures containing cracks

    International Nuclear Information System (INIS)

    Pan, Y.C.; Marchertas, A.H.; Kennedy, J.M.

    1984-06-01

    An approach is presented for studying the cracking and radioactive release of a reactor containment during severe accidents and extreme environments. The cracking of concrete is modeled as the blunt crack. The initiation and propagation of a crack are determined by using the maximum strength and the J-integral criteria. Furthermore, the extent of cracking is related to the leakage calculation by using a model developed by Rizkalla, Lau and Simmonds. Numerical examples are given for a three-point bending problem and a hypothetical case of a concrete containment structure subjected to high internal pressure during an accident

  7. Research requirements for improved design of reinforced concrete containment structures

    International Nuclear Information System (INIS)

    Banerjee, A.K.; Holley, M.J. Jr.

    1978-01-01

    Reinforced concrete is a competitive material for the construction of nuclear power plant containment structures. However, the designer is constrained by limited data on the behavior of certain construction details which require him to use what may be excessive rebar quantities and lead to difficult and costly construction. This paper discusses several design situations where research is recommended to increase the designer's options, to facilitate construction, and to extend the applicability of reinforced concrete to such changing containment requirements as may be imposed by an evolving nuclear technology. (Auth.)

  8. Plant life management of the ACR-1000 Concrete containment structure

    International Nuclear Information System (INIS)

    Abrishami, H.H.; Ricciuti, R.; Elgohary, M.

    2009-01-01

    The Ageing of reinforced concrete structures due to service conditions, aggressive environments, or accidents may cause their strength, serviceability and durability to decrease over time. For a new plant, a Plant Life Management (PLiM) program should start in the design process and then continues through the plant operation and decommissioning. Hence, PLiM must provide not only Ageing Management program (AMP) but also provide requirements on material characteristic and design criteria as well. The purpose of this paper is to present the Plant Life Management (PLiM) strategy for the concrete containment structure of the ACR-10001 (Advanced CANDU Reactor) designed by AECL. The ACR-1000 is designed for a 100-year plant life including 60-year operating life and an additional 40-year decommissioning period. The approach adopted for the PLiM strategy of the concrete containment structure is a preventive one, key areas being: 1) design methodology, 2) material performance and 3) ageing management program. During the design phase, in addition to strength and serviceability, durability, throughout the service life and decommissioning phase of the ACR-1000 structure, is a major consideration. Factors affecting durability design include: a) concrete performance, b) structural application, and c) consideration of environmental conditions. In addition to addressing the design methodology and material performance requirements, a systematic approach for the ageing management program for the concrete containment structure is presented. (authors)

  9. High stress monitoring of prestressing tendons in nuclear concrete vessels using fibre-optic sensors

    Energy Technology Data Exchange (ETDEWEB)

    Perry, M., E-mail: marcus.perry@strath.ac.uk [Institute for Energy and Environment, University of Strathclyde, 204 George Street, Glasgow G1 1XW (United Kingdom); Yan, Z.; Sun, Z.; Zhang, L. [Aston Institute of Photonic Technologies, Aston University, Birmingham B4 7ET (United Kingdom); Niewczas, P. [Institute for Energy and Environment, University of Strathclyde, 204 George Street, Glasgow G1 1XW (United Kingdom); Johnston, M. [Civil Design Group, EDF Energy, Nuclear Generation, East Kilbride G74 5PG (United Kingdom)

    2014-03-15

    Highlights: • We weld radiation-resistant optical fibre strain sensors to steel prestressing tendons. • We prove the sensors can survive 1300 MPa stress (80% of steel's tensile strength). • Mechanical relaxation of sensors is characterised under 1300 MPa stress over 10 h. • Strain transfer between tendon and sensor remains at 69% after relaxation. • Sensors can withstand and measure deflection of tendon around a 4.5 m bend radius. - Abstract: Maintaining the structural health of prestressed concrete nuclear containments is a key element in ensuring nuclear reactors are capable of meeting their safety requirements. This paper discusses the attachment, fabrication and characterisation of optical fibre strain sensors suitable for the prestress monitoring of irradiated steel prestressing tendons. The all-metal fabrication and welding process allowed the instrumented strand to simultaneously monitor and apply stresses up to 1300 MPa (80% of steel's ultimate tensile strength). There were no adverse effects to the strand's mechanical properties or integrity. After sensor relaxation through cyclic stress treatment, strain transfer between the optical fibre sensors and the strand remained at 69%. The fibre strain sensors could also withstand the non-axial forces induced as the strand was deflected around a 4.5 m bend radius. Further development of this technology has the potential to augment current prestress monitoring practices, allowing distributed measurements of short- and long-term prestress losses in nuclear prestressed-concrete vessels.

  10. Structural optimization of reinforced concrete container for radioactive wastes

    International Nuclear Information System (INIS)

    Tamura, M.

    1984-01-01

    A structural optimization study of reinforced concrete container for transportation and disposal of the low level radioactive waste generated in Brazilian nuclear power plants. The code requires the structural integrity of these containers when subjected to fall from specified height, avoiding environmental contamination. The structural optimization allows material and transportation cost reduction by container wall thickness reduction. The structural analysis is performed by tridimensional mathematical model using finite element method. (Author) [pt

  11. Behaviour of concrete nuclear containment structures upto ultimate failure with special reference to MAPP-1 containment

    International Nuclear Information System (INIS)

    Appa Rao, T.V.S.R.

    1975-01-01

    Theoretical and experimental methods for investigating the behaviour of concrete secondary containment structures subjected to loads upto their ultimate failure have been discussed in the paper. Need for inelastic nonlinear analysis of containments has been emphasized. Different contitutive models of concrete that can be employed in the nonlinear analysis of concrete structures were briefly reviewed. Based on the experimental results obtained in a 1:12 scale model test conducted at the Structural Engineering Research (Regional) Centre, Madras, behaviour of the MAPP-1 containment to internal pressure loading upto its ultimate failure has been discussed. (author)

  12. Thermodynamic study on the in-vessel corium - Application to the corium/concrete interaction

    International Nuclear Information System (INIS)

    Quaini, Andrea

    2015-01-01

    calculation and experimental results. Heat treatments on two ex-vessel corium samples showed the influence of the concrete composition on the nature of the liquid phases formed at high temperature. The observed microstructures have been interpreted by means of calculation performed with the TAF-ID database. In parallel, a novel experimental setup named ATTILHA based on aerodynamic levitation and laser heating has been conceived and developed to obtain high temperature phase diagram data. This setup has been validated on well-known oxide systems. Furthermore, this technique allowed to observe in-situ, by using an infrared camera, the formation of a miscibility gap in the liquid phase of the O-Fe-Zr system by oxidation of a Fe-Zr sample. The next step of the development will be the nuclearization of the apparatus to investigate U-containing samples. The implementation of a very fast visible camera (5000 Hz) to investigate the thermo-physical properties of in-vessel and ex-vessel corium mixtures is also underway. The synergy between the development of experimental and calculation tools will allow to improve the thermodynamic description of the corium and the severe accident code using thermodynamic input data. (author) [fr

  13. Elements of thought on corium containment strategy in reactor vessel

    International Nuclear Information System (INIS)

    2015-01-01

    As accidents with core fusion are taken into account for the design of third-generation nuclear reactors, this brief document presents the corium containment strategy for a reactor vessel, its limitations, as well as research programs undertaken by the IRSN in this field. The report describes the controlled management of a severe accident, the major objective being to minimise releases in the environment, that which requires to maintain the reactor containment enclosure tightness. Practical actions are briefly indicated. Key points indicating the feasibility of a strategy of containment in vessel are discussed. The impact of reactor power on the robustness of an approach with containment in vessel is also discussed. An overview of technological evolutions and contributions of researches made by the IRSN is finally proposed

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

    International Nuclear Information System (INIS)

    Lafitte, R.; Marchand, J.D.

    1979-01-01

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

  15. Proving Test on the Reliability for Reactor Containment Vessel

    International Nuclear Information System (INIS)

    Takumi, K.; Nonaka, A.

    1988-01-01

    NUPEC (Nuclear Power Engineering Test Center) has started an eight-year project of Proving Test on the Reliability for Reactor Containment Vessel since June 1987. The objective of this project is to confirm the integrity of containment vessels under severe accident conditions. This paper shows the outline of this project. The test Items are (1) Hydrogen mixing and distribution test, (2) Hydrogen burning test, (3) Iodine trapping characteristics test, and (4) Structural behavior test. Based on the test results, computer codes are verified and as the results of analysis and evaluation by the computer codes, containment integrity is to be confirmed

  16. Non destructive Testing (NDT) of concrete containing hematite

    International Nuclear Information System (INIS)

    Mohamad Pauzi Ismail; Noor Azreen Masenwat; Suhairy Sani; Nasharuddin Isa; Mohamad Haniza Mahmud

    2014-01-01

    This paper described the results of Non-destructive ultrasonic and rebound hammer measurements on concrete containing hematite. Local hematite stones were used as aggregates to produce high density concrete for application in X-and gamma shielding. Concrete cube samples (150 mm x 150 mm x 150 mm) containing hematite as coarse aggregates were prepared by changing mix ratio, water to cement ratio (w/c) and types of fine aggregate. All samples were cured in water for 7 days and then tested after 28 days. Density, rebound number(N) and ultrasonic pulse velocity (UPV) of the samples were taken before compressed to failure. The measurement results are explained and discussed. (author)

  17. The instrumentation of the prestressed concrete vessel with hot liner at Seibersdorf Research Centre

    International Nuclear Information System (INIS)

    Zemann, H.

    1975-11-01

    The joint project ''Prestressed Concrete Pressure Vessel with Hot Liner'' at Seibersdorf Research Centre now is in the process of testing the PCPV both in construction and operation from the safety point of view. The physical state of the PCPV (modulus of elasticity, humidity of concrete, creeping, etc.) is brought to stable conditions by ''pre-aging''. In order to control this process of stabilisation, an extensive knowledge of the concrete and an elaborated instrumentation is a necessity. This paper presents a survey about the philosophy and the realisation of the instrumentation of the PCPV and the investigations we performed to interpret the measurements. (author)

  18. Ultimate load analysis of prestressed concrete reactor pressure vessels considering a general material law

    International Nuclear Information System (INIS)

    Schimmelpfennig, K.

    1975-01-01

    A method of analysis is presented, by which progressive fracture processes in axisymmetric prestressed concrete pressure vessels during increasing internal pressure can be evaulated by means of a continuum calculation considering a general material law. Formulations used in the analysis concerning material behaviour are derived on one hand from appropriate results of testing small concrete specimens, and are on the other hand gained by parametric studies in order to solve questions still existing by recalulating fracture tests on concrete bodies with more complex states of stress. Due attention is focussed on investigating the behaviour of construction members subjected to high shear forces (end slabs.). (Auth.)

  19. Application of high strength steel to nuclear reactor containment vessel

    International Nuclear Information System (INIS)

    Susukida, H.; Sato, M.; Takano, G.; Uebayashi, T.; Yoshida, K.

    1976-01-01

    Nuclear reactor containment vessels are becoming larger in size with the increase in the power generating capacity of nuclear power plants. For example, a containment vessel for a PWR power plant with an output of 1,000 MWe becomes an extremely large one if it is made of the conventional JIS SGV 49 (ASTM A 516 Gr. 70) steel plates less than 38 mm in thickness. In order to design the steel containment vessel within the conventional dimensional range, therefore, it is necessary to use a high strength steel having a higher tensile strength than SGV 49 steel, good weldability and a higher fracture toughness and moreover, possessing satisfactory properties without undergoing post-weld heat treatment. The authors conducted a series of verification tests on high strength steel developed by modifying the ASTM A 543 Grade B Class 1 steel with a view to adopting it as a material for the nuclear reactor containment vessels. As the result of evaluation of the test results from various angles, we confirmed that the high strength steel is quite suitable for the manufacture of nuclear reactor containment vessels. (auth.)

  20. Ultimate internal pressure capacity of concrete containment structures

    International Nuclear Information System (INIS)

    Krishnaswamy, C.N.; Namperumal, R.; Al-Dabbagh, A.

    1983-01-01

    Lesson learned from the accident at Three-Mile Island nuclear plant has necessitated the computation of the ultimate internal pressure capacity of containment structures as a licensing requirement in the U.S. In general, a containment structure is designed to be essentially elastic under design accident pressure. However, as the containment pressure builds up beyond the design value due to a more severe postulated accident, the containment response turns nonlinear as it sequentially passes through cracking of concrete, yielding of linear plate, yielding of rebar, and yielding of post-tensioning tendon (if the containment concrete is prestressed). This paper reports on the determination of the ultimate internal pressure capacity and nonlinear behavior of typical reinforced and prestressed concrete BWR containments. The probable modes of failure, the criteria for ultimate pressure capacity, and the most critical sections are described. Simple equations to hand-calculate the ultimate pressure capacity and the nonlinear behavior at membrane sections of the containment shell are presented. A nonlinear finite element analysis performed to determine the nonlinear behavior of the entire shell including nonmembrane sections is briefly discribed. The analysis model consisted of laminated axisymmetric shell finite elements with nonlinear stress-strain properties for each material. Results presented for typical BWR concrete containments include nonlinear response plots of internal pressure versus containment deflection and strains in the liner, rebar, and post-tensioning tendons at the most stressed section in the shell. Leak-tightness of the containment liner and the effect of thermal loads on the ultimate capacity are discussed. (orig.)

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

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.

    1979-01-01

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

  2. Analytical model for shear strength of end slabs of prestressed concrete nuclear reactor vessels

    International Nuclear Information System (INIS)

    Abdulrahman, H.O.; Sozen, M.A.; Schnobrich, W.C.

    1979-04-01

    The results are presented of an investigation of the behavior and strength of flat end slabs of cylindrical prestressed concrete nuclear reactor vessels. The investigation included tests of ten small-scale pressure vessels and development of a nonlinear finite-element model to simulate the deformation response and strength of the end slabs. Because earlier experimental studies had shown that the flexural strength of the end slab could be calculated using intelligible procedures, the emphasis of this investigation was on shear strength

  3. A model to predict moisture conditions in concrete reactor containments

    International Nuclear Information System (INIS)

    Ahs, M.; Nilsson, L.O.; Poyet, S.; L'Hostis, V.

    2015-01-01

    Moisture has an impact in many of the degradation mechanisms that appear in the structures of a nuclear power plant. Moisture conditions in a reactor containment wall have been simulated by using a hygro-thermal model of drying concrete. Methods to estimate the temperature dependency of the sorption isotherms and moisture transport properties is suggested and applied in the model. This temperature dependency is included as there is a temperature gradient present through the containment wall. The hygro-thermal model was applied on a full scale 3D model of a real reactor containment building and the concrete relative humidity has been computed at 4 different moments: 1, 10, 20 and 30 years. The results show that the major part of the concrete is not dried at all even after 30 years of operation. It is also clear that the temperature distribution inside the whole concrete volume is affected by the variable boundary conditions. It was concluded that the suggested hygro-thermal model was appropriate to use as a method to estimate the existing conditions in a PWR reactor containment wall

  4. NOx photocatalytic degradation employing concrete pavement containing titanium dioxide

    NARCIS (Netherlands)

    Ballari, M.M.; Hunger, Martin; Hüsken, Götz; Brouwers, Jos

    2010-01-01

    In the present work the degradation of nitrogen oxides (NOx) by concrete paving stones containing TiO2 to be applied in road construction is studied. A kinetic model is proposed to describe the photocatalytic reaction of NOx (combining the degradation of NO and the appearance and disappearance of

  5. Seismic proving test of PWR reactor containment vessel

    International Nuclear Information System (INIS)

    Akiyama, H.; Yoshikawa, T.; Tokumaru, Y.

    1987-01-01

    The seismic reliability proving tests of nuclear power plant facilities are carried out by Nuclear Power Engineering Test Center (NUPEC), using the large-scale, high-performance vibration of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry (MITI). In 1982, the seismic reliability proving test of PWR containment vessel started using the test component of reduced scale 1/3.7 and the test component proved to have structural soundness against earthquakes. Subsequently, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. Whereupon, the seismic analysis and evaluation on the actual containment vessel were performed by these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed

  6. Concrete as secondary containment for interior wall embedded waste lines

    International Nuclear Information System (INIS)

    Porter, C.L.

    1993-01-01

    Throughout the Department of Energy (DOE) complex are numerous facilities that handle hazardous waste solutions. Secondary containment of tank systems and their ancillary piping is a major concern for existing facilities. The Idaho Division of Environmental Quality was petitioned in 1990 for an Equivalent Device determination regarding secondary containment of waste lines embedded in interior concrete walls. The petition was granted, however it expires in 1996. To address the secondary containment issue, additional studies were undertaken. One study verified the hypothesis that an interior wall pipe leak would follow the path of least resistance through the naturally occurring void found below a rigidly supported pipe and pass into an adjacent room where detection could occur, before any significant deterioration of the concrete takes place. Other tests demonstrated that with acidic waste solutions rebar and cold joints are not an accelerated path to the environment. The results from these latest studies confirm that the subject configuration meets all the requirements of secondary containment

  7. Thermal effects, creep and nonlinear responde of concrete reactor vessels

    International Nuclear Information System (INIS)

    Bazant, Z.P.

    1978-01-01

    A new mathematical model for prediction of pore pressure and moisture transfer in concrete heated well beyond 100 0 C is outlined. The salient features of the model are:(1) the hypothesis taht the pore space available to capillary water grows with increasing temperature as well as increasing pressure in excess of saturation pressure, and (2) the hypothesis that moisture permeability increases by two orders of magnitude when passing 100 0 C. Permaability below 100 0 C is controlled by migration of adsorbed water through gel-pore sized necks on passages through the material; these necks are lost above 100 0 C and viscosity then governs. The driving force of moisture transfer may be considered as the gradient of pore pressure, which is defined as pressure of vapor rather than liquid water if concrete is not saturated. Thermodynamic properties of water may be used to determine sorption isotherms in saturated concrete. The theory is the necessary first step in rationally predicting thermal stresses and deformations, and assessing the danger of explosive spalling. However, analysis of creep and nonlinear triaxial behavior is also needed for this purpose. A brief review of recent achievements in these subjects is also given. (Author)

  8. Device for protecting the containment vessel dome of a nuclear reactor

    International Nuclear Information System (INIS)

    Allain, A.; Filloleau, E.; Mulot, P.

    1976-01-01

    A device is disclosed for protecting the dome of a nuclear reactor containment vessel against the upward displacement of the concrete shield slab of said reactor and the resultant effects of tilting of an equipment unit mounted on the shield slab at the periphery of said slab, wherein said device comprises: (1) means for separating the equipment unit into two sections consisting of an upper section and a lower section, said lower section being rigidly fixed to said shield slab and said means being actuated by the upward displacement of said slab, (2) a system for vertical rectilinear guiding of said upper section within the containment vessel, and (3) rigid mechanical components which provide a coupling between the aforesaid upper and lower sections of the equipment unit and exert on said upper section under the action of the tilting motion of said lower section a thrust which causes the upward displacement of said upper section

  9. 77 FR 69508 - Inservice Inspection of Prestressed Concrete Containment Structures With Grouted Tendons

    Science.gov (United States)

    2012-11-19

    ... Containment Structures With Grouted Tendons AGENCY: Nuclear Regulatory Commission. ACTION: Regulatory guide... (RG) 1.90, ``Inservice Inspection of Prestressed Concrete Containment Structures with Grouted Tendons... appropriate surveillance program for prestressed concrete containment structures with grouted tendons...

  10. To the problem of reinforced concrete reactor vessel design and calculation

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  11. Biaxial behavior of plain concrete of nuclear containment building

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang-Keun E-mail: sklee0806@bcline.com; Song, Young-Chul; Han, Sang-Hoon

    2004-01-01

    To provide biaxial failure behavior characteristics of concrete of a standard Korean nuclear containment building, the concrete specimens with the dimensions of 200 mmx200 mmx60 mm were tested under different biaxial load combinations. The specimens were subjected to biaxial load combinations covering the three regions of compression-compression, compression-tension, nd tension-tension. To avoid a confining effect due to friction in the boundary surface between the concrete specimen and the loading platen, the loading platens with Teflon pads were used. The principal deformations in the specimens were recorded, and the failure modes along with each stress ratio were examined. Based on the strength data, the biaxial ultimate strength envelopes were developed and the biaxial stress-strain responses in three different biaxial loading regions were plotted. The test results indicated hat the concrete strength under equal biaxial compression, f{sub 1}=f{sub 2}, is higher by about 17% on the average than that under the uniaxial compression and the concrete strength under biaxial tension is almost independent of the stress ratio and is similar to that under the uniaxial tension.

  12. Aging of concrete containment structures in nuclear power plants

    International Nuclear Information System (INIS)

    Naus, D.J.; Oland, C.B.; Ellingwood, B.; Mori, Yasuhiro; Arndt, E.G.

    1992-01-01

    Concrete structures play a vital role in the safe operation of all light-water reactor plants in the US Pertinent concrete structures are described in terms of their importance design, considerations, and materials of construction. Degradation factors which can potentially impact the ability of these structures to meet their functional and performance requirements are identified. Current inservice inspection requirements for concrete containments are summarized. A review of the performance history of the concrete components in nuclear power plants is provided. A summary is presented. A summary is presented of the Structural Aging (SAG) Program being conducted at the Oak Ridge National Laboratory for the US Nuclear Regulatory Commission. The SAG Program is addressing the aging management of safety-related concrete structures in nuclear power plants for the purpose of providing improved bases for their continued service. The program consists of a management task and three technical tasks: materials property data base, structural component assessment/repair technologies, and quantitiative methodology for continued service conditions. Objectives and a summary of accomplishments under each of these tasks are presented

  13. Biaxial behavior of plain concrete of nuclear containment building

    International Nuclear Information System (INIS)

    Lee, Sang-Keun; Song, Young-Chul; Han, Sang-Hoon

    2004-01-01

    To provide biaxial failure behavior characteristics of concrete of a standard Korean nuclear containment building, the concrete specimens with the dimensions of 200 mmx200 mmx60 mm were tested under different biaxial load combinations. The specimens were subjected to biaxial load combinations covering the three regions of compression-compression, compression-tension, nd tension-tension. To avoid a confining effect due to friction in the boundary surface between the concrete specimen and the loading platen, the loading platens with Teflon pads were used. The principal deformations in the specimens were recorded, and the failure modes along with each stress ratio were examined. Based on the strength data, the biaxial ultimate strength envelopes were developed and the biaxial stress-strain responses in three different biaxial loading regions were plotted. The test results indicated hat the concrete strength under equal biaxial compression, f 1 =f 2 , is higher by about 17% on the average than that under the uniaxial compression and the concrete strength under biaxial tension is almost independent of the stress ratio and is similar to that under the uniaxial tension

  14. Constitutive models for concrete and finite element analysis of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Smith, P.D.; Anderson, C.A.

    1977-01-01

    Two constitutive models for concrete are discussed. For short-term loads, the orthotropic variable modulus model is described, and for long-term loads a viscoelastic model utilizing a Dirichlet series approximation for the creep compliance function is summarized. The orthotropic variable modulus model is demonstrated in an analysis of a PCRV head with penetrations. The viscoelastic model is illustrated with a simulation of a prestressed concrete cylinder subject to non-uniform temperatures

  15. A system for the thermal insulation of a pre-stressed concrete vessel

    International Nuclear Information System (INIS)

    Aubert, Gilles; Petit, Guy.

    1975-01-01

    This invention concerns the thermal insulation of a pre-stressed concrete vessel for a pressurised water nuclear reactor, this vessel being fitted internally with a leak-proof metal lining. Two rings are placed at the lower and upper parts of the vessel respectively. The upper ring is closed with a cover. These rings differ in diameter, are fitted with a metal insulating and mark the limits of a chamber between the vaporisable fluid and the internal wall of the vessel. This chamber is filled with a fluid in the liquid phase up to the liquid/vapor interface level of the fluid and with a gas above that level, the covering of the rings forming a cold fluid liquid seal. Each ring is supported by the vessel. Leak-proof components take up the radial expansion of the rings [fr

  16. Radiolytic gas production from concrete containing Savannah River Plant waste

    International Nuclear Information System (INIS)

    Bibler, N.E.

    1978-01-01

    To determine the extent of gas production from radiolysis of concrete containing radioactive Savannah River Plant waste, samples of concrete and simulated waste were irradiated by 60 Co gamma rays and 244 Cm alpha particles. Gamma radiolysis simulated radiolysis by beta particles from fission products in the waste. Alpha radiolysis indicated the effect of alpha particles from transuranic isotopes in the waste. With gamma radiolysis, hydrogen was the only significant product; hydrogen reached a steady-state pressure that increased with increasing radiation intensity. Hydrogen was produced faster, and a higher steady-state pressure resulted when an organic set retarder was present. Oxygen that was sealed with the wastes was depleted. Gamma radiolysis also produced nitrous oxide gas when nitrate or nitrite was present in the concrete. With alpha radiolysis, hydrogen and oxygen were produced. Hydrogen did not reach a steady-state pressure at 137 Cs and 90 Sr), hydrogen will reach a steady-state pressure of 8 to 28 psi, and oxygen will be partially consumed. These predictions were confirmed by measurement of gas produced over a short time in a container of concrete and actual SRP waste. The tests with simulated waste also indicated that nitrous oxide may form, but because of the low nitrate or nitrite content of the waste, the maximum pressure of nitrous oxide after 300 years will be 238 Pu and 239 Pu will predominate; the hydrogen and oxygen pressures will increase to >200 psi

  17. Evaluation of calculational and material models for concrete containment structures

    International Nuclear Information System (INIS)

    Dunham, R.S.; Rashid, Y.R.; Yuan, K.A.

    1984-01-01

    A computer code utilizing an appropriate finite element, material and constitutive model has been under development as a part of a comprehensive effort by the Electric Power Research Institute (EPRI) to develop and validate a realistic methodology for the ultimate load analysis of concrete containment structures. A preliminary evaluation of the reinforced and prestressed concrete modeling capabilities recently implemented in the ABAQUS-EPGEN code has been completed. This effort focuses on using a state-of-the-art calculational model to predict the behavior of large-scale reinforced concrete slabs tested under uniaxial and biaxial tension to simulate the wall of a typical concrete containment structure under internal pressure. This paper gives comparisons between calculations and experimental measurements for a uniaxially-loaded specimen. The calculated strains compare well with the measured strains in the reinforcing steel; however, the calculations gave diffused cracking patterns that do not agree with the discrete cracking observed in the experiments. Recommendations for improvement of the calculational models are given. (orig.)

  18. Pressurization of Containment Vessels from Plutonium Oxide Contents

    International Nuclear Information System (INIS)

    Hensel, S.

    2012-01-01

    Transportation and storage of plutonium oxide is typically done using a convenience container to hold the oxide powder which is then placed inside a containment vessel. Intermediate containers which act as uncredited confinement barriers may also be used. The containment vessel is subject to an internal pressure due to several sources including; (1) plutonium oxide provides a heat source which raises the temperature of the gas space, (2) helium generation due to alpha decay of the plutonium, (3) hydrogen generation due to radiolysis of the water which has been adsorbed onto the plutonium oxide, and (4) degradation of plastic bags which may be used to bag out the convenience can from a glove box. The contributions of these sources are evaluated in a reasonably conservative manner.

  19. Prestressed concrete nuclear reactor containment structures. Revision 3

    International Nuclear Information System (INIS)

    Reuter, H.R.; Chang-Lo, P.L.C.; Pfeifer, B.W.; Shah, G.H.; Whitcraft, J.S.

    1975-02-01

    A discussion of the techniques and procedures used for the design of prestressed concrete nuclear reactor containment structures is presented. A physical description of Bechtel designed containment structures is presented. The design bases and load combinations are given for anticipated conditions of service. Reference design documents which include industry codes, specifications, AEC Regulatory Guides, Bechtel Topical Reports and additional criteria as appropriate to containment design are listed. Stepwise procedures typically followed by Bechtel for design of containments is discussed and design examples are presented. A description of currently used analytical methods and the practical application of these methods for containment design is also presented. The principal containment construction materials are identified and codes of practice pertaining to construction procedures are listed. Preoperational structural testing procedures and post-operational surveillance programs are furnished along with results of tests on completed containment structures. (U.S.)

  20. Ageing degradation in the Gentilly-1 concrete containment building

    International Nuclear Information System (INIS)

    Jaffer, S.; Pentecost, S.; Angell, P.; Shenton, B.

    2015-01-01

    Concrete containment buildings (CCBs) are designed for a service life up to 40 years, but nuclear power plant (NPP) refurbishment can extend service life beyond 60 years. Only limited testing can be conducted on an in-service CCB. The Gentilly-1 (G-1) NPP is in a safe, sustainable shutdown state and the G-1 CCB was available for testing to determine age-related degradation that may be relevant to operating CCBs. Visual observation of the G-1 CCB helped to identify various signs of degradation. However, field testing, via concrete removal, was performed to: (i) examine reinforcing bars and concrete to determine their condition and in-situ stresses and (ii) examine condition of post-tensioned (P-T) wires. The concrete was also subjected to laboratory tests to evaluate its physical, mechanical and chemical properties such as compressive strength, carbonation depth, chloride content and presence of internal degradation. The degradation mechanisms that were clearly visible include macro- and micro-cracking, efflorescence, and weathering. The reinforcing bars in the perimeter wall and dome exposed during the program showed no evidence of active corrosion. Corrosion products were observed on the surfaces of most exposed P-T wires in the perimeter wall, but none were present on P-T wires exposed in the dome. Laboratory testing on the concrete cores extracted from the CCB revealed compressive strength in excess of the design requirements, low carbonation depths (< 10 mm) and no appreciable chlorides. Micro-cracking was observed in the samples recovered from the wall and dome. To date, the observed micro-cracking has had no apparent visible affect on the performance of the CCB concrete. (authors)

  1. Concrete containers in radioactive waste management: a review

    International Nuclear Information System (INIS)

    Tavares, Bárbara L.; Tello, Clédola Cássia O. de

    2017-01-01

    Nuclear power is considered a clean energy, because it does not produce the gases responsible for greenhouse effect. However, like all human activities, it is susceptible to waste generation. With increasing demand for energy in Brazil, the use of nuclear power is being expanded, as a result, the implementation of correct treatment and disposal are a necessity, in order to ensure the non-contamination of the public or environment and that exposure doses are lower than limits by legislation. Most of waste produced in Brazil are classified as low and intermediate radiation level; consequently, the national repository will be near surface, in accordance with the legislation. Considering the multi-barrier concept for the repository, the radioactive waste product is the first barrier. To have a qualified radioactive waste product, it should be solid or solidified using an inert material. With the intention of standardize the disposal process, all radioactive waste products will be placed in concrete containers. These containers will be settled in a concrete cell, the final engineered barrier of the repository. The state of the art is the first part of the study of the concrete containers and its specific criteria acceptation. Since the repository’s operational and surveillance period is 60 and 300 years, respectively, tests still need to be fulfilled in order to ensure the stability and resistance of the material. (author)

  2. Concrete containers in radioactive waste management: a review

    Energy Technology Data Exchange (ETDEWEB)

    Tavares, Bárbara L.; Tello, Clédola Cássia O. de, E-mail: barbaralacerdat@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte/MG (Brazil)

    2017-07-01

    Nuclear power is considered a clean energy, because it does not produce the gases responsible for greenhouse effect. However, like all human activities, it is susceptible to waste generation. With increasing demand for energy in Brazil, the use of nuclear power is being expanded, as a result, the implementation of correct treatment and disposal are a necessity, in order to ensure the non-contamination of the public or environment and that exposure doses are lower than limits by legislation. Most of waste produced in Brazil are classified as low and intermediate radiation level; consequently, the national repository will be near surface, in accordance with the legislation. Considering the multi-barrier concept for the repository, the radioactive waste product is the first barrier. To have a qualified radioactive waste product, it should be solid or solidified using an inert material. With the intention of standardize the disposal process, all radioactive waste products will be placed in concrete containers. These containers will be settled in a concrete cell, the final engineered barrier of the repository. The state of the art is the first part of the study of the concrete containers and its specific criteria acceptation. Since the repository’s operational and surveillance period is 60 and 300 years, respectively, tests still need to be fulfilled in order to ensure the stability and resistance of the material. (author)

  3. Design of the prestressed concrete reactor vessel for gas-cooled heating reactors

    International Nuclear Information System (INIS)

    Becker, G.; Notheisen, C.; Steffen, G.

    1987-01-01

    The GHR pebble bed reactor offers a simple, safe and economic possibility of heat generation. An essential component of this concept is the prestressed concrete reactor vessel. A system of cooling pipes welded to the outer surface of the liner is used to transfer the heat from the reactor to the intermediate circuit. The high safety of this vessel concept results from the clear separation of the functions of the individual components and from the design principle of the prestressed conncrete. The prestressed concrete structure is so designed that failure can be reliably ruled out under all operating and accident conditions. Even in the extremely improbable event of failure of all decay heat removal systems when decay heat and accumulated heat are transferred passively by natural convection only, the integrity of the vessel remains intact. For reasons of plant availability the liner and the liner cooling system shall be designed so as to ensure safe elimination of failure over the total operating life. The calculations which were peformed partly on the basis of extremely adverse assumption, also resulted in very low loads. The prestressed concrete vessel is prefabricated to the greatest possible extent. Thus a high quality and optimized fabrication technology can be achieved especially for the liner and the liner cooling system. (orig./HP)

  4. Radioactivity concentration measuring device for radiation waste containing vessel

    International Nuclear Information System (INIS)

    Goto, Tetsuo.

    1994-01-01

    The device of the present invention can precisely and accurately measure a radioactive concentration of radioactive wastes irrespective of the radioactivity concentration distribution. Namely, a Ge detector having a collimator and a plurality of radiation detectors are placed at the outside of the radioactive waste containing vessel in such a way that it can rotate and move vertically relative to the vessel. The plurality of radiation detectors detect radiation coefficient signals at an assumed segment unit of a predetermined length in vertical direction and for every predetermined angle unit in the rotational direction. A weight measuring device determines the weight of the vessel. A computer calculates an average density of radioactivity for the region filled with radioactivity based on the determined net weight and radiation coefficient signals assuming that the volume of the radioactivity is constant. In addition, the computer calculates the amount of radioactivity in the assumed segment by conducting γ -ray absorption compensation calculation for the material in the vessel. Each of the amount of radioactivity is integrated to determine the amount of radioactivity in the vessel. (I.S.)

  5. Plant Life Management of the EC6 Concrete Containment Structure

    Energy Technology Data Exchange (ETDEWEB)

    Abrishami, Homayoun; Ricciuti, Rick; Khan, Azhar [CANDU Energy Inc., Mississauga (Canada)

    2012-03-15

    Aging of reinforced concrete structures due to service conditions, aggressive environments, or accidents may cause their strength, serviceability and durability to decrease over time. Due to the complex nature of safety-related structures in nuclear power plants in comparison to other structures, they possess a number of characteristics that make them comparison to other structures, they possess a number of characteristics that make them unique. These characteristics are: thick concrete cross-sections, heavy reinforcement, often one-side access only, subjected to such ageing stresses as irradiation and elevated temperature, in addition to other typical ageing mechanisms (i. e., exposure to freeze/thaw cycles, aggressive chemicals, etc.) that typically affects other types of non-nuclear structures. For a new plant, the Plant Life Management Program (PLiM) should start in the design process and then continues through construction, plant operation and decommissioning. Hence PLiM must provide not only Ageing Management program (AMP) but also provide requirements on material characteristic and the design criteria as well. The purpose of this paper is to present the Plant Life Management (PLiM) strategy for the concrete containment structure of EC6 (Enhanced CANDU 6) Nuclear Power Plant designed by CANDU Energy Inc. The EC6 is designed for 100-year plant life including a 60-year operating life and an additional 40-year decommissioning period of time. The approach adopted for the PLiM strategy of the concrete containment structure is a preventive one, key areas being: 1) design methodology, 2) material performance and 3) life cycle management and ageing management program. In addition to strength and serviceability, durability is a major consideration during the design phase, service life and up to the completion of decommissioning. Factors affecting durability design include: a) concrete performance, b) structural application, and c) consideration of environmental

  6. Plant Life Management of the EC6 Concrete Containment Structure

    International Nuclear Information System (INIS)

    Abrishami, Homayoun; Ricciuti, Rick; Khan, Azhar

    2012-01-01

    Aging of reinforced concrete structures due to service conditions, aggressive environments, or accidents may cause their strength, serviceability and durability to decrease over time. Due to the complex nature of safety-related structures in nuclear power plants in comparison to other structures, they possess a number of characteristics that make them comparison to other structures, they possess a number of characteristics that make them unique. These characteristics are: thick concrete cross-sections, heavy reinforcement, often one-side access only, subjected to such ageing stresses as irradiation and elevated temperature, in addition to other typical ageing mechanisms (i. e., exposure to freeze/thaw cycles, aggressive chemicals, etc.) that typically affects other types of non-nuclear structures. For a new plant, the Plant Life Management Program (PLiM) should start in the design process and then continues through construction, plant operation and decommissioning. Hence PLiM must provide not only Ageing Management program (AMP) but also provide requirements on material characteristic and the design criteria as well. The purpose of this paper is to present the Plant Life Management (PLiM) strategy for the concrete containment structure of EC6 (Enhanced CANDU 6) Nuclear Power Plant designed by CANDU Energy Inc. The EC6 is designed for 100-year plant life including a 60-year operating life and an additional 40-year decommissioning period of time. The approach adopted for the PLiM strategy of the concrete containment structure is a preventive one, key areas being: 1) design methodology, 2) material performance and 3) life cycle management and ageing management program. In addition to strength and serviceability, durability is a major consideration during the design phase, service life and up to the completion of decommissioning. Factors affecting durability design include: a) concrete performance, b) structural application, and c) consideration of environmental

  7. Experimental investigations concerning the suitability of channel systems for liner leak detection and drainage of a prestressed concrete vessel

    International Nuclear Information System (INIS)

    Nickel, M.; Breitbach, G.; Altes, J.; Escherich, K.H.; Wolters, J.

    1985-02-01

    The iternal surfaces of prestressed concrete pressure vessels are fitted with a steel liner to preserve the gas tightness of the primary circuit. Because of the high quality manufacture and the loading conditions a linear failure can be practically excluded. However, if it is hypothetically assumed, that a leak develops during reactor operation, it may be difficult to determine the position of the leak, because the linear area is very large. For tightness surveillance and for venting channel systems installed in close proximity to the linear are suitable. The suitability of such channels for leak detection, localisation and venting was investigated experimentally. A concrete wall (length 2.5 m, height 2.0 m, thickness 0.5 m) was constructed, covered on one side with a steel liner. Behind the liner two different channel systems have been installed. For the simulation of leaks holes were drilled into the liner. The experimental programm contained the following measurements: determination of gas flow rates into the different leaks, distribution of leakage gas over the array of channels and determination of pressures into the concrete and immediately behind the liner. The experiments have shown, that channel arrays immediately adjacent to the liner are the most suitable systems for localisation and controlled exhaust of leakage gas. The suitability decreases, if the channels are set into the concrete somewhat distant from the liner. (orig.) [de

  8. Design of radial reinforcement for prestressed concrete containments

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shen, E-mail: swang@bechtel.com [Bechtel Power Corporation, 5275 Westview Drive, BP2-2C3, Frederick, MD 21703 (United States); Munshi, Javeed A., E-mail: jamunshi@bechtel.com [Bechtel Power Corporation, 5275 Westview Drive, BP2-2C3, Frederick, MD 21703 (United States)

    2013-02-15

    Highlights: ► A rigorous formulae is proposed to calculate radial stress within prestressed concrete containments. ► The proposed method is validated by finite element analysis in an illustrative practical example. ► A partially prestressed condition is more critical than a fully prestressed condition for radial tension. ► Practical design consideration is provided for detailing of radial reinforcement. -- Abstract: Nuclear containments are critical components for safety of nuclear power plants. Failure can result in catastrophic safety consequences as a result of leakage of radiation. Prestressed concrete containments have been used in large nuclear power plants with significant design internal pressure. These containments are generally reinforced with prestressing tendons in the circumferential (hoop) and meridional (vertical) directions. The curvature effect of the tendons introduces radial tensile stresses in the concrete shell which are generally neglected in the design of such structures. It is assumed that such tensile radial stresses are small as such no radial reinforcement is provided for this purpose. But recent instances of significant delaminations in Crystal River Unit 3 in Florida have elevated the need for reevaluation of the radial tension issue in prestressed containment. Note that currently there are no well accepted industry standards for design and detailing of radial reinforcement. This paper discusses the issue of radial tension in prestressed cylindrical and dome shaped structures and proposes formulae to calculate radial stresses. A practical example is presented to illustrate the use of the proposed method which is then verified by using state of art finite element analysis. This paper also provides some practical design consideration for detailing of radial reinforcement in prestressed containments.

  9. Pre-test analysis results of a PWR steel lined pre-stressed concrete containment model

    International Nuclear Information System (INIS)

    Basha, S.M.; Ghosh, Barnali; Patnaik, R.; Ramanujam, S.; Singh, R.K.; Kushwaha, H.S.; Venkat Raj, V.

    2000-02-01

    Pre-stressed concrete nuclear containment serves as the ultimate barrier against the release of radioactivity to the environment. This ultimate barrier must be checked for its ultimate load carrying capacity. BARC participated in a Round Robin analysis activity which is co-sponsored by Sandia National Laboratory, USA and Nuclear Power Engineering Corporation Japan for the pre-test prediction of a 1:4 size Pre-stressed Concrete Containment Vessel. In house finite element code ULCA was used to make the test predictions of displacements and strains at the standard output locations. The present report focuses on the important landmarks of the pre-test results, in sequential terms of first crack appearance, loss of pre-stress, first through thickness crack, rebar and liner yielding and finally liner tearing at the ultimate load. Global and local failure modes of the containment have been obtained from the analysis. Finally sensitivity of the numerical results with respect to different types of liners and different constitutive models in terms of bond strength between concrete and steel and tension-stiffening parameters are examined. The report highlights the important features which could be observed during the test and guidelines are given for improving the prediction in the post test computation after the test data is available. (author)

  10. Seismic damage assessment of reinforced concrete containment structures

    International Nuclear Information System (INIS)

    Cho, HoHyun; Koh, Hyun-Moo; Hyun, Chang-Hun; Kim, Moon-Soo; Shin, Hyun Mock

    2003-01-01

    This paper presents a procedure for assessing seismic damage of concrete containment structures using the nonlinear time-history numerical analysis. For this purpose, two kinds of damage index are introduced at finite element and structural levels. Nonlinear finite element analysis for the containment structure applies PSC shell elements using a layered approach leading to damage indices at finite element and structural levels, which are then used to assess the seismic damage of the containment structure. As an example of such seismic damage assessment, seismic damages of the containment structure of Wolsong I nuclear power plant in Korea are evaluated against 30 artificial earthquakes generated with a wide range of PGA according to US NRC regulatory guide 1.60. Structural responses and corresponding damage index according to the level of PGA and nonlinearity are investigated. It is also shown that the containment structure behaves elastically for earthquakes corresponding to or lower than DBE. (author)

  11. Monitoring of prestressed concrete pressure vessels. II. performance of selected concrete embedment strain meters under normal and extreme environmental conditions

    International Nuclear Information System (INIS)

    Naus, D.J.; Hurtt, C.C.

    1978-10-01

    Unique types of instrumentation are used in prestressed concrete pressure vessels (PCPVs) to measure strains, stresses, deflections, prestressing forces, moisture content, temperatures, and possibly cracking. Their primary purpose is to monitor these complex structures throughout their 20- to 30-year operating lifetime in order to provide continuing assurance of their reliability and safety. Numerous concrete embedment instrumentation systems are available commercially. Since this instrumentation is important in providing continuing assurance of satisfactory performance of PCPVs, the information provided must be reliable. Therefore, laboratory studies were conducted to evaluate the reliability of these commercially available instrumentation systems. This report, the second in a series related to instrumentation embedded in concrete, presents performance-reliability data for 13 types of selected concrete embedment strain meters which were subjected to a variety of loading environments, including unloaded, thermally loaded, simulated PCPV, and extreme environments. Although only a limited number of meters of each type were tested in any one test series, the composite results of the investigation indicate that the majority of these meters would not be able to provide reliable data throughout the 20- to 30-year anticipated operating life of a PCPV. Specific conclusions drawn from the study are: (1) Improved corrosion-resistant materials and sealing techniques should be developed for meters that are to be used in PCPV environments. (2) There is a need for the development of meters that are capable of surviving in concretes where temperatures in excess of 66 0 C are present for extended periods of time. (3) Research should be conducted on other measurement techniques, such as inductance, capacitance, and fluidics

  12. Structure simulation of a pre-stressed concrete containment model

    International Nuclear Information System (INIS)

    Grebner, H.; Sievers, J.

    2004-01-01

    An axisymmetric Finite-Element-Model of the 1:4 pre-stressed containment model tested at SANDIA was developed. The model is loaded by the pre-stressing of the tendons and by increasing internal pressure (up to 1.3 MPa). The analyses results in terms of displacements and strains in the liner, the rebars, the tendons and the concrete of the cylindrical part agree well with measured data up to about 0.6 MPa internal pressure (i.e. 1.5 times design pressure). First circumferential micro-cracks in the concrete are found at about 0.75 MPa. With increasing pressure micro-cracks are present through the whole wall. Above about 0.9 MPa the formation of micro-cracks in radial and meridional direction is calculated. At the maximum load (1.3 MPa) almost all concrete parts of the model have micro-cracks which may cause leaks. Nevertheless the failure of the containment model is not expected for loads up to 1.3 MPa without consideration of geometric inhomogeneities due to penetrations in the wall. Although the calculated strains in liner, rebars and tendons show some plastification, the maximum values are below the critical ones. The safety margin against failure is smallest in some hoop tendons. At present parametric studies are performed to investigate the differences between calculations and measured data. Furthermore three-dimensional models are developed for a better simulation of the meridional tendons in the dome region. (orig.)

  13. Properties of slag concrete for low-level waste containment

    International Nuclear Information System (INIS)

    Langton, C.A.; Wong, P.B.

    1991-01-01

    Ground granulated blast furnace slag was incorporated in the concrete mix used for construction of low-level radioactive waste disposal vaults. The vaults were constructed as six 100 x 100 x 25 ft cells with each cell sharing internal walls with the two adjacent cells. The vaults were designed to contain a low-level radioactive wasteform called saltstone and to isolate the saltstone from the environment until the landfill is closed. Closure involves backfilling with native soil, installation of clay cap, and run-off control. The design criteria for the slag-substituted concrete included compressive strength, 4000 psi after 28 days; slump, 6 inch; permeability, less than 10 -7 cm/sec; and effective nitrate, chromium and technetium diffusivities of 10 -8 , 10 -12 and 10 -12 cm 2 /sec, respectively. The reducing capacity of the slag resulted in chemically reducing Cr +6 to Cr +3 and Tc +7 to Tc +4 and subsequent precipitation of the respective hydroxides in the alkaline pore solution. Consequently, the concrete vault enhances containment of otherwise mobile waste ions and contributes to the overall protection of the groundwater at the disposal site

  14. Reliability-based inspection of prestressed concrete containment structures

    International Nuclear Information System (INIS)

    Pandey, M.D.

    1996-03-01

    A study was undertaken to develop a reliability-based approach to the planning of inspection programs for prestressed concrete containment structures. The main function of the prestressing system is to ensure the leak integrity of the containment by maintaining a compressive state of stress under the tensile forces which arise in a hypothesized loss of coolant accident. Prestressing force losses (due to creep and shrinkage, stress relaxation or tendon corrosion) can lead to tensile stresses under accident pressure, resulting in loss of containment leak integrity due to concrete cracking and tensile yielding of the non-prestressed reinforcement. Therefore, the evaluation of prestressing inspection programs was based on their effectiveness in maintaining an acceptable reliability level with respect to a limit state representing yeilding of non-prestressed reinforcement. An annual target reliability of 10 -4 was used for this limit state. As specified in CSA-N287.7, the evaluation of prestressing systems for containment structures is based on the results of lift-off tests to determine the prestressing force. For unbonded systems the tests are carried out on a randomly selected sample from each tendon group in the structure. For bonded systems, the test is carried out on an unbonded test beam that matches the section geometry and material properties of the containment structure. It was found that flexural testing is useful in updating the probability of concrete cracking under accident pressure. For unbonded systems, the analysis indicated that the sample size recommended by the CSA Standard (4% of the tendon population) is adequate. The CSA recommendation for a five year inspection interval is conservative unless severe degradation of the prestressing system, characterized by a high prestressing loss rate (>3%) and a large coefficient of variation of the measured prestressing force (>15%), is observed

  15. Aging management of light water reactor concrete containments

    International Nuclear Information System (INIS)

    Shah, V.N.; Hookhman, C.J.

    1994-01-01

    This paper evaluates aging of light water reactor concrete containments and identifies three degradation mechanisms that have potential to cause widespread aging damage after years of satisfactory experience: alkali-silica reaction, corrosion of reinforcing steel, and sulfate attack. The evaluation is based on a comprehensive review of the relevant technical literature. Low-alkali cement and slow-reacting aggregates selected according to ASTM requirements cause deleterious alkali-silica reactions. Low concentrations of chloride ions can initiate corrosion of the reinforcing steel if the hydroxyl ions are sufficiently reduced by carbonation, leaching, or magnesium sulfate attack. Magnesium sulfate attack on concrete can cause loss of strength and cementitious properties after long exposure. Techniques to detect and mitigate these long-term aging effects are discussed

  16. Analysis of chloride diffusivity in concrete containing red mud

    Directory of Open Access Journals (Sweden)

    D.V. Ribeiro

    Full Text Available Red mud is a solid waste produced in the alumina production process and, due to its high pH, is classified as hazardous. Its incorporation in concrete mixtures, acting as filler due to the particles fineness, might be an interesting reuse alternative. The focus of this paper is to study the chloride diffusivity of concrete mixtures containing red-mud. The concentration of chlorides was monitored by measuring the conductivity of the anolyte, which was distilled water initially. In addition, the estimation of the chloride ions diffusion coefficients in steady and non-steady conditions, Ds and Dns, was obtained from the ''time-lag'' and ''equivalent time'' between diffusion and migration experiments. Due to superfine particle-size distribution and the "filler" effect, the red mud addition seems to assure lower chloride diffusivity.

  17. Method of detecting leakage in nuclear reactor containment vessel

    International Nuclear Information System (INIS)

    Koba, Akitoshi; Goto, Seiichiro.

    1974-01-01

    Object: To permit accurate and prompt detection of leakage of a radioactive substance. Structure: The rate of change of such factors as radiation dose, temperature and pressure in the containment vessel, and each detected rate of change is compared with a reference value. The running cycle of the condensed drain exhausting pump in a drain collecting tank within a predetermined period is detected, and it is also compared with a reference value. These comparisons determine the absence or presence of leakage. (Kamimura, M.)

  18. Prestressed-concrete pressure vessels and their applicability to advanced-energy-system concepts

    International Nuclear Information System (INIS)

    Naus, D.J.

    1983-01-01

    Prestressed concrete pressure vessels (PCPVs) are, in essence, spaced steel structures since their strength is derived from a multitude of steel elements made up of deformed reinforcing bars and prestressing tendons which are present in sufficient quantities to carry tension loads imposed on the vessel. Other major components of a PCPV include the concrete, liner and cooling system, and insulation. PCPVs exhibit a number of advantages which make them ideally suited for application to advanced energy concepts: fabricability in virtually any size and shape using available technology, improved safety, reduced capital costs, and a history of proven performance. PCPVs have many applications to both nuclear- and non-nuclear-based energy systems concepts. Several of these concepts will be discussed as well as the research and development activities conducted at ORNL in support of PCPV development

  19. Reinforced concrete containment structures in high seismic zones

    International Nuclear Information System (INIS)

    Aziz, T.S.

    1977-01-01

    A new structural concept for reinforced concrete containment structures at sites where earthquake ground motions in terms of the Safe Shutdown Earthquake (SSE) exceeds 0.3 g is presented. The structural concept is based on: (1) an inner steel-lined concrete shell which houses the reactor and provides shielding and containment in the event of loss of coolant accident; (2) an outer annular concrete shell structure which houses auxilary reactor equipment and safeguards systems. These shell structures are supported on a common foundation mat which is embeded in the subgrade. Under stipulated earthquake conditions the two shell structures interact to resist lateral inertia forces. Thus the annular structure which is not a pressure boundary acts as a lateral support for the inner containment shell. The concept is practical, economically feasible and new to practice. An integrated configuration which includes the interior shell, the annular structure and the subgrade is analyzed for several static and dynamic loading conditions. The analysis is done using a finite difference solution scheme for the static loading conditions. A semi-analytical three-dimensional finite element scheme combined with a Fast Fourier Transform (FFT) algorithm is used for the dynamic loading conditions. The effects of cracking of the containment structure due to pressurization in conjunction with earthquake loading are discussed. Analytical results for both the finite difference and the finite element schemes are presented and the sensitivity of the results to changes in the input parameters is studied. General recommendations are given for plant configurations where high seismic loading is a major design consideration

  20. Design and construction of the prestressed concrete boiler closures for the Hartlepool and Heysham pressure vessels

    International Nuclear Information System (INIS)

    Crowder, R.; Howells, R.M.; Paton, A.A.

    1976-01-01

    At a relatively late stage in the station design, the boiler closures for the reactor vessels at Hartlepool and Heysham were changed from steel to prestressed concrete. This paper sets out the criteria which were finally evolved for the new style of closure and describes the way in which the prestressed concrete closure's parts were designed to satisfy these criteria. With both the civil and mechanical components of the closure having their own specific requirements, close co-operation was necessary between these disciplines to ensure that a compatible and practical closure design resulted. This close interrelationship has been carried through into the construction stage and a special concreting and prestressing factory has been built adjacent to the works of the mechanical component fabricator. This enabled an optimum manufacturing cycle to be followed and the important aspects of this are described in the paper. (author)

  1. Sensitivity analysis of numerical model of prestressed concrete containment

    Energy Technology Data Exchange (ETDEWEB)

    Bílý, Petr, E-mail: petr.bily@fsv.cvut.cz; Kohoutková, Alena, E-mail: akohout@fsv.cvut.cz

    2015-12-15

    Graphical abstract: - Highlights: • FEM model of prestressed concrete containment with steel liner was created. • Sensitivity analysis of changes in geometry and loads was conducted. • Steel liner and temperature effects are the most important factors. • Creep and shrinkage parameters are essential for the long time analysis. • Prestressing schedule is a key factor in the early stages. - Abstract: Safety is always the main consideration in the design of containment of nuclear power plant. However, efficiency of the design process should be also taken into consideration. Despite the advances in computational abilities in recent years, simplified analyses may be found useful for preliminary scoping or trade studies. In the paper, a study on sensitivity of finite element model of prestressed concrete containment to changes in geometry, loads and other factors is presented. Importance of steel liner, reinforcement, prestressing process, temperature changes, nonlinearity of materials as well as density of finite elements mesh is assessed in the main stages of life cycle of the containment. Although the modeling adjustments have not produced any significant changes in computation time, it was found that in some cases simplified modeling process can lead to significant reduction of work time without degradation of the results.

  2. Dynamic testing of MFTF containment-vessel structural system

    International Nuclear Information System (INIS)

    Weaver, H.J.; McCallen, D.B.; Eli, M.W.

    1982-01-01

    Dynamic (modal) testing was performed on the Magnetic Fusion Test Facility (MFTF) containment vessel. The seismic design of this vessel was heavily dependent upon the value of structural damping used in the analysis. Typically for welded steel vessels, a value of 2 to 3% of critical is used. However, due to the large mass of the vessel and magnet supported inside, we felt that the interaction between the structure and its foundation would be enhanced. This would result in a larger value of damping because vibrational energy in the structure would be transferred through the foundation into the surrounding soil. The dynamic test performed on this structure (with the magnet in place) confirmed this later theory and resulted in damping values of approximately 4 to 5% for the whole body modes. This report presents a brief description of dynamic testing emphasizing the specific test procedure used on the MFTF-A system. It also presents an interpretation of the damping mechanisms observed (material and geometric) based upon the spatial characteristics of the modal parameters

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

    International Nuclear Information System (INIS)

    Bazant, Z.P.

    1984-01-01

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

  4. Preliminary results of steel containment vessel model test

    International Nuclear Information System (INIS)

    Matsumoto, T.; Komine, K.; Arai, S.

    1997-01-01

    A high pressure test of a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of a steel containment vessel (SCV), representing an improved boiling water reactor (BWR) Mark II containment, was conducted on December 11-12, 1996 at Sandia National Laboratories. This paper describes the preliminary results of the high pressure test. In addition, the preliminary post-test measurement data and the preliminary comparison of test data with pretest analysis predictions are also presented

  5. Probability based load factors for design of concrete containment structures

    International Nuclear Information System (INIS)

    Hwang, H.; Kagami, S.; Reich, M.; Ellingwood, B.; Shinozuka, M.

    1985-01-01

    This paper describes a procedure for developing probability-based load combinations for the design of concrete containments. The proposed criteria are in a load and resistance factor design (LRFD) format. The load factors and resistance factors are derived for use in limit states design and are based on a target limit state probability. In this paper, the load factors for accident pressure and safe shutdown earthquake are derived for three target limit state probabilities. Other load factors are recommended on the basis of prior experience with probability-based design criteria for ordinary building construction. 6 refs

  6. Seismic reliability assessment methodology for CANDU concrete containment structures

    International Nuclear Information System (INIS)

    Stephens, M.J.; Nessim, M.A.; Hong, H.P.

    1995-05-01

    A study was undertaken to develop a reliability-based methodology for the assessment of existing CANDU concrete containment structures with respect to seismic loading. The focus of the study was on defining appropriate specified values and partial safety factors for earthquake loading and resistance parameters. Key issues addressed in the work were the identification of an approach to select design earthquake spectra that satisfy consistent safety levels, and the use of structure-specific data in the evaluation of structural resistance. (author). 23 refs., 9 tabs., 15 figs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-01-15

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

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

    International Nuclear Information System (INIS)

    Lafitte, R.; Marchand, J.D.

    1981-01-01

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

  9. Design and analysis of reactor containment of steel-concrete composite laminated shell

    International Nuclear Information System (INIS)

    Ichikawa, K.; Isobata, O.; Kawamata, S.

    1977-01-01

    A new scheme of containment consisting of steel-concrete laminated shell is being developed. In the main part of a cylindrical vessel, the shell consists of two layers of thin steel plates located at the inner and outer surfaces, and a layer of concrete core into which both the steel plates are anchored. Because of the compressive and shearing resistance of the concrete core, the layers behave as a composite solid shell. Membrane forces are shared by steel plates and partly by concrete core. Bending moment is effectively resisted by the section with extreme layers of steel. Therefore, both surfaces can be designed as extremely thin plates: the inner plate, which is a load carrying members as well as a liner, can be welded without the laborious process of stress-relieving, and various jointing methods can be applied to the outer plate which is free from the need for leak tightness. The capability of the composite layers of behaving as a unified solid shell section depends largely on the shearing rigidity of the concrete core. However, as its resisting capacity to transverse shearing force is comparatively low, a device for reducing the shearing stress at the junction to the base mat is needed. In the new scheme, this part of the cylindrical shell is divided into multiple layers of the same kind of composite shell. This device makes the stiffness of the bottom of the cylindrical shell to lateral movement minimum while maintaining the proper resistance to membrane forces. The analysis shows that the transverse shearing stress can be reduced to less than 1√n of the ordinary case by dividing the thickness of the shell into n layers which are able to slip against each other at the contact surface. In order to validate the feasibility and safety of this new design, the results of analysis on the basis of up-to-date design loads are presented

  10. Structural Integrity Evaluation of Containment Vessel under Severe Accident for PGSFR

    International Nuclear Information System (INIS)

    Lee, Seong-Hyeon; Koo, Gyeong-Hoi; Kim, Sung-Kyun

    2016-01-01

    This paper provides structural integrity evaluation results of CV of the PGSFR(Prototype Gen-IV Sodium Fast Reactor) under severe accident through transient analysis. The evaluation was carried out according to ASME B and PV Code Sec. III-Subsection NH rule. Structural integrity of CV was evaluated through transient analysis of structure in case of severe accident. Stress evaluation results for selected evaluation sections satisfy design criteria of ASME B and PV Code Sec. III Subsection NH. The transient load condition of normal operation will considered in the future work. The purpose of RVCS is to maintain the integrity of concrete structure during normal power operation. Therefore RVCS should be designed to keep the temperature of concrete surface under design limit and to minimize heat loss through CV(Containment Vessel). And in case of severe accident, the integrity of reactor structure and concrete structure should be maintained. Therefore RVCS should be designed to satisfy ASME Level D service limits. When RVCS works with breakdown of DHRS after severe accident, the temperature change of inner and outer surface of CV over time can affect structural integrity of CV. To verify the structural integrity, it is necessary to perform transient analysis of CV structure under changing temperature over time

  11. Concrete containments in Swedish nuclear power plants. A review of construction and material

    International Nuclear Information System (INIS)

    Roth, Thomas; Silfwerbrand, Johan; Sundquist, Haakan

    2002-12-01

    The purpose of project is the long-term accumulation of knowledge related to the status of existing structures in order to facilitate answers to questions that may arise in the future. We have visited all the power stations in Sweden and in conjunction with these visits we have gone through all the relevant documents relating to the constructional concrete. An assessment of the structural integrity, related to the question of cracking and hence seepage, has been conducted. Currently, the work has only been done on a random sampling basis as in many cases important information is still missing. Generally, it can be said that the relevant constructions are, from a structural integrity point-of-view, correctly designed and detailed and have very high safety margins for the load cases which constitute the functional demands placed upon the installation. Each containment structure (vessel) appears to have been designed and built using the best available knowledge at the time of construction. It may be of interest to note that when these structures were built there was a very high level of competence and experience of how to design, detail, and construct large concrete structures. The cement used for the majority of these large concrete structures forming nuclear power stations, namely a slowly hardening cement (LH cement), had very good properties, perhaps even better than those available today. Later structures were built with other cements and concrete mixes, although this has been partly compensated for by a choice of a higher nominal quality. The environment is favourable regarding potential degradation of the concrete, the reinforcement steel and the steel liner. Questions remain regarding the uncertainties of the methods used for continuous inspection of the cement injected prestressing steel. This is even the case for possibly insufficient injection around grouting mounting parts for manholes and other openings. Assessment of prestressing losses may also require

  12. Air and gas cleaning methods for reactor containment vessels

    Energy Technology Data Exchange (ETDEWEB)

    Silverman, L.

    1963-11-15

    In this paper, a survey is made of the existing and some proposed new methods for the control and purification of air and gases which might be released from a reactor contained or confined for protection of the health and safety of the public from potential accidents. The difference between confinement and containment concepts must be considered. The problems involved and the need for decontamination, site selection, exclusion area, population density, distance, etc., have been discussed elsewhere. We propose to discuss here the safety measures necessary to control the release of radioactive materials to the environment. This requires special systems which must function effectively to minimize loss of fission products such as halogens and particulates. These can penetrate the confinement filters or the containment vessel to a limited extent even after cleaning.

  13. DURABILITY OF GREEN CONCRETE WITH TERNARY CEMENTITIOUS SYSTEM CONTAINING RECYCLED AGGREGATE CONCRETE AND TIRE RUBBER WASTES

    Directory of Open Access Journals (Sweden)

    MAJID MATOUQ ASSAS

    2016-06-01

    Full Text Available All over the world billions of tires are being discarded and buried representing a serious ecological threat. Up to now a small part is recycled and millions of tires are just stockpiled, landfilled or buried. This paper presents results about the properties and the durability of green concrete contains recycled concrete as a coarse aggregate with partial replacement of sand by tire rubber wastes for pavement use. Ternary cementious system, Silica fume, Fly ash and Cement Kiln Dust are used as partial replacement of cement by weight. Each one replaced 10% of cement weight to give a total replacement of 30%. The durability performance was assessed by means of water absorption, chloride ion permeability at 28 and 90 days, and resistance to sulphuric acid attack at 1, 7, 14 and 28 days. Also to the compression behaviors for the tested specimens at 7, 14, 28 and 90 days were detected. The results show the existence of ternary cementitious system, silica fly ash and Cement Kiln Dust minimizes the strength loss associated to the use of rubber waste. In this way, up to 10% rubber content and 30% ternary cementious system an adequate strength class value (30 MPa, as required for a wide range of common structural uses, can be reached both through natural aggregate concrete and recycled aggregate concrete. Results also show that, it is possible to use rubber waste up to 15% and still maintain a high resistance to acid attack. The mixes with 10%silica fume, 10% fly ash and 10% Cement Kiln Dust show a higher resistance to sulphuric acid attack than the reference mix independently of the rubber waste content. The mixes with rubber waste and ternary cementious system was a lower resistance to sulphuric acid attack than the reference mix.

  14. A comparison of elastic-plastic and variable modulus-cracking constitutive models for prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Anderson, C.A.; Smith, P.D.

    1979-01-01

    Numerical prediction of the behavior of prestressed concrete reactor vessels (PCRVs) under static, dynamic and long term loadings is complicated by the currently ill-defined behavior of concrete under stress and the three-dimensional nature of PCRVs. Which constitutive model most closely approximates the behavior of concrete in PCRVs under load has not yet been decided. Many equations for accurately modeling the three-dimensional behavior of PCRVs tax the capability of a most up-to-date computing system. The main purpose of this paper is to compare the characteristics of two constitutive models which have been proposed for concrete, variable modulus cracking model and elastic-plastic model. Moreover, the behavior of typical concrete structures was compared, the materials of which obey these constitutive laws. The response to internal pressure of PCRV structure, the constitutive models for concrete, the test problems using a thick-walled concrete ring and a rectangular concrete plate, and the analysis of an axisymmetric concrete pressure vessel PV-26 using the variable modulus cracking model of the ADINA code are explained. The variable modulus cracking model can predict the behavior of reinforced concrete structures well into the range of nonlinear behavior. (Kako, I.)

  15. Environmental performance and mechanical analysis of concrete containing recycled asphalt pavement (RAP) and waste precast concrete as aggregate.

    Science.gov (United States)

    Erdem, Savaş; Blankson, Marva Angela

    2014-01-15

    The overall objective of this research project was to investigate the feasibility of incorporating 100% recycled aggregates, either waste precast concrete or waste asphalt planning, as replacements for virgin aggregates in structural concrete and to determine the mechanical and environmental performance of concrete containing these aggregates. Four different types of concrete mixtures were designed with the same total water cement ratio (w/c=0.74) either by using natural aggregate as reference or by totally replacing the natural aggregate with recycled material. Ground granulated blast furnace slag (GGBS) was used as a mineral addition (35%) in all mixtures. The test results showed that it is possible to obtain satisfactory performance for strength characteristics of concrete containing recycled aggregates, if these aggregates are sourced from old precast concrete. However, from the perspective of the mechanical properties, the test results indicated that concrete with RAP aggregate cannot be used for structural applications. In terms of leaching, the results also showed that the environmental behaviour of the recycled aggregate concrete is similar to that of the natural aggregate concrete. Copyright © 2013 Elsevier B.V. All rights reserved.

  16. Ultimate load design and testing of a cylindrical prestressed concrete vessel

    International Nuclear Information System (INIS)

    Stefanou, G.D.

    1982-01-01

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

  17. The Use of Prestressed Concrete Vessels in the French Power Reactor Programme

    International Nuclear Information System (INIS)

    Conte, F.; Dambrine, C.; Gaussot, D.

    1963-01-01

    This paper deals with the use of pre-stressed concrete for the G2 and G3 reactors at Marcoule and for the EDF3 reactor now under construction at Chinon. The first two reactors have been operating at power since 1959 and 1960 respectively. Messrs. Conte and Dambrine discuss the problems that arose during construction of the vessels for G2 and G3 and also deal with the experience gained in operation - experience which suggests that they are extremely safe- Work on the EDF3 vessel, begun at Chinon in the second half of 1961, is still under way and should be finished towards the end of 1963. Mr. Gaussot discusses the reasons for choosing this type of vessel, the results of calculations and mock-up tests, and the problems presented by the construction itself. A number of studies have been devoted to the future prospects of prestressed concrete structures for reactors. It would seem that working pressures could be increased, if desired, and, in any case, that dimensions could be considerably enlarged, thus offering the chance of integral-type solutions. (author) [fr

  18. Integrated leak rate test results of JOYO reactor containment vessel

    International Nuclear Information System (INIS)

    Tamura, M.; Endo, J.

    1982-02-01

    Integrated leak rate tests of JOYO after the reactor coolant system had been filled with sodium have been performed two times since 1978 (February 1978 and December 1979). The tests were conducted with the in-containment sodium systems, primary argon cover gas system and air conditioning systems operating. Both the absolute pressure method and the reference chamber method were employed during the test. The results of both tests confirmed the functioning of the containment vessel, and leak rate limits were satisfied. In Addition, the adequancy of the test instrumentation system and the test method was demonstrated. Finally the plant conditions required to maintain reasonable accuracy for the leak rate testing of LMFBR were established. In this paper, the test conditions and the test results are described. (author)

  19. Exposure rates from concrete covered cylindrical units containing radioactive waste

    International Nuclear Information System (INIS)

    Hedemann Jensen, P.

    1983-03-01

    Exposure rates from cylindrical waste units containing the nuclides 60 Co, 134 Cs and 137 Cs homogeneously mixed in a solidification product have been calculated. Analyses have been made for single drums and for two disposal geometries, one with the units placed below ground near the surface in a circular geometry, and one with the units placed on the ground in a pile behind a concrete wall. Due to self-shielding of the units, the exposure rate from the two geometries will be a factor of only 10 - 20 higher than from a single unit, even without soil or wall shielding. With one meter of soil above the circular pile below ground, a reduction factor of 5.10 3 to 5.10 4 can be achieved, depending on the nuclide considered. Placing a one-meter concrete wall in front of the drum pile on the ground gives rise to a reduction factor in the range of 5.10 5 to 2.10 7 . (author)

  20. Bonded or Unbonded Technologies for Nuclear Reactor Prestressed Concrete Containments

    International Nuclear Information System (INIS)

    Abrishami, Homayoun; Tcherner, Julia; Barre, Francis; Borgerhoff, Michael; Bumann, Urs; Calonius, Kim; Courtois, Alexis; Debattista, Jean-Marc; Gallitre, Etienne; Isard, Cedric; Elison, Oscar; Graves, Herman; Sircar, Madhumita; Huerta, Alejandro; White, Andrew; ); Jackson, Paul; Kjellin, Daniel; Lillhoek, Sofia; Louhivirta, Jari; Myllymaeki, Jukka; Vaelikangas, Pekka; Martin, Jose; Nakano, Makio; Puttonen, Jari; Rambach, Jean-Mathieu; Tarallo, Francois; Smith, Leslie; Stepan, Jan; Touret, Jean-Pierre; Varpasuo, Pentti

    2015-01-01

    OECD/NEA/CSNI Working Group on Integrity and Ageing of Components and Structures (WGIAGE) has the main mission to advance the current understanding of those aspects relevant to ensuring the integrity of structures, systems and components under design and beyond design loads, to provide guidance in choosing the optimal ways of dealing with challenges to the integrity of operating as well as new nuclear power plants, and to make use of an integrated approach to design, safety and plant life management. The work related to the risks of the loss of pre-stressing force in concrete structures has been in high priority during the activities of the concrete sub-group of WGIAGE. Therefore, the CAPS of WGIAGE: Study on post-tensioning methodologies in containments, was approved by CSNI in June 2009. In this study the two post-tensioning methodologies: bonded and un-bonded methods and their technological features are analysed. In the bonded technology, the tendon cannot slide in its duct due to the cement grouting which is injected after tensioning. In the un-bonded technology, the tendon can slide inside its duct, the corrosion protection is given by grease, wax or dry air. A key point concerning the assessment of durability and safety of prestressed concrete containments is the technology chosen for tendon protection: bonded with cement grout or un-bonded and protected by grease or soft products. The mechanical behaviour of the containment is directly influenced by the adherence of the tendons to the concrete, locally and under high stresses in case of severe accident. The bonded or un-bonded tendons of post-tensioned concrete containment of the Nuclear Power Plants have the major role of containment (balance of the pressure effect during design basis and beyond design accident). Many difficulties around the design, the construction and the in service inspection are related to the tendons. The main goal of the CAPS work was to clarify the consequences and necessary

  1. Physical and mechanical properties of self-compacting concrete containing superplasticizer and metakaolin

    Science.gov (United States)

    Shahidan, Shahiron; Tayeh, Bassam A.; Jamaludin, A. A.; Bahari, N. A. A. S.; Mohd, S. S.; Zuki Ali, N.; Khalid, F. S.

    2017-11-01

    The development of concrete technology shows a variety of admixtures in concrete to produce special concrete. This includes the production of self-compacting concrete which is able to fill up all spaces, take formwork shapes and pass through congested reinforcement bars without vibrating or needing any external energy. In this study, the main objective is to compare the physical and mechanical properties of self-compacting concrete containing metakaolin with normal concrete. Four types of samples were produced to study the effect of metakaolin towards the physical and mechanical properties of self-compacting concrete where 0%, 5%, 10% and 15% of metakaolin were used as cement replacement. The physical properties were investigated using slump test for normal concrete and slump flow test for self-compacting concrete. The mechanical properties were tested for compressive strength and tensile strength. The findings of this study show that the inclusion of metakaolin as cement replacement can increase both compressive and tensile strength compared to normal concrete. The highest compressive strength was found in self-compacting concrete with 15% metakaolin replacement at 53.3 MPa while self-compacting concrete with 10% metakaolin replacement showed the highest tensile strength at 3.6 MPa. On top of that, the finishing or concrete surface of both cube and cylinder samples made of self-compacting concrete produced a smooth surface with the appearance of less honeycombs compared to normal concrete.

  2. The design of bonded reinforcement for thermal stresses in prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Kotulla, B.; Hansson, V.

    1977-01-01

    This paper deals with examples of thermal loadings where instationary growth of tensile zones and redistribution of stresses by cracking are of importance. Temperatures produce, in addition to prestressing and internal pressure, the most important stresses in a prestressed concrete reactor pressure vessel. Characteristic of thermal stresses is that they are influenced to a large extent by creep of concrete and that they influence stress redistributions by temperature dependent creep data. Computations show that during the first instationary heating process of the vessel stresses are reduced by creep effects to about fifty percent of the values of the stationary elastic case at the hot face. With a following cooling, creep effects are generally much less, so this case may produce tensile stresses on the internal face of the wall which lead to cracking of the concrete. Tensile stresses first occur due to the instationary growth of the temperature field in a narrow zone near the liner. If outside this zone compressive stresses exist due to prestressing then crack spreading is limited and restraint by the parts of the wall under compression causes crack distribution even without reinforcement in this zone. Growth of cracks with the instationary spreading of tensile zones according to temperature development was calculated. These calculations take into account discrete cracks, reinforcement and different assumptions for tensile strength. Reinforcement of small diameter near the surface has the best influence on crack spacing. Calculations show that for the stationary state of cooling the forces in the reinforcement may be as low as twenty to thirty percent of the tensile force not taking into account cracking of the concrete

  3. High temperature concrete composites containing organosiloxane crosslinked copolymers

    Science.gov (United States)

    Zeldin, A.; Carciello, N.; Kukacka, L.; Fontana, J.

    High temperature polymer concrete composites comprising about 10 to 30% by weight of a liquid monomer mixture is described. It consists essentially of an organosiloxane polymer crosslinked with an olefinically unsaturated monomer selected from the group consisting of styrene, methyl methacrylate, trimethylolpropane trimethacrylate, triallyl cyanurate, n-phenylmalimide, divinyl benzene and mixtures thereof. About 70 to 90% by weight of an inert inorganic filler system containing silica sand and portland cement, Fe/sub 2/O/sub 3/, carbon black or mixtures thereof. Optionally a free radical initiator such as di-tert-butyl peroxide, azobisisobyutyronitrile, benzoyl peroxide, lauryl peroxide and other organic peroxides are used to initiate crosspolymerization of the monomer mixture in the presence of the inorganic filler.

  4. Development of fast reactor containment safety analysis code, CONTAIN-LMR. (3) Improvement of sodium-concrete reaction model

    International Nuclear Information System (INIS)

    Kawaguchi, Munemichi; Doi, Daisuke; Seino, Hiroshi; Miyahara, Shinya

    2015-01-01

    A computer code, CONTAIN-LMR, is an integrated analysis tool to predict the consequence of severe accident in a liquid metal fast reactor. Because a sodium-concrete reaction behavior is one of the most important phenomena in the accident, a Sodium-Limestone Concrete Ablation Model (SLAM) has been developed and installed into the original CONTAIN code at Sandia National Laboratories (SNL) in the U.S. The SLAM treats chemical reaction kinetics between the sodium and the concrete compositions mechanistically using a three-region model, containing a pool (sodium and reaction debris) region, a dry (boundary layer (B/L) and dehydrated concrete) region, and a wet (hydrated concrete) region, the application is limited to the reaction between sodium and limestone concrete. In order to apply SLAM to the reaction between sodium and siliceous concrete which is an ordinary structural concrete in Japan, the chemical reaction kinetics model has been improved to consider the new chemical reactions between sodium and silicon dioxide. The improved model was validated to analyze a series of sodium-concrete experiments which were conducted in Japan Atomic Energy Agency (JAEA). It has been found that relatively good agreement between calculation and experimental results is obtained and the CONTAIN-LMR code has been validated with regard to the sodium-concrete reaction phenomena. (author)

  5. Review of current practices and requirements for the inspection of prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Reimann, K.J.

    1980-12-01

    Code requirements for pre- and in-service inspection of prestressed concrete pressure vessels as utilized in gas-cooled reactors are reviewed and compared with practices and experiences during construction, commissioning, and operation of such reactors. The pre-service inspection relies heavily on embedded instrumentation for measurements of stresses, temperatures, and displacements. The same instrumentation is later used for in-service surveillance, which additionally includes visual examination of exposed surfaces, monitoring of tendon conditions, and measurement of tendon loads. Improvement of present monitoring instrumentation and/or techniques, rather than development of new in-service inspection methods, is recommended

  6. Calculation of Prestressed Pressure Vessel Taking into Account the Concrete Temperature Inhomogeneity

    Science.gov (United States)

    Andreev, Vladimir

    2018-03-01

    The paper deals with the problem of determining the stress state of the pressure vessel (PV) with considering the concrete temperature inhomogeneity. Such structures are widely used in heat power engineering, for example, in nuclear power engineering. The structures of such buildings are quite complex and a comprehensive analysis of the stress state in them can be carried out either by numerical or experimental methods. However, a number of fundamental questions can be solved on the basis of simplified models, in particular, studies of the effect on the stressed state of the inhomogeneity caused by the temperature field.

  7. PARCS - A pre-stressed and reinforced concrete shell element for analysis of containment structures

    International Nuclear Information System (INIS)

    Buragohain, D.N.; Mukherjee, A.

    1993-01-01

    Containment structures are designed as pressure vessels against a huge internal pressure build up in the event of a postulated LOCA. In such situations the containment structures experience predominantly in-plane stress in tension. Therefore, pre-stressed concrete has been very frequently used for the construction of containment. For larger plants a dual containment with a pre-stressed concrete inner containment and a reinforced concrete outer containment has been adopted. These structures are required to perform within very stringent safety requirements under extremely severe loading. Naturally, their design has attracted a lot of investigators and a huge volume of literature has been published in previous SMiRT conferences. However, it seems that the structural modeling of the containment has not developed accordingly. It is a common practice to consider the concrete section only in the model and the effects of pre-stress and reinforcements are usually neglected. This is due to the difficulty in including these effects without generating an unduly large model. To include these effects using the existing software, the concrete can be modeled with 3D elements. The reinforcements can be included in the model as bar or cable elements. However, that would require a nodal line along every reinforcement. Therefore, this method would generate a huge model unmanageable even with modern computing facilities. Alternatively, the reinforcements can be assumed to be smeared uniformly within the structure and an average property can be included. This model is acceptable when the reinforcements are very closely spaced. However, for sparsely spaced reinforcements it would result in loss of accuracy, especially in important areas like the vicinity of large openings. In this paper a shell element for the analysis of pre-stressed and reinforced concrete structures has been proposed which alleviates this difficulty. This element can accommodate the reinforcing bars or cables anywhere

  8. Latest developments in prestressed concrete vessels for gas-cooled reactors

    International Nuclear Information System (INIS)

    Ople, F.S. Jr.

    1979-01-01

    This paper is an update of the design development of prestressed concrete vessels, commonly referred to as 'PCRVs' starting with the first single-cavity PCRV for the Fort St. Vrain Nuclear Generating Station to the latest multi-cavity PCRV configurations being utilized as the primary reactor vessels for both the High Temperature Gas-Cooled Reactor (HTGR) and the Gas-Cooled Fast Breeder Reactor (GCFR) in the U.S.A. The complexity of PCRV design varies not only due to the type of vessel configuration (single versus multi-cavity) but also on the application to the specific type of reactor concept. PCRV technology as applied to the Steam Cycle HTGR is fairly well established; however, some significant technical complexities are associated with PCRV design for the Gas Turbine HTGR and the GCFR. For the Gas Turbine HTGR, for instance, the fluid dynamics of the turbo-machinery cause multi-pressure conditions to exist in various portions of the power conversion loops during operation. This condition complicates the design approach and the proof test specification for the PCRV. The geometric configuration of the multi-cavity PCRV is also more complex due to the introduction of large horizontal cylindrical cavities (housing the turbo/machines for the Gas Turbine HTGR and circulators for the GCFR) in addition to the vertical cylindrical cavities for the core and heat exchangers. Because of this complex geometry, it becomes difficult to achieve an optimum prestressing arrangement for the PCRV. Other novel features of the multi-cavity PCRV resulting from the continuing design optimization effort are the incorporation of an asymmetric (offset core) configuration and the use of large vessel cavity/penetration concrete closures directly held down by prestressing tendons for both economic and safety reasons. (orig.)

  9. A study on the improvement of ISI methods for a prestressed concrete containment building

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Jeong Moon; Choi, In Kil

    2001-12-01

    The ISI (In-Service Inspection) of a PCCV (Prestressed Concrete Containment Vessel) consists of the tendon ISI and the SIT (Structural Integrity Test) which evaluate the effective or residual prestress in the post-tensioned prestress system, and of the ILRT (Integrated Leakage Rate Test) which ensures the leak-tightness of a PCCV. The tendon system adopted in Korean PCCVs is either grouted or ungrouted one. The grouted tendon system was used in the Ulchin Unit 1 and 2 and the Wolsong Units 1-4, whereas the rest NPPs except Kori Unit 1 and 2 adopted an ungrouted tendon system. In this report, the issues were identified on the ISI of ungrouted tendon system and on the SIT of all the PCCVs. The ILRT issues are implicitly included in the SIT issues as the ILRT is performed in parallel with the SIT. Improvements were suggested on the issues identified after the analyses of the domestic and foreign experiences and researches.

  10. Containment vessel bottom head transport and lifting technique

    International Nuclear Information System (INIS)

    Zheng Donghong; Tian Shiyong; Hu Dequan; Xiao Hongtao

    2013-01-01

    The challengeable transport and lifting techniques and high safety assurance measures are needed for the onsite construction of the AP1000 containment vessel bottom head (CVBH), which is a large component with heavy weight, big size, high center of gravity, and easy to deformation. During transport, the infra structural road foundation is heavily loaded with big turning radius, and the requirement for synchronization of transport vehicles is strict. During lifting, the crane lifting capacities are high, requirement for the lifting and rigging tools is strict, nuclear island being put into place is difficult, and the crane operating foundation is heavily loaded. The transport and lifting techniques and safety assurance measures for CVBH are elaborated in detail, so as to provide a reference for the follow-up transport and lifting of large components of nuclear island. (authors)

  11. Review on experiments relating to primary containment vessel failure

    International Nuclear Information System (INIS)

    Suzuki, Hiroyuki; Okada, Hidetoshi; Uchida, Sunsuke; Naitoh, Masanori

    2015-01-01

    Experiments regarding failures of primary containment vessels (PCVs) are reviewed and remained issues to be investigated in the future are discussed. Experiments are categorized as those relating to criteria of PCV failures and to FP releases through breaches on PCV boundaries. In the experiments categorized as those relating to criteria of PCV failures, experiments with full-scale, scale models, and compounds used for sealing are surveyed. Experiments relating to an amount of radioactive fission products (FPs) trapped at breaches on PCV boundaries are also reviewed. As remained issues to be investigated in the future, two items are pointed out: Evaluating degradation behavior of PCV boundaries exposed to temperature and pressure from the failure onset criteria to far above them, and evaluating an amount of FPs trapped at breaches on PCV boundaries. (author)

  12. Ductile fracture of cylindrical vessels containing a large flaw

    Science.gov (United States)

    Erdogan, F.; Irwin, G. R.; Ratwani, M.

    1976-01-01

    The fracture process in pressurized cylindrical vessels containing a relatively large flaw is considered. The flaw is assumed to be a part-through or through meridional crack. The flaw geometry, the yield behavior of the material, and the internal pressure are assumed to be such that in the neighborhood of the flaw the cylinder wall undergoes large-scale plastic deformations. Thus, the problem falls outside the range of applicability of conventional brittle fracture theories. To study the problem, plasticity considerations are introduced into the shell theory through the assumptions of fully-yielded net ligaments using a plastic strip model. Then a ductile fracture criterion is developed which is based on the concept of net ligament plastic instability. A limited verification is attempted by comparing the theoretical predictions with some existing experimental results.

  13. Development of neutron shielding concrete containing iron content materials

    Science.gov (United States)

    Sariyer, Demet; Küçer, Rahmi

    2018-02-01

    Concrete is one of the most important construction materials which widely used as a neutron shielding. Neutron shield is obtained of interaction with matter depends on neutron energy and the density of the shielding material. Shielding properties of concrete could be improved by changing its composition and density. High density materials such as iron or high atomic number elements are added to concrete to increase the radiation resistance property. In this study, shielding properties of concrete were investigated by adding iron, FeB, Fe2B, stainless - steel at different ratios into concrete. Neutron dose distributions and shield design was obtained by using FLUKA Monte Carlo code. The determined shield thicknesses vary depending on the densities of the mixture formed by the additional material and ratio. It is seen that a combination of iron rich materials is enhanced the neutron shielding of capabilities of concrete. Also, the thicknesses of shield are reduced.

  14. Design criteria for prestressed concrete pressure vessels for high temperature reactors

    International Nuclear Information System (INIS)

    Schimmelpfennig, K.

    1991-01-01

    This paper summarizes the work on design criteria for concrete structures of Prestressed Concrete Reactor Vessels (PCRVs), which has been carried out since 1984 by a couple of competent institutions. After some basic considerations on the safety demands on PCRVs, especially their Prestressed Concrete Structure (PCS), and the consequences for an elevated level of quality to be ensured by the design criteria, an impression is given, first, by what means a higher quality standard is gained with respect to selection of materials and specification of material data in comparison to the usual building industry and what kind of criteria on this behalf should be fixed in a PCRV code. As a further quality increasing feature, the specific demands on design analysis as practised according to the present state of science and as to be treated within a code are discussed. This concerns analyses for steady state and transient temperatures as well as stress and strain analyses for service and ultimate load conditions. It is outlined to what degree calculation models should be detailed, which includes statements about admissible idealizations. As a central topic the question is discussed in what way the ultimate load capacity has to be evaluated, thereby presenting results of some investigations pointing out the conditions under which the design is determined by the different kinds of ultimate load conditions. Finally, some reflections on the demands on monitoring the PCS behaviour during its lifetime and on several questions still to be answered in this field are expressed. (orig.)

  15. Study on Concrete Containing Recycled Aggregates Immersed in Epoxy Resin

    Directory of Open Access Journals (Sweden)

    Adnan Suraya Hani

    2017-01-01

    Full Text Available In recent decades, engineers have sought a more sustainable method to dispose of concrete construction and demolition waste. One solution is to crush this waste concrete into a usable gradation for new concrete mixes. This not only reduces the amount of waste entering landfills but also alleviates the burden on existing sources of quality natural concrete aggregates. There are too many kinds of waste but here constructions waste will be the priority target that should be solved. It could be managed by several ways such as recycling and reusing the concrete components, and the best choice of these components is the aggregate, because of the ease process of recycle it. In addition, recycled aggregates and normal aggregates were immersed in epoxy resin and put in concrete mixtures with 0%, 5%, 10% and 20% which affected the concrete mixtures properties. The strength of the concrete for both normal and recycled aggregates has increased after immersed the aggregates in epoxy resin. The percentage of water absorption and the coefficient of water permeability decreased with the increasing of the normal and the recycled aggregates immersed in epoxy resin. Generally the tests which have been conducted to the concrete mixtures have a significant results after using the epoxy resin with both normal and recycled aggregates.

  16. Tests on model of a prestressed concrete nuclear pressure vessel with multiple cavities

    International Nuclear Information System (INIS)

    Favre, R.; Koprna, M.; Jaccoud, J.P.

    1977-01-01

    The prestressed concrete pressure vessel (prototype) is a cylinder having a diameter of 48 m and a height of 39 m. It has 25 vertical cavities (reactor, heat exchangers, heat recuperators) and 3 horizontal cavities (gas turbines of 500 kw). The cavities are closed by plugs, and their tightness is ensured by a steel lining. A model, on a scale of 1/20, made of microconcrete, was loaded in several cycles, by a uniform inner pressure in the cavities, increasing to the point of failure. The three successive stages were examined: stage of globally elastic behavior, cracking stage, ultimate stage. The behavior of the model is globally elastic up to an inner pressure of 120 to 130 kp/cm 2 , corresponding to about twice the maximum pressure of service, equal to 65 kp/cm 2 . The prestressed tendons at this stage show practically no stress increase. The first detectable cracks appear on the lateral side half-way up the model, as soon as the pressure exceeded 120 kp/cm 2 . From 150-165 kp/cm 2 , the cracking stage can be considered as achieved and the main crack pattern entirely formed. A horizontal crack continues in the middle of the barrel, as well as vertical cracks at each outer cavity. Beyond a pressure of 150-165 kp/cm 2 the ultimate stage begins. The strains of the stresses in the tendons grow more rapidly. The steel lining is highly solicited. Above about 210 kp/cm 2 the model behaves like a structure composed of a group of concrete blocks bound by the tendons and the lining. The failure (240 kp/cm 2 ) occurred through a mechanism of ejection and bending of the concrete ring at the periphery of the barrel of the vessel, which was solicited mainly in tension

  17. 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)

  18. Stress analysis of LOFT containment vessel attachments for the mainsteam and feedwater piping support structures

    International Nuclear Information System (INIS)

    Finicle, D.P.

    1977-01-01

    The LOFT Containment Vessel attachments for the Mainsteam and Feedwater Piping Support Structures have been analyzed for operating and faulted loading conditions. This report contains the analysis of the connections to the containment vessel for the most current design and loading. Also contained in this report is the analysis of the piping supports

  19. Constitutive relation of concrete containing meso-structural characteristics

    Directory of Open Access Journals (Sweden)

    Li Guo

    Full Text Available A constitutive model of concrete is proposed based on the mixture theory of porous media within thermodynamic framework. By treating concrete as a multi-phase multi-component mixture, we constructed the constitutive functions for elastic, interfacial, and plastic strain energy respectively. A constitutive law of concrete accommodating internal micro-cracks and interfacial boundaries was established. The peak stress predicted with the developed model depends primarily on the volume ratio of aggregate, and the results explain very well reported experimental phenomena. The strain-stress curve under uniaxial loading was found in a good agreement with experimental data for concrete with three different mixing proportions. Keywords: Constitutive model of concrete, Mixture theory of porous media, Meso-structure, Interfacial energy

  20. Meeting 'Prestressed-concrete reactor pressure vessels', 13th and 14th october 1975, Berlin

    International Nuclear Information System (INIS)

    Schickert, G.

    1976-01-01

    Influence of radioactive radiation on the mechanical properties of concrete; behaviour of concrete in short-time testing under multiaxial mechanical stresses; behaviour of concrete in long-time testing under multiaxial mechanical stresses at higher temperatures; temperature stress of concrete; strength formation of concrete; steel fiber concrete. (LH) [de

  1. Reliability-based design code calibration for concrete containment structures

    International Nuclear Information System (INIS)

    Han, B.K.; Cho, H.N.; Chang, S.P.

    1991-01-01

    In this study, a load combination criteria for design and a probability-based reliability analysis were proposed on the basis of a FEM-based random vibration analysis. The limit state model defined for the study is a serviceability limit state of the crack failure that causes the emission of radioactive materials, and the results are compared with the case of strength limit state. More accurate reliability analyses under various dynamic loads such as earthquake loads were made possible by incorporating the FEM and random vibration theory, which is different from the conventional reliability analysis method. The uncertainties in loads and resistance available in Korea and the references were adapted to the situation of Korea, and especially in case of earthquake, the design earthquake was assessed based on the available data for the probabilistic description of earthquake ground acceleration in the Korea peninsula. The SAP V-2 is used for a three-dimensional finite element analysis of concrete containment structure, and the reliability analysis is carried out by modifying HRAS reliability analysis program for this study. (orig./GL)

  2. General requirements for concrete containment structures for CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1993-07-01

    This standard provides the general requirements used in the design, construction, testing, and commissioning of concrete containment structures for CANDU nuclear power plants designated as class containment and is directed to the owners, designers, manufacturers, fabricators, and constructors of the concrete components and parts

  3. High-impact concrete for fill in US Department of Transportation type shipping containers

    International Nuclear Information System (INIS)

    Greenhalgh, W.O.; Cash, R.J.

    1990-01-01

    This report describes the use of light-weight, high-impact concrete in U.S. Department of Transportation-type shipments. The formulations described are substantially lighter in weight (20 to 50 percent) than construction concrete, but product test specimens generally yield superior impact characteristics. The use of this specialty concrete for container fill, encapsulations, or liquid-waste solidification can be advantageous. Use of the material for container or cask construction has the advantage of lighter weight for easier handling, and the container consistently exhibits better performance on drop tests. High-impact concrete does have the disadvantage of less gamma radiation shielding per volume, but some formulation changes discussed in this report can be used to prepare better shielding concrete. Test characteristics of high-impact concrete are included. 3 refs., 6 figs., 7 tabs

  4. Design and construction of a prestressed concrete pressure vessel for a working pressure of 69N/mm2 (10,000 p.s.i)

    International Nuclear Information System (INIS)

    Dawson, P.

    1977-01-01

    Construction is nearing completion of a pressure vessel with a chamber 9.15 m (30 ft.) high and 3.05 m (10 ft.) internal diameter for hydraulic tests on marine components up to 69 N/mm 2 (10,000 p.s.i.) working pressure. The chamber comprises a steel cylinder, with independent end plates contained within a prestressed concrete structure. The cylinder is constructed in two halves, each consisting of three forged rings, 170 mm thick, shrink-fitted onto a 90 mm thick liner. It rests on a 100 mm thick bottom plate, provided with a band of hard-facing overlay on which the cylinder slides in response to changes of test medium pressure. Models to be tested within the chamber are hung from a removeable 150 mm thick top plate. A central elliptical hatch provides access into the chamber. Special sealing assemblies are fitted at the junction of the cylinder sections and between the cylinder and end plates. These seals are capable of accepting radial expansion of the cylinder and corresponding vertical movements at the upper seal arising from elastic movements of the enclosing structure. The top plate is restrained by a wire-wound prestressed concrete closure plug, itself located by twelve bifurcated inclined steel struts which transfer the load on the top plate into the concrete structure. The struts are retractable to allow removal of the closure plug and top plate. The enclosing concrete structure is 25 m (82 ft.) high and 11 m (36 ft.) diameter. It is vertically prestressed by 180 no. 540 Tonne tendons and circumferentially prestressed by 5 mm wire laid under tension in pre-cast concrete channels by the Taylor Woodrow Wire-Winding System. The structure was analysed, using limit state principles, by computerised elastic and non-elastic dynamic relaxation techniques. The results were evaluated against triaxial stress criteria established from relevant research work and experience obtained from nuclear prestressed concrete pressure vessels

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

    International Nuclear Information System (INIS)

    Fisher, Kane J.

    2000-01-01

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

  6. The application of external vibration monitoring to reactors with concrete pressure vessels

    International Nuclear Information System (INIS)

    Hammill, W.J.

    1979-01-01

    The application of external vibration monitoring techniques to advanced gas cooled reactors (AGR) which have concrete pressure vessels is considered. A monitoring system for a particular AGR coolant circuit structure is developed, whose primary objective is to detect impacting of two components, although the detection of forced vibration response is also considered. Experimental results from instrumented components in the reactor and data from rig tests on full size units have been used together with a mathematical model of some elements of the transmission path in order to establish its dynamic characteristics and relate internal component vibration to externally measured signals. The application of external vibration monitoring to the external detection of the forced vibration response of an internal reactor assembly and the remote monitoring of circulator sound output is discussed. (author)

  7. Experimental analysis of a nuclear reactor prestressed concrete pressure vessels model

    International Nuclear Information System (INIS)

    Vallin, C.

    1980-01-01

    A comprehensible analysis was made of the performance of each set of sensors used to measure the strain and displacement of a 1/20 scale Prestressed Concrete Pressure Vessel (PCPV) model tested at the Instituto de Pesquisas Energeticas e Nucleares (IPEN). Among the three Kinds of sensors used (strain gage, displacement transducers and load cells) the displacement transducers showed the best behavior. The displacemente transducers data was statistically analysed and a linear behavior of the model was observed during the first pressurizations tests. By means of a linear statistical correlation between experimental and expected theoretical data it was found that the model looses the linearity at a pressure between 110-125 atm. (Author) [pt

  8. Experience in surveillance of the prestress of concrete reactor vessels in Wylfa nuclear power station

    International Nuclear Information System (INIS)

    Dawson, P.; Paton, A.A.; Walsh, S.R.

    1989-01-01

    This paper describes experience gained in the in-service surveillance of the prestressing system for the prestressed concrete reactor vessels (PCRVs) at Wylfa nuclear power station. The paper gives details of results for the prestressing system obtained from the statutory in-service inspection program of the PCRVs. The program includes a detailed examination of a selection of prestressing tendon anchorages, anchorage load checks using a lift-off technique on a one percent sample of tendons and corrosion inspection of samples of prestressing strand and determination of their mechanical properties. The results obtained from the above in-service inspections have shown that the prestressing system continues to function within its design limits

  9. Leakage detecting method and device for water tight vessel of wet-type container apparatus

    International Nuclear Information System (INIS)

    Tanaka, Yoshimi.

    1995-01-01

    The present invention provides a method of and a device for detecting leakage of a water tight vessel of a wet-type container apparatus for containing a reactor pressure vessel while immersing it water in a reactor container. Namely, in the wet-type container apparatus, the periphery of the pressure vessel is coated with a heat insulation material and the periphery of the heat insulation material is coated with a water tight vessel. The water tight vessel is immersed under water in the reactor container. As a method of detecting leakage of the wet-type container apparatus, gases mixed with helium are supplied into the water tight vessel at a pressure higher than the inner pressure of the reactor container at a lowest position of the reactor pressure vessel. A water level in the reactor container is determined so as to form a space at the top portion of the inside of the reactor container. The helium at the top portion is detected to monitor the leakage of the water tight vessel. With such procedures, even if the water tight vessel is ruptured at any position, helium mixed to the gases is released to water in the reactor container and rise up to the top space and detected by a helium leakage detection device. (I.S.)

  10. TEMP-STRESS analysis of a reinforced concrete vessel under internal pressure

    International Nuclear Information System (INIS)

    Marchertas, A.H.; Kennedy, J.M.; Pfeiffer, P.A.

    1987-01-01

    Prediction of the response of the Sandia National laboratory 1/6-scale reinforced concrete containment model test was obtained by Argonne National Laboratory (ANL) employing a computer program developed by ANL. The test model was internally pressurized to failure. The two-dimensional code TEMP-STRESS [1-5] has been developed at ANL for stress analysis of plane and axisymmetric 2-D reinforced structures under various thermal conditions. The program is applicable to a wide variety of nonlinear problems, and is utilized in the present study. The comparison of these pretest computations with test data on the containment model should be a good indication of the state of the code

  11. Properties of dune sand concrete containing coffee waste

    Directory of Open Access Journals (Sweden)

    Mohamed Guendouz

    2018-01-01

    Full Text Available In the last years, an increase of coffee beverages consumption has been observed all over the world; and its consumption increases the waste coffee grounds which will become an environmental problems. Recycling of this waste to produce new materials like sand concrete appears as one of the best solutions for reduces the problem of pollution. This work aims to study the possibility of recycling waste coffee grounds (Spent Coffee Grounds (SCG as a fine aggregate by replacing the sand in the manufacturing of dune sand concrete. For this; sand concrete mixes were prepared with substitution of sand with the spent coffee grounds waste at different percentage (0%, 5%, 10%, 15% and 20% by volume of the sand in order to study the influence of this wastes on physical (Workability, bulk density and porosity, mechanical (compressive and flexural strength and Thermal (Thermal conductivity and thermal diffusivity properties of dune sand concrete. The results showed that the use of spent coffee grounds waste as partial replacement of natural sand contributes to reduce workability, bulk density and mechanical strength of sand concrete mixes with an increase on its porosity. However, the thermal characteristics are improved and especially for a level of 15% and 20% of substitution. So, it is possible to obtain an insulating material which can be used in the various types of structural components. This study ensures that reusing of waste coffee grounds in dune sand concrete gives a positive approach to reduce the cost of materials and solve some environmental problems.

  12. Radioactive waste processing vessel

    International Nuclear Information System (INIS)

    Hayashi, Masaru; Suzuki, Osamu; Ishizaki, Kanjiro.

    1987-01-01

    Purpose: To obtain a vessel of a reduced weight and with no external leaching of radioactive materials. Constitution: The vessel main body is constituted, for example, with light weight concretes or foamed concretes, particularly, foamed concretes containing fine closed bubbles in the inside. Then, layers having dense texture made of synthetic resin such as polystylene, vinylchloride resin, etc. or metal plate such as stainless plate are integrally disposed to the inner surface of the vessel main body. The cover member also has the same structure. (Sekiya, K.)

  13. The nonlinear finite element analysis program NUCAS (NUclear Containment Analysis System) for reinforced concrete containment building

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Jin; Lee, Hong Pyo; Seo, Jeong Moon [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-03-01

    The maim goal of this research is to develop a nonlinear finite element analysis program NUCAS to accurately predict global and local failure modes of containment building subjected to internal pressure. In this report, we describe the techniques we developed throught this research. An adequate model to the analysis of containment building such as microscopic material model is adopted and it applied into the development Reissner-Mindlin degenerated shell element. To avoid finite element deficiencies, the substitute strains based on the assumed strain method is used in the shell formulation. Arc-length control method is also adopted to fully trace the peak load-displacement path due to crack formation. In addition, a benchmark test suite is developed to investigate the performance of NUCAS and proposed as the future benchmark tests for nonlinear analysis of reinforced concrete. Finally, the input format of NUCAS and the examples of input/output file are described. 39 refs., 65 figs., 8 tabs. (Author)

  14. Comparison of elastic--plastic and variable modulus-cracking constitutive models for prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Anderson, C.A.; Smith, P.D.

    1978-01-01

    The variable modulus-cracking model is capable of predicting the behavior of reinforced concrete structures (such as the reinforced plate under transverse pressure described previously) well into the range of nonlinear behavior including the prediction of the ultimate load. For unreinforced thick-walled concrete vessels under internal pressure the use of elastic--plastic concrete models in finite element codes enhances the apparent ductility of the vessels in contrast to variable modulus-cracking models that predict nearly instantaneous rupture whenever the tensile strength at the inner wall is exceeded. For unreinforced thick-walled end slabs representative of PCRV heads, the behavior predicted by finite element codes using variable modulus-cracking models is much stiffer in the nonlinear range than that observed experimentally. Although the shear type failures and crack patterns that are observed experimentally are predicted by such concrete models, the ultimate load carrying capacity and vessel-ductility are significantly underestimated. It appears that such models do not adequately model such features as aggregate interlock that could lead to an enhanced vessel reserve strength and ductility

  15. Analysis code for pressure in reactor containment vessel of ATR. CONPOL

    International Nuclear Information System (INIS)

    1997-08-01

    For the evaluation of the pressure and temperature in containment vessels in the events which are classified in the abnormal change of pressure, atmosphere and others in reactor containment vessels in accident among the safety evaluation events of the ATR, the analysis code for the pressure in reactor containment vessels CONPOL is used. In this report, the functions of the analysis code and the analysis model are shown. By using this analysis code, the rise of the pressure and temperature in a containment vessel is evaluated when loss of coolant accident occurs, and high temperature, high pressure coolant flows into it. This code possesses the functions of computing blow-down quantity and heat dissipation from reactor cooling facility, steam condensing heat transfer to containment vessel walls, and the cooling effect by containment vessel spray system. As for the analysis techniques, the models of reactor cooling system, containment vessel and steam discharge pool, and the computation models for the pressure and temperature in containment vessels, wall surface temperature, condensing heat transfer, spray condensation and blow-down are explained. The experimental analysis of the evaluation of the pressure and temperature in containment vessels at the time of loss of coolant accident is reported. (K.I.)

  16. Mechanical properties of cement concrete composites containing nano-metakaolin

    Science.gov (United States)

    Supit, Steve Wilben Macquarie; Rumbayan, Rilya; Ticoalu, Adriana

    2017-11-01

    The use of nano materials in building construction has been recognized because of its high specific surface area, very small particle sizes and more amorphous nature of particles. These characteristics lead to increase the mechanical properties and durability of cement concrete composites. Metakaolin is one of the supplementary cementitious materials that has been used to replace cement in concrete. Therefore, it is interesting to investigate the effectiveness of metakaolin (in nano scale) in improving the mechanical properties including compressive strength, tensile strength and flexural strength of cement concretes. In this experiment, metakaolin was pulverized by using High Energy Milling before adding to the concrete mixes. The pozzolan Portland cement was replaced with 5% and 10% nano-metakaolin (by wt.). The result shows that the optimum amount of nano-metakaolin in cement concrete mixes is 10% (by wt.). The improvement in compressive strength is approximately 123% at 3 days, 85% at 7 days and 53% at 28 days, respectively. The tensile and flexural strength results also showed the influence of adding 10% nano-metakaolin (NK-10) in improving the properties of cement concrete (NK-0). Furthermore, the Backscattered Electron images and X-Ray Diffraction analysis were evaluated to support the above findings. The results analysis confirm the pores modification due to nano-metakaolin addition, the consumption of calcium hydroxide (CH) and the formation of Calcium Silicate Hydrate (CSH) gel as one of the beneficial effects of amorphous nano-metakaolin in improving the mechanical properties and densification of microstructure of mortar and concrete.

  17. Calculations of concrete containment tight loss: Studies of a reinforced concrete slab with non uniform thickness

    International Nuclear Information System (INIS)

    Jamet, P.; Berriaud, C.; Humbert, J.M.; Millard, A.; Nahas, G.

    1983-01-01

    A study was carried out in order to investigate the validity of a concrete model including tensile fracture and strain-softening under compressive loading. Triaxial tests were performed on micro-concrete specimens, and the post-peak behaviour of the material was characterized. The parameters required by the model were therefore obtained. The case of a circular slab loaded up to failure was then considered, in order to compare the numerical results obtained by a finite elements analysis including the concrete model, to the experimental data. (orig.)

  18. Calculations of concrete containment tight loss: studies of a reinforced concrete SLAB with non uniform thickness

    International Nuclear Information System (INIS)

    Jamet, P.

    1983-08-01

    A study was carried out in order to investigate the validity of a concrete model including tensile fracture and strain-softening under compressive loading. Triaxial tests were performed on micro-concrete specimens, and the post-peak behaviour of the material was characterized. The parameters required by the model were therefore obtained. The case of a circular slab loaded up to failure was then considered, in order to compare the numerical results obtained by a finite elements analysis including the concrete model, to the experimental data

  19. Overview of experimental results obtained under the Prestressed Concrete Nuclear Pressure Vessel Development Program at the Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Naus, D.J.

    1978-01-01

    Under the Prestressed Concrete Nuclear Pressure Vessel Development Program at the Oak Ridge National Laboratory, various aspects of Prestressed Concrete Pressure Vessels (PCPVs) are investigated and evaluated with respect to reliability, structural performance, constructability, and economy. Based upon identified needs, analytical and experimental investigations are conducted. Areas of interest include finite-element analysis development, materials and structural behavior tests, instrumentation evaluation and development, and structural model tests. Studies have been recently completed in the following areas: concrete embedment instrumentation systems for PCPVs, grouted-nongrouted prestressing systems, acoustic emission as a technique for structural integrity monitoring, and model tests of steam-generator cavity closure plugs for a Gas-Cooled Fast Reactor (GCFR). An overview of results is presented

  20. Mechanical Properties of Fiber Reinforced Lightweight Concrete Containing Surfactant

    Directory of Open Access Journals (Sweden)

    Yoo-Jae Kim

    2010-01-01

    Full Text Available Fiber reinforced aerated lightweight concrete (FALC was developed to reduce concrete's density and to improve its fire resistance, thermal conductivity, and energy absorption. Compression tests were performed to determine basic properties of FALC. The primary independent variables were the types and volume fraction of fibers, and the amount of air in the concrete. Polypropylene and carbon fibers were investigated at 0, 1, 2, 3, and 4% volume ratios. The lightweight aggregate used was made of expanded clay. A self-compaction agent was used to reduce the water-cement ratio and keep good workability. A surfactant was also added to introduce air into the concrete. This study provides basic information regarding the mechanical properties of FALC and compares FALC with fiber reinforced lightweight concrete. The properties investigated include the unit weight, uniaxial compressive strength, modulus of elasticity, and toughness index. Based on the properties, a stress-strain prediction model was proposed. It was demonstrated that the proposed model accurately predicts the stress-strain behavior of FALC.

  1. Durability of conventional concretes containing black rice husk ash.

    Science.gov (United States)

    Chatveera, B; Lertwattanaruk, P

    2011-01-01

    In this study, black rice husk ash (BRHA) from a rice mill in Thailand was ground and used as a partial cement replacement. The durability of conventional concretes with high water-binder ratios was investigated including drying shrinkage, autogenous shrinkage, depth of carbonation, and weight loss of concretes exposed to hydrochloric (HCl) and sulfuric (H(2)SO(4)) acid attacks. Two different replacement percentages of cement by BRHA, 20% and 40%, and three different water-binder ratios (0.6, 0.7 and 0.8) were used. The ratios of paste volume to void content of the compacted aggregate (γ) were 1.2, 1.4, and 1.6. As a result, when increasing the percentage replacement of BRHA, the drying shrinkage and depth of carbonation reaction of concretes increased. However, the BRHA provides a positive effect on the autogenous shrinkage and weight loss of concretes exposed to hydrochloric and sulfuric acid attacks. In addition, the resistance to acid attack was directly varied with the (SiO(2) + Al(2)O(3) + Fe(2)O(3))/CaO ratio. Results show that ground BRHA can be applied as a pozzolanic material and also improve the durability of concrete. Copyright © 2010 Elsevier Ltd. All rights reserved.

  2. A Review of the Mechanical Properties of Concrete Containing Biofillers

    Science.gov (United States)

    Ezdiani Mohamad, Mazizah; Mahmood, Ali A.; Min, Alicia Yik Yee; Khalid, Nur Hafizah A.

    2016-11-01

    Sustainable construction is a rapidly increasing research area. Investigators of all backgrounds are using industrial and agro wastes to replace Portland cement in concrete to reduce greenhouse emissions and the corresponding decline in general health. Many types of wastes have been used as cement replacements in concrete including: fly ash, slag and rice husk ash in addition to others. This study investigates the possibility of producing a sustainable approach to construction through the partial replacement of concrete using biofillers. This will be achieved by studying the physical and mechanical properties of two widely available biological wastes in Malaysia; eggshell and palm oil fuel ash (POFA). The mechanical properties tests that were studied and compared are the compression, tensile and flexural tests.

  3. Optimisation by mathematical modeling of physicochemical characteristics of concrete containers in radioactive waste management

    Directory of Open Access Journals (Sweden)

    Plećaš Ilija

    2013-01-01

    Full Text Available A method for obtaining an optimal concrete container composition used for storing radioactive waste from nuclear power plants is developed. It is applied to the radionuclides 60Co, 137Cs, 85Sr, and 54Mn. A set of recipes for concrete composition leading to an optimal solution is given.

  4. Strategy for 100-year life of the ACR-1000 concrete containment structure

    International Nuclear Information System (INIS)

    Abrishami, H.; Elgohary, M.

    2006-01-01

    The purpose of this paper is to present the Plant Life Management (PLiM) strategy for the concrete containment structure of the ACR-1000 (Advanced CANDU Reactor) designed by AECL. The ACR-1000 is designed for 100-year plant life including 60-year operating life and additional 40-year decommissioning period of time. The approach adopted for the PLiM strategy of the concrete containment structure is a preventive one, key areas being: 1) design methodology, 2) material performance and 3) life cycle management and ageing management program. In the design phase, in addition to strength and serviceability, durability is a major requirement during the service life and decommissioning phase of the ACR structure. Parameters affecting durability design include: a) concrete performance, b) structural application, and c) environmental conditions. Due to the complex nature of the environmental effects acting on structures during the service life of project, it is considered that true improved performance during the service life can be achieved by improving the material characteristics. Many recent innovations in advanced concrete materials technology have made it possible to produce modern concrete such as high-performance concrete with exceptional performance characteristics. In this paper, the PLiM strategy for the ACR-1000 concrete containment is presented. In addition to addressing the design methodology and material performance areas, a systematic approach for ageing management program for the concrete containment structure is presented. (author)

  5. Corrosion of steel tendons in concrete pressure vessels: review of recent literature and experimental investigations

    International Nuclear Information System (INIS)

    Griess, J.C.

    1978-01-01

    The fundamentals of localized corrosion are briefly discussed, and the literature concerning corrosion of carbon steel in aqueous environments, in particular the stress-corrosion cracking of carbon steels, is reviewed. The behavior of high strength steels in specific environments, including concrete and organic substances, is also summarized. The available information indicates that the corrosion of steels in correctly formulated concrete is minimal. Even appreciable concentrations of chloride, sulfate, sulfide, and nitrate salts can be tolerated in the concrete or grout without detrimental effects. Adherence to established standards in the preparation and application of grouts in tendon-bearing conduits should guarantee very long tendon lifetimes. Little is reported about the behavior of tendons in proprietary organic greases or waxes, but very good corrosion resistance is expected if the organic material remains intact. Stress-corrosion cracking tests performed with AISI 1080 steel tendon wires, using the constant-strain-rate method, produced results expected from data in the literature. Cracking was observed only in neutral or acid solutions containing hydrogen sulfide, in ammonium nitrate solutions, and possibly in a dilute solution of sodium bisulfite. General corrosion tests in water and in dilute solutions of sodium nitrate, chloride, or sulfate showed that oxygen was an important factor; corrosion was substantially greater when oxygen had free access to the solution than when access to oxygen was restricted. In the tests with oxygen the heaviest attack on the steel tendons was at the waterline of the solution

  6. Modeling of delayed strains of concrete under biaxial loadings. Application to the reactor containment of nuclear power plants

    International Nuclear Information System (INIS)

    Benboudjema, F.

    2002-12-01

    The prediction of delayed strains is of crucial importance for durability and long-term serviceability of concrete structures (bridges, containment vessels of nuclear power plants, etc.). Indeed, creep and shrinkage cause cracking, losses of pre-stress and redistribution of stresses, and also, rarely, the ruin of the structure. The objective of this work is to develop numerical tools, able to predict the long-term behavior of concrete structures. Thus, a new hydro mechanical model is developed, including the description of drying, shrinkage, creep and cracking phenomena for concrete as a non-saturated porous medium. The modeling of drying shrinkage is based on an unified approach of creep and shrinkage. Basic and drying creep models are based on relevant chemo-physical mechanisms, which occur at different scales of the cement paste. The basic creep is explicitly related to the micro-diffusion of the adsorbed water between inter-hydrates and intra-hydrates and the capillary pores, and the sliding of the C-S-H gel at the nano-porosity level. The drying creep is induced by the micro-diffusion of the adsorbed water at different scales of the porosity, under the simultaneous effects of drying and mechanical loadings. Drying shrinkage is, therefore, assumed to result from the elastic and delayed response of the solid skeleton, submitted to both capillary and disjoining pressures. Furthermore, the cracking behavior of concrete is described by an orthotropic elastoplastic damage model. The coupling between all these phenomena is performed by using effective stresses which account for both external applied stresses and pore pressures. This model has been incorporated into a finite element code. The analysis of the long-term behavior is also performed on concrete specimens and prestressed concrete structures submitted to simultaneous drying and mechanical loadings. (author)

  7. HECLA experiments on interaction between metallic melt and hematite-containing concrete

    Energy Technology Data Exchange (ETDEWEB)

    Sevon, Tuomo, E-mail: tuomo.sevon@vtt.f [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT, Espoo (Finland); Kinnunen, Tuomo; Virta, Jouko; Holmstroem, Stefan; Kekki, Tommi; Lindholm, Ilona [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT, Espoo (Finland)

    2010-10-15

    In a hypothetical severe accident in a nuclear power plant, molten materials may come into contact with concrete, causing concrete ablation. In five HECLA experiments the interaction between metallic melt and concrete was investigated by pouring molten stainless steel at almost 1800 {sup o}C into cylindrical concrete crucibles. The tests were transient, i.e. no decay heat simulation was used. The main objective was to test the behavior of the FeSi concrete, containing hematite (Fe{sub 2}O{sub 3}) and siliceous aggregates. This special concrete type is used as a sacrificial layer in the Olkiluoto 3 EPR reactor pit, and very scarce experimental data is available about its behavior at high temperatures. It is concluded that no clear differences between the ablation of FeSi concrete and ordinary siliceous concrete were observed. The ablation depths were small, 25 mm at maximum. No dramatic effects, such as cracking of large pieces of concrete due to the thermal shock, took place. An important side result of the test series was gaining knowledge of the properties of the special concrete type. Chemical analyses were conducted and mechanical properties were measured.

  8. Biaxial failure criteria and stress-strain response for concrete of containment structure

    International Nuclear Information System (INIS)

    Lee, S. K.; Woo, S. K.; Song, Y. C.; Kweon, Y. K.; Cho, C. H.

    2001-01-01

    Biaxial failure criteria and stress-strain response for plain concrete of containment structure on nuclear power plants are studied under uniaxial and biaxial stress(compression-compression, compression-tension, and tension-tension combined stress). The concrete specimens of a square plate type are used for uniaxial and biaxial loading. The experimental data indicate that the strength of concrete under biaxial compression, f 2 /f 1 =-1/-1, is 17 percent larger than under uniaxial compression and the poisson's ratio of concrete is 0.1745. On the base of the results, a biaxial failure envelope for plain concrete that the uniaxial strength is 5660 psi are provided, and the biaxial failure behaviors for three biaxial loading areas are plotted respectively. And, various analytical equations having the reliability are proposed for representations of the biaxial failure criteria and stress-strain response curves of concrete

  9. Properties of concrete containing coconut shell powder (CSP) as a filler

    Science.gov (United States)

    Leman, A. S.; Shahidan, S.; Nasir, A. J.; Senin, M. S.; Zuki, S. S. Mohd; Ibrahim, M. H. Wan; Deraman, R.; Khalid, F. S.; Azhar, A. T. S.

    2017-11-01

    Coconut shellsare a type of agricultural waste which can be converted into useful material. Therefore,this study was conducted to investigate the properties of concrete which uses coconut shell powder (CSP) filler material and to define the optimum percentage of CSP which can be used asfiller material in concrete. Comparisons have been made between normal concrete mixes andconcrete containing CSP. In this study, CSP was added into concrete mixes invaryingpercentages (0%, 2%, 4%, 6%, 8% and 10%). The coconut shell was grounded into afine powder before use. Experimental tests which have been conducted in this study include theslump test, compressive test and splitting tensile strength test. CSP have the potential to be used as a concrete filler and thus the findings of this study may be applied to the construction industry. The use of CSP as a filler in concrete can help make the earth a more sustainable and greener place to live in.

  10. Analytical capability for predicting structural response of NPP concrete containments to severe loads

    International Nuclear Information System (INIS)

    Planas, J.; Guinea, G.; Trbojevic, V.M.; Marti, J.; Martinez, F.; Cortes, P.

    1989-12-01

    A survey has been conducted on the state-of-the-art of analytical techniques for predicting the structural response of concrete containment buildings under severe accident conditions. The validity of inelastic analysis is often limited by the inadequacy of the material models adopted. This is specially true in the case of materials which undergo localization phenomena in the course of the deformation process. Because of this, the Joint Research Centre at Ispra has given a high priority to the review of existing constitutive models for concrete. Such models must be able to describe concrete behaviour with and without steel reinforcement across the complete stress range, from initial elastic behaviour to and beyond the point of failure. For reinforced and prestressed concrete, segregated models (where concrete and steel are independently simulated) are preferred. A review of existing constitutive models for mass concrete has been conducted. The review focused on necessary features for describing the near-peak and post-peak stages of deformation. Special attention was dedicated to the localization of strains in tension and the post-peak softening behaviour. Existing models for representing the concrete steel bond were also reviewed. These models are still relatively simplistic and incorporate seldom a number of effects of considerable importance: sustained, dynamic and cyclic loading, environmental effects, etc. Finally, the computational procedures currently available for modelling problems involving the ultimate capacity of concrete containments have also been reviewed. This includes methodologies for modelling amongst other mass concrete, cracking procedures, bond behaviour, in existing computer codes

  11. Corrosion Measurements in Reinforced Fly Ash Concrete Containing Steel Fibres Using Strain Gauge Technique

    Directory of Open Access Journals (Sweden)

    V. M. Sounthararajan

    2013-01-01

    Full Text Available Corrosion of steel bars in concrete is a serious problem leading to phenomenal volume expansion and thereby leading to cover concrete spalling. It is well known that the reinforced concrete structures subjected to chloride attack during its service life cause these detrimental effects. The early detection of this damage potential can extend the service life of concrete. This study reports the comprehensive experimental studies conducted on the identification of corrosion mechanism in different types of reinforced concrete containing class-F fly ash and hooked steel fibres. Fly ash replaced concrete mixes were prepared with 25% and 50% fly ash containing steel fibres at 0.5%, 1.0%, and 1.5% by volume fraction. Corrosion process was investigated in an embedded steel bar (8 mm diameter reinforced in concrete by passing an impressed current in sodium chloride solution. Strain gauge attached to the rebars was monitored for electrical measurements using strain conditioner. Strain gauge readings observed during the corrosion process exhibited the volume changes of the reinforcement embedded inside the concrete. The corrosion potential of different steel fibre reinforced concrete mixes with fly ash addition showed higher resistance towards the corrosion initiation.

  12. Compressive and tensile strength for concrete containing coal bottom ash

    Science.gov (United States)

    Maliki, A. I. F. Ahmad; Shahidan, S.; Ali, N.; Ramzi Hannan, N. I. R.; Zuki, S. S. Mohd; Ibrahim, M. H. W.; Azmi, M. A. Mohammad; Rahim, M. Abdul

    2017-11-01

    The increasing demand in the construction industry will lead to the depletion of materials used in construction sites such as sand. Due to this situation, coal bottom ash (CBA) was selected as a replacement for sand. CBA is a by-product of coal combustion from power plants. CBA has particles which are angular, irregular and porous with a rough surface texture. CBA also has the appearance and particle size distribution similar to river sand. Therefore, these properties of CBA make it attractive to be used as fine aggregate replacement in concrete. The objectives of this study were to determine the properties of CBA concrete and to evaluate the optimum percentage of CBA to be used in concrete as fine aggregate replacement. The CBA was collected at Tanjung Bin power plant. The mechanical experiment (compressive and tensile strength test) was conducted on CBA concrete. Before starting the mechanical experiment, cubic and cylindrical specimens with dimensions measuring 100 × 100 × 100 mm and 150 × 300 mm were produced based on the percentage of coal bottom ash in this study which is 0% as the control specimen. Meanwhile 10%, 20%, 30%, 40%, 50%, 60%, 70%, 80%, 90% and 100% of CBA were used to replace the fine aggregates. The CBA concrete samples were cured for 7 days and 28 days respectively to maintain the rate of hydration and moisture. After the experimental work was done, it can be concluded that the optimum percentage of CBA as fine aggregate is 60% for a curing period of both 7 days and 28 days with the total compressive strength of 36.4 Mpa and 46.2 Mpa respectively. However, the optimum percentage for tensile strength is at 70% CBA for a curing period of both 7 days and 28 days with a tensile strength of 3.03 MPa and 3.63 MPa respectively.

  13. 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)

  14. The study on the mechanical characteristics of concrete of nuclear reactor containment structure

    International Nuclear Information System (INIS)

    Jung, W. S.; Kwon, K. J.; Cho, M. S.; Song, Y. C.

    2000-01-01

    Reactor containment structure of nuclear power plant designed by prestressed concrete causes time-dependent prestress loss due to the mechanical characteristics of concrete. Prestress loss strongly affects to the safety factor of structure under the circumstances of designing, construction and inspection. Thus, this study is to investigate the mechanical characteristics of reactor containment concrete structure of Yonggwang No. 5 and 6. In this study, the compressive strength, modulus of elasticity, poisson's ratio and creep test followed by ASTM code are performed to investigate the mechanical characteristics of concrete made by V type cement. Additionally, since creep causes more time-dependent prestress loss than the other, the measurement value from the creep test is compared with the results from the creep prediction equations by KSCE, JSCE, Hansen, ACI and CEB-FIP model for the effective application. Hereafter, the results of this study may enable to assist the calculation effective stress considering time-dependent prestress loss of the prestressed concrete structures

  15. Economic aspect comparison between steel plate reinforced concrete and reinforced concrete technique in reactor containment wall construction

    International Nuclear Information System (INIS)

    Yuliastuti; Sriyana

    2008-01-01

    Construction costs of nuclear power plant were high due to the construction delays, regulatory delays, redesign requirement, and difficulties in construction management. Based on US DOE (United States Department of Energy) study in 2004, there were thirteen advanced construction technologies which were potential to reduce the construction time of nuclear power plant. Among these technologies was the application of steel-plate reinforced concrete (SC) on reactor containment construction. The conventional reinforced concrete (RC) technique were built in place and require more time to remove framework since the external form is temporary. Meanwhile, the SC technique offered a more efficient way to placing concrete by using a permanent external form made of steel. The objective of this study was to calculate construction duration and economic comparison between RC and SC technique. The result of this study showed that SC technique could reduce the construction time by 60% and 29,7% cost reduced compare to the RC technique. (author)

  16. Parametric model to estimate containment loads following an ex-vessel steam spike

    International Nuclear Information System (INIS)

    Lopez, R.; Hernandez, J.; Huerta, A.

    1998-01-01

    This paper describes the use of a relatively simple parametric model to estimate containment loads following an ex-vessel steam spike. The study was motivated because several PSAs have identified containment loads accompanying reactor vessel failures as a major contributor to early containment failure. The paper includes a detailed description of the simple but physically sound parametric model which was adopted to estimate containment loads following a steam spike into the reactor cavity. (author)

  17. Development and application of a material law for steel-fibre-reinforced concrete with regard to its use for pre-stressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Schimmelpfennig, K.; Borgerhoff, M.

    1995-01-01

    On the basis of the evaluation of many publications on the mechanical behaviour of steel fibre reinforced concrete (SFRC) and on the results of experiments using an SFRC especially developed for pre-stressed concrete reactor vessels (PCRVs), a material law for SFRC including general multiaxial stress conditions has been developed. From fibre pull-out tests described in the literature and by use of the experimental results, relations describing the capable tensile stress in SFRC after cracking, as a function of crack width, have been derived. There is a significant increase in the biaxial compressive strength of SFRC compared with plain concrete. The improved behaviour under multiaxial stress conditions, with one of the principal stresses being tensile, is outlined in comparison with different formulations of failure envelopes of plain concrete. For the purpose of verifying the material law implemented in the computer program used, analyses have been carried out for experiments with SFRC beams. After some modification concerning the shear behaviour, load-displacement curves and realistic crack propagations which correspond well have been obtained. In the stand-tube area in the centre of a PCRV top cap the use of SFRC is advantageous because of the difficulties concerning the arrangement of reinforcement in the concrete between the tubes. (orig.)

  18. Stress analysis and review of a prestressed concrete reactor pressure vessel of a HTR 500-type

    International Nuclear Information System (INIS)

    Wang, T.J.; Altes, J.

    1988-12-01

    The main aim of this first step of analysis is to test the feasibility of the SMART-code for a complete calculation of PCRV's and to establish the control programs for the finite element analysis. Curved triangular, quadrilateral and membrane shell elements are used. The incremental model in the form of tangential stress-strain law has been chosen as the constitutive model for the concrete. The three parameter failure envelope is used as the failure model of the concrete. For the numerical solution the incremental initial iteration method with constant stiffness is utilized. The creep strain is treated as a liner functional of the stress history and dependent on temperature, humidity and aging. The calculation of the creep behaviour is carried out up to 7 years of operation using the model of SEKI and KAWASUMI. In this model the influence of temperature, humidity and the interaction between them is fully considered. The effects of interaction between temperature and creep with and without humidity's influence are studied and some interesting results are presented. The total creep curves vs time are gained, the deformations of nodal points after 7 years are 1.8 - 5.5 times larger then those of the initial elastic deformation after the first loading. Under the action of prestressing along most parts of the PCRV and under the service condition the main part of the PCRV are in compression. Due to increasing the loading over the operating pressure some parts are cracked and the material behaves nonlinearly. At a loading value of 3.25 times the operating pressure the whole transverse section is fully cracked. For the stage of prestressing, design operating pressure and design limit pressure the vessel behaves elastically. The global safety factory is 1.5 times larger than the design value of 2.25 that shows the conservative design. The analysis method and computer codes, which are used in this review, are confirmed efficiently. (orig./HP)

  19. Monitoring the Durability Performance of Concrete in Nuclear Waste Containment. Technical Progress Report No. 3

    International Nuclear Information System (INIS)

    Ulm, Franz-Josef

    2000-01-01

    OAK-B135 Monitoring the Durability Performance of Concrete in Nuclear Waste Containment. Technical Progress Report No. 3(NOTE: Part II A item 1 indicates ''PAPER'', but a report is attached electronically)

  20. Characterization of Concrete Mixes Containing Phase Change Materials

    Science.gov (United States)

    Paksoy, H.; Kardas, G.; Konuklu, Y.; Cellat, K.; Tezcan, F.

    2017-10-01

    Phase change materials (PCM) can be used in passive building applications to achieve near zero energy building goals. For this purpose PCM can be added in building structures and materials in different forms. Direct incorporation, form stabilization and microencapsulation are different forms used for PCM integration in building materials. In addition to thermal properties of PCM itself, there are several other criteria that need to be fulfilled for the PCM enhanced building materials. Mechanical properties, corrosive effects, morphology and thermal buffering have to be determined for reliable and long-term applications in buildings. This paper aims to give an overview of characterization methods used to determine these properties in PCM added fresh concrete mixes. Thermal, compressive strength, corrosion, and microscopic test results for concrete mixes with PCM are discussed.

  1. Acoustic Resonance Characteristics of Rock and Concrete Containing Fractures

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Seiji [Univ. of California, Berkeley, CA (United States)

    1998-08-01

    In recent years, acoustic resonance has drawn great attention as a quantitative tool for characterizing properties of materials and detecting defects in both engineering and geological materials. In quasi-brittle materials such as rock and concrete, inherent fractures have a significant influence on their mechanical and hydraulic properties. Most of these fractures are partially open, providing internal boundaries that are visible to propagating seismic waves. Acoustic resonance occurs as a result of constructive and destructive interferences of propagating waves. Therefore the geometrical and mechanical properties of the fracture are also interrogated by the acoustic resonance characteristics of materials. The objective of this dissertation is to understand the acoustic resonance characteristics of fractured rock and concrete.

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  3. Computation of shrinkage stresses in prestressed concrete containments

    International Nuclear Information System (INIS)

    Wu, R.F.; Ouyang, H.

    1989-01-01

    According to a survey, surface cracking on PCRVs and PCCs under the investigations is confined to drying shrinkage and thermal strain effects and no instances of structurally significant cracking was been found. In this paper, the authors use FEM to compute humidity distribution in drying concrete and shrinkage stresses by internal restraint. Since PCC is built segment by segment in several years, a computational model taking into account construction sequence is presented and shrinkage stresses by external restraints are calculated with the model

  4. A method for three-dimensional structural analysis of reinforced concrete containment

    International Nuclear Information System (INIS)

    Kulak, R.F.; Fiala, C.

    1989-01-01

    A finite element method designed to assist reactor safety analysts in the three-dimensional numerical simulation of reinforced concrete containments to normal and off-normal mechanical loadings is presented. The development of a lined reinforced concrete plate element is described in detail, and the implementation of an empirical transverse shear failure criteria is discussed. The method is applied to the analysis of a 1/6th scale reinforced concrete containment model subjected to static internal pressurization. 11 refs., 14 figs., 1 tab

  5. The durability of concrete containing recycled tyres as a partial replacement of fine aggregate

    Science.gov (United States)

    Syamir Senin, Mohamad; Shahidan, Shahiron; Syazani Leman, Alif; Othman, Nurulain; Shamsuddin, Shamrul-mar; Ibrahim, M. H. W.; Zuki, S. S. Mohd

    2017-11-01

    Nowadays, uncontrolled disposal of waste materials such as tyres can affect the environment. Therefore, careful management of waste disposal must be done in order to conserve the environment. Waste tyres can be use as a replacement for both fine aggregate and coarse aggregate in the production of concrete. This research was conducted to assess the durability of concrete containing recycled tyres which have been crushed into fine fragments to replace fine aggregate in the concrete mix. This study presents an overview of the use of waste rubber as a partial replacement of natural fine aggregate in a concrete mix. 36 concrete cubes measuring 100mm × 100mm × 100mm and 12 concrete cubes measuring 150mm × 150mm × 150mm were prepared and added with different percentages of rubber from recycled tyres (0%, 3%, 5% and 7%) as fine aggregate replacement. The results obtained show that the replacement of fine aggregate with 7% of rubber recorded a compressive strength of 43.7MPa while the addition of 3% of rubber in the concrete sample recorded a high compressive strength of 50.8MPa. This shows that there is a decrease in the strength and workability of concrete as the amount of rubber used a replacement for fine aggregate in concrete increases. On the other hand, the water absorption test indicated that concrete which contains rubber has better water absorption ability. In this study, 3% of rubber was found to be the optimal percentage as a partial replacement for fine aggregate in the production of concrete.

  6. Serviceability design load factors and reliability assessments for reinforced concrete containment structures

    International Nuclear Information System (INIS)

    Han Bong Koo

    1998-01-01

    A reinforced concrete nuclear power plant containment structure is subjected to various random static and stochastic loads during its lifetime. Since these loads involve inherent randomness and other uncertainties, an appropriate probabilistic model for each load must be established in order to perform reliability analysis. The current ASME code for reinforced concrete containment structures are not based on probability concepts. The stochastic nature of natural hazard or accidental loads and the variations of material properties require a probabilistic approach for a rational assessment of structural safety and performance. The paper develops probability-based load factors for the limit state design of reinforced concrete containment structures. The purpose of constructing reinforced concrete containment structure is to protect against radioactive release, and so the use of a serviceability limit state against crack failure that can cause the emission of radioactive materials is suggested as a critical limit state for reinforced concrete containment structures. Load factors for the design of reinforced concrete containment structures are proposed and carried out the reliability assessments. (orig.)

  7. Nuclear reactors sited deep underground in steel containment vessels

    Energy Technology Data Exchange (ETDEWEB)

    Bourque, Robert [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

    2006-07-01

    Although nuclear power plants are certainly very safe, they are not perceived as safe by the general populace. Also, there are concerns about overland transport of spent fuel rods and other irradiated components. It is hereby proposed that the nuclear components of nuclear power plants be placed in deep underground steel vessels with secondary coolant fed from them to turbines at or near the surface. All irradiated components, including spent fuel, would remain in the chamber indefinitely. This general concept was suggested by the late Edward Teller, generated some activity 20-25 years ago and appears to be recently reviving in interest. Previous work dealt with issues of geologic stability of underground, possibly reinforced, caverns. This paper presents another approach that makes siting independent of geology by placing the reactor components in a robust steel vessel capable of resisting full overburden pressure as well as pressures resulting from accident scenarios. Structural analysis of the two vessel concepts and approximate estimated costs are presented. This work clears the way for the extensive discussions required to evaluate the advantages of this concept. (author)

  8. Prediction of Splitting Tensile Strength of Concrete Containing Zeolite and Diatomite by ANN

    Directory of Open Access Journals (Sweden)

    E. Gülbandılar

    2017-01-01

    Full Text Available This study was designed to investigate with two different artificial neural network (ANN prediction model for the behavior of concrete containing zeolite and diatomite. For purpose of constructing this model, 7 different mixes with 63 specimens of the 28, 56 and 90 days splitting tensile strength experimental results of concrete containing zeolite, diatomite, both zeolite and diatomite used in training and testing for ANN systems was gathered from the tests. The data used in the ANN models are arranged in a format of seven input parameters that cover the age of samples, Portland cement, zeolite, diatomite, aggregate, water and hyper plasticizer and an output parameter which is splitting tensile strength of concrete. In the model, the training and testing results have shown that two different ANN systems have strong potential as a feasible tool for predicting 28, 56 and 90 days the splitting tensile strength of concrete containing zeolite and diatomite.

  9. Ultimate analysis of PWR prestressed concrete containment subjected to internal pressure

    International Nuclear Information System (INIS)

    Hu, H.-T.; Lin, Y.-H.

    2006-01-01

    Numerical analyses are carried out by using the ABAQUS finite element program to predict the ultimate pressure capacity and the failure mode of the PWR prestressed concrete containment at Maanshan nuclear power plant. Material nonlinearity such as concrete cracking, tension stiffening, shear retention, concrete plasticity, yielding of prestressing tendon, yielding of steel reinforcing bar and degradation of material properties due to high temperature are all simulated with proper constitutive models. Geometric nonlinearity due to finite deformation has also been considered. The results of the analysis show that when the prestressed concrete containment fails, extensive cracks take place at the apex of the dome, the junction of the dome and cylinder, and the bottom of the cylinder connecting to the base slab. In addition, the ultimate pressure capacity of the containment is higher than the design pressure by 86%

  10. Design of reinforced concrete containment structures for thermal gradients effects

    International Nuclear Information System (INIS)

    Bhat, P.D.; Vecchio, F.

    1983-01-01

    The need for more accurate prediction of structural behaviour, particularly under extreme load conditions, has made the consideration of thermal gradient effects and increasingly important part of the design of reinforced concrete structures for nuclear applications. While the thermal effects phenomenon itself has been qualitatively well understood, the analytical complications involved in theoretical analysis have made it necessary to resort to major simplifications for practical design applications. A number of methods utilizing different variations in approach have been developed and are in use today, including one by Ontario Hydro which uses an empirical relationship for determining an effective moment of inertia for cracked members. (orig./WL)

  11. An international data base of nuclear concrete containment ageing

    International Nuclear Information System (INIS)

    Seni, C.; Ianko, L.

    1994-01-01

    The ageing of nuclear structures is of special interest because of the extended service life expected of these structures, and the potential impact of their deterioration on safety and reliability. Although there are databases about concrete, they address properties in general, not performance. In 1992, the IAEA, in collaboration with AECL, set out to create a new database that would fill the gap. Functional ageing, i.e. deterioration of leak-tightness, was to be included, not just structural ageing, i.e. deterioration of load-bearing capacity. This paper outlines the project of creating the database

  12. Flexural Behaviour Of Reinforced Concrete Beams Containing Expanded Glass As Lightweight Aggregates

    Directory of Open Access Journals (Sweden)

    Khatib Jamal

    2015-12-01

    Full Text Available The flexural properties of reinforced concrete beams containing expanded glass as a partial fine aggregate (sand replacement are investigated. Four concrete mixes were employed to conduct this study. The fine aggregate was replaced with 0%, 25%, 50% and 100% (by volume expanded glass. The results suggest that the incorporation of 50% expanded glass increased the workability of the concrete. The compressive strength was decreasing linearly with the increasing amount of expanded glass. The ductility of the concrete beam significantly improved with the incorporation of the expanded glass. However, the load-carrying capacity of the beam and load at which the first crack occurs was reduced. It was concluded that the inclusion of expanded glass in structural concrete applications is feasible.

  13. A realistic structural analysis of the integrity of the liner of reinforced and prestressed concrete containments

    International Nuclear Information System (INIS)

    Buchhardt, F.; Brandl, P.

    1979-01-01

    The BWR Gundremmingen II is the first German nuclear power plant with a concrete containment having a thin steel plate liner directly attached to the interior concrete surface to provide an air-tight seal. Due to this monolithic way of anchorage a bonded system of concrete and metal liner membrane is obtained so that the same deformations of the loading or strain conditions are induced to the very stiff concrete hull as well as to the liner. Because of the complex structural behaviour of the bonded system the evaluation is carried out by the finite element method. The overall system is decoupled in several steps. Due to its considerable stiffness the concrete structure can be regarded as the liner supporting basis. The liner system itself might be subdivided into perfect and imperfect sections discretized by plain or curved elements which are supported by point-wise spring elements representing the stud anchors. (orig.)

  14. Carbonation of ternary cementitious concrete systems containing fly ash and silica fume

    Directory of Open Access Journals (Sweden)

    Eehab Ahmed Badreldin Khalil

    2015-04-01

    Full Text Available Carbonation is quite a complex physical negative effect phenomenon on concrete especially in the ones containing ternary blends of Portland Cement, fly ash, and silica fume. Nine selected concrete mixtures were prepared with various water to cementitious materials’ ratios and various cementitious contents. The concrete mixtures were adapted in such a way to have the same workability and air content. The fresh concrete properties were kept near identical in slump, air content, and unit weight. The variation was in the hardened concrete mechanical properties of compression and tension strength. The carbonation phenomenon was studied for these mixes showing at which mixes of ternary cementitious content heavy carbonation attacks maybe produced. The main components of such mixes that do affect the carbonation process with time were presented.

  15. Energy release and its containment within thin-walled, backed vessels

    International Nuclear Information System (INIS)

    Chambers, D.I.

    1983-01-01

    The problem adressed is the containment of a sudden release of energy of a magnitude up to 4 x 10 11 joules in a reusable vessel. The design process began by formulating dynamic models for both the input to such a vessel and the vessel itself and using these models to generate a general response. Modifications to the input and a more specific response are discussed. Computer codes used in calculations are described and listed

  16. Using An Adapter To Perform The Chalfant-Style Containment Vessel Periodic Maintenance Leak Rate Test

    International Nuclear Information System (INIS)

    Loftin, B.; Abramczyk, G.; Trapp, D.

    2011-01-01

    Recently the Packaging Technology and Pressurized Systems (PT and PS) organization at the Savannah River National Laboratory was asked to develop an adapter for performing the leak-rate test of a Chalfant-style containment vessel. The PT and PS organization collaborated with designers at the Department of Energy's Pantex Plant to develop the adapter currently in use for performing the leak-rate testing on the containment vessels. This paper will give the history of leak-rate testing of the Chalfant-style containment vessels, discuss the design concept for the adapter, give an overview of the design, and will present results of the testing done using the adapter.

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

    International Nuclear Information System (INIS)

    Raske, D.T.; Wang, Z.

    1992-01-01

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

  18. Sulphuric Acid Resistant of Self Compacted Geopolymer Concrete Containing Slag and Ceramic Waste

    Directory of Open Access Journals (Sweden)

    Shafiq I.

    2017-01-01

    Full Text Available Malaysia is a one of the developing countries where the constructions of infrastructure is still ongoing, resulting in a high demand for concrete. In order to gain sustainability factors in the innovations for producing concrete, geopolymer concrete containing granulated blast-furnace slag and ceramics was selected as a cement replacement in concrete for this study. Since Malaysia had many ceramic productions and uses, the increment of the ceramic waste will also be high. Thus, a new idea to reuse this waste in construction materials have been tested by doing research on this waste. Furthermore, a previous research stated that Ordinary Portland Cement concrete has a lower durability compared to the geopolymer concrete. Geopolymer binders have been reported as being acid resistant and thus are a promising and alternative binder for sewer pipe manufacture. Lack of study regarding the durability of the geopolymer self-compacting concrete was also one of the problems. The waste will be undergoing a few processes in the laboratory in order to get it in the best form before undergoing the next process as a binder in geopolymer concrete. This research is very significant in order to apply the concept of sustainability in the construction field. In addition, the impact of this geopolymer binder is that it emits up to nine times less CO2 than Portland Cement.

  19. Nonlinear failure analysis of a reinforced concrete containment under internal pressure

    International Nuclear Information System (INIS)

    Sharma, S.; Wang, Y.K.; Reich, M.

    1984-01-01

    A detailed nonlinear finite element model is used to investigate the failure response of the Indian Point containment building under severe accident pressures. Refined material models are used to describe the complex stress-strain behavior of the liner and rebar steels, the plain concrete and the reinforced concrete. Structural geometry of the containment is idealized by eight layers of axisymmetric finite elements through the wall thickness in order to closely model the actual placement of the rebars. Soil stiffness under the containment base mat is modeled by a series of nonlinear spring elements. Numerical results presented in the paper describe cracking and plastic deformation (in compression) of the concrete, yielding of the liner and rebar steels and eventual loss of the load carrying capacity of the containment. The results are compared with available data from the previous studies for this containment. 8 references, 9 figures

  20. Overview of concrete containment design practice in the U.S.A

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    1976-01-01

    This paper presents a historical summary of the engineering practices and their evolution applied to the design of concrete containment structures in the U.S.A. during the period 1965 to 1974. It reviews the broad spectrum of concrete containment designs developed for the three major Nuclear Steam Supply Systems, Pressurized Water Reactor, Boiling Water Reactor and High Temperature Gas Reactor employed or planned in the U.S.A. during this period. The development of deformed rebar and one way prestress as well as fully prestressed reinforced concrete containment is discussed. Particular attention is paid to base mat-containment shell joint design details as well as the design of reinforcement around large penetrations and those penetrations subject to large pipe thrust loads. In addition to the historical summary, current trends in containment design are identified and projections of future developments are presented. Finally, potential innovations such as plastic liners are discussed. (author)

  1. The measured contribution of whipping and springing on the fatigue and extreme loading of container vessels

    Science.gov (United States)

    Storhaug, Gaute

    2014-12-01

    Whipping/springing research started in the 50'ies. In the 60'ies inland water vessels design rules became stricter due to whipping/springing. The research during the 70-90'ies may be regarded as academic. In 2000 a large ore carrier was strengthened due to severe cracking from North Atlantic operation, and whipping/springing contributed to half of the fatigue damage. Measurement campaigns on blunt and slender vessels were initiated. A few blunt ships were designed to account for whipping/springing. Based on the measurements, the focus shifted from fatigue to extreme loading. In 2005 model tests of a 4,400 TEU container vessel included extreme whipping scenarios. In 2007 the 4400 TEU vessel MSC Napoli broke in two under similar conditions. In 2009 model tests of an 8,600 TEU container vessel container vessel included extreme whipping scenarios. In 2013 the 8,100 TEU vessel MOL COMFORT broke in two under similar conditions. Several classification societies have published voluntary guidelines, which have been used to include whipping/springing in the design of several container vessels. This paper covers results from model tests and full scale measurements used as background for the DNV Legacy guideline. Uncertainties are discussed and recommendations are given in order to obtain useful data. Whipping/springing is no longer academic.

  2. Sand Cement Brick Containing Recycled Concrete Aggregate as Fine-Aggregate Replacement

    Directory of Open Access Journals (Sweden)

    Sheikh Khalid Faisal

    2017-01-01

    Full Text Available Nowadays, the usage amount of the concrete is increasing drastically. The construction industry is a huge consumer of natural consumer. It is also producing the huge wastage products. The usage of concrete has been charged to be not environmentally friendly due to depletion of reserve natural resources, high energy consumption and disposal issues. The conservation of natural resources and reduction of disposal site by reuse and recycling waste material was interest possibilites. The aim of this study is to determine the physical and mechanical properties of sand cement brick containing recycled concrete aggregate and to determine the optimum mix ratio containing recycled concrete aggregate. An experiment done by comparing the result of control specimen using 100% natural sand with recycled concrete aggregate replacement specimen by weight for 55%, 65%, and 75%. The sample was tested under density, compressive strength, flexural strength and water absorption to study the effect of using recycled concrete aggregate on the physical and mechanical properties of bricks. The result shows that the replacement of natural sand by recycled concrete aggregate at the level of 55% provide the highest compressive and flexural strength compared to other percentage and control specimen. However, if the replacement higher than 55%, the strength of brick was decreased for compressive and flexural strength, respectively. The relationship of compressive-flexural strength is determined from statistical analysis and the predicted result can be obtained by using equation ff,RCA = 0.5375 (fc0.3272.

  3. Prestressed concrete reactor vessel for the HHT-670 MW(e) demonstration plant. Pt.2. Three-dimensional analysis of the temperature and stress fields in a HHT vessel, including effects of the thermal creep

    International Nuclear Information System (INIS)

    Rodriguez, C.; Rebora, B.

    1979-01-01

    The thermal rheological calculation of the prestressed concrete reactor vessel for the HHT-670 MW(e) Demonstration Plant is presented in the paper. The main aim of this calculation is to evaluate the effects of the elevated temperature and various loads on the liner as well as on the hot concrete

  4. Heissdampfreaktor (HDR) steel-containment-vessel and floodwater-storage-tank structural-dynamics tests

    International Nuclear Information System (INIS)

    Arendts, J.G.

    1982-01-01

    Inertance (vibration) testing of two significant vessels at the Heissdampfreaktor (HDR) facility, located near Kahl, West Germany, was recently completed. Transfer functions were obtained for determination of the modal properties (frequencies, mode shapes and damping) of the vessels using two different test methods for comparative purposes. One of the vessels tested was the steel containment vessel (SCV). The SCV is approximately 180 feet high and 65 feet in diameter with a 1.2-inch wall thickness. The other vessel, called the floodwater storage tank (FWST), is a vertically standing vessel approximately 40 feet high and 10 feet in diameter with a 1/2-inch wall thickness. The FWST support skirt is square (in plan views) with its corners intersecting the ellipsoidal bottom head near the knuckle region

  5. Development and investigation of the prestressed reinforced concrete vessels for the water cooled reactors in the FRG

    International Nuclear Information System (INIS)

    Medovikov, A.I.; Bogopol'skij, V.G.; Nikolaev, Yu.B.; Konevskij, V.N.

    1980-01-01

    An analysis of calculation results for characteristics of stress-strained state of reactor vessel made of prestressed reinforced concrete is presented. Experimental data obtained during the investigation into a model of reactor vessel top cover are given. Thermal shielding system both for boiling water and pressurized-water reactors has been considered and its working capacity has been evaluated. An analysis of experimental data show correctness of the method assumed for calculation of the reactor top cover which permits to exactly determine its stressed-strained state as well as the nature of crack propagation in the vessel and the structure supporting power. Ceramics is suggested to be used as a heat-insulating material

  6. Study on effective prestressing effects on concrete containment under the design-basis pressure condition

    International Nuclear Information System (INIS)

    Sun Feng; Pan Rong; Wang Lu; Mao Huan; Yang Yu

    2013-01-01

    Prestressing technology is widely used in nuclear power plant containment building, and the durability of containment structure is affected directly by the distribution and loss of prestressing value under design-basis pressure. Containment structure and the distribution of prestressing system are introduced briefly. Furthermore, the calculating process of horizontal prestressing bunch loss near the equipment hatch hole is put forward in details, and the containment structure prestressing loss when 5-year pressure test is obtained. Based above analysis, the finite element model of the prestressed concrete containment structure is built by using ANSYS code, the prestressing effect on concrete containment is analysed. The results show that most of the design pressure is bore by the prestressing system under the design-basis pressure, so the containment structure is safe. These conclusions are consistent with prestressing containment system design concepts, which can provide reference to the engineering staff. (authors)

  7. 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

  8. Examination and testing requirements for concrete containment structures for CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-07-01

    This Standard provides the examination and testing requirements that will apply to the work of any organization participating in the construction, installation, and fabrication of parts or components of concrete containment structures, or both, that are defined as class containment. 2 tabs.

  9. Examination and testing requirements for concrete containment structures for CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1993-07-01

    This Standard provides the examination and testing requirements that will apply to the work of any organization participating in the construction, installation, and fabrication of parts or components of concrete containment structures, or both, that are defined as class containment. 2 tabs

  10. Leaching studies of heavy concrete material for nuclear fuel waste immobilization containers

    International Nuclear Information System (INIS)

    Onofrei, M.; Raine, D.; Brown, L.; Hooton, R.D.

    1989-08-01

    The leaching behaviour of a high-density concrete was studied as part of a program to evaluate its potential use as a container material for nuclear fuel waste under conditions of deep geologic disposal. Samples of concrete material were leached in deionized distilled water, Standard Canadian Shield Saline Solution (SCSSS), SCSSS plus 20% Na-bentonite, and SCSSS plus granite and 20% Na-bentonite under static conditions at 100 degrees celsius for periods up to 365 days. The results of these leaching experiments suggest that the stability of concrete depends on the possible internal structural changes due to hydration reactions of unhydrated components, leading to the formation of C-S-H gel plus portlandite (Ca(OH) 2 ). The factors controlling the concrete leaching process were the composition of the leachant and the concentration of elements in solution capable of forming precipitates on the concrete surface, e.g., silicon, Mg 2+ and Ca 2+ . The main effect observed during leaching was an increase in groundwater pH (from 7 to 9). However, the addition of Na-bentonite suppressed the normal tendency of the pH of the groundwater in contact with concrete to rise rapidly. It was shown that the solution concentration of elements released from the concrete, particularly potassium, increased in the presence of Na-bentonite

  11. Corrosion on reinforced concrete structures. An application for the intermediate level radioactive waste container

    International Nuclear Information System (INIS)

    Arva, Alejandro; Alvarez, Marta G.; Duffo, Gustavo S.

    2003-01-01

    The behavior of steel reinforcement bars (rebars) for a high performance reinforced concrete made of sulfate resistant portland cement was evaluated from the rebars corrosion point of view. The results from the present work will be used to evaluate the materials properties to be used in the construction of the intermediate level radioactive waste disposal containers. The study is carried out evaluating the incidence of chloride and sulfate ions, as well as, concrete carbonation in the rebar corrosion process. The electrochemical parameters that characterize the corrosion process (corrosion potential [E corr ], polarisation resistance [Rp] and concrete electrical resistivity [ρ]) were monitored on specially designed reinforced concrete specimens. The results up to date (about 1000 days of exposure) reveal that the concrete under study provides to the steel reinforcement bars of a passive state against corrosion under the test conditions. An increasing tendency as a function of time of ρ is observed that corroborates the continuous curing process of concrete. The chloride and carbonation diffusion coefficients were also determined, and their values are comparable with those of high quality concrete. (author)

  12. Microstructure of ultra high performance concrete containing lithium slag.

    Science.gov (United States)

    He, Zhi-Hai; Du, Shi-Gui; Chen, Deng

    2018-04-03

    Lithium slag (LS) is discharged as a byproduct in the process of the lithium carbonate, and it is very urgent to explore an efficient way to recycle LS in order to protect the environments and save resources. Many available supplementary cementitious materials for partial replacement of cement and/or silica fume (SF) can be used to prepare ultra high performance concrete (UHPC). The effect of LS to replace SF partially by weight used as a supplementary cementitious material (0%, 5%, 10% and 15% of binder) on the compressive strengths and microstructure evolution of UHPC has experimentally been studied by multi-techniques including mercury intrusion porosimetry, scanning electron microscope and nanoindentation technique. The results show that the use of LS degrades the microstructure of UHPC at early ages, and however, the use of LS with the appropriate content improves microstructure of UHPC at later ages. The hydration products of UHPC are mainly dominated by ultra-high density calcium-silicate-hydrate (UHD C-S-H) and interfacial transition zone (ITZ) in UHPC has similar compact microstructure with the matrix. The use of LS improves the hydration degree of UHPC and increases the elastic modulus of ITZ in UHPC. LS is a promising substitute for SF for preparation UHPC. Copyright © 2018 Elsevier B.V. All rights reserved.

  13. Analysis of radioactivity increase of rad waste filled in fibre-reinforced concrete container regarding external exposure of workers

    International Nuclear Information System (INIS)

    Baratova, D.; Hrncir, T.; Necas, V.

    2012-01-01

    The paper deals with the assessment of the external radiation exposure of workers performing the individual tasks associated with disposal of the fibre-reinforced concrete containers in the National Radioactive Waste Repository in Mochovce. Models for fibre-reinforced concrete containers with maximum activity allowable for transport and for fibre-reinforced concrete containers contained radionuclides at the common level of activity concentration were created in order to analyze the option of fibre-reinforced concrete containers radioactivity increase. Calculations of individual effective doses have been carried out for three workers who work in the control area of the waste disposal facility dosimetrist, assistant and crane worker. (Authors)

  14. Dynamic Properties of Container Vessel with Low Metacentric Height

    DEFF Research Database (Denmark)

    Blanke, M.; Jensen, A.G.

    1997-01-01

    between roll and lateral motions. This was changed with the construction of a unique four degrees of freedom roll planar motion mechanism (RPMM) at the Danish Maritime Institute. The paper presents complete nonlinear models for a container ship obtained with this facility. Model scale predictions...

  15. Observations on analysis, testing and failure of prestressed concrete containments

    International Nuclear Information System (INIS)

    Murray, D.W.

    1984-01-01

    The paper reviews the mechanics which indicate that a bursting failure with large energy release is the failure mechanism to be expected from ductile lined containment structures pressurized to failure. It reviews a study which shows that, because of leakage, this is not the case for unlined prestressed containments. It argues that current practice, since it does not specifically address the bursting failure problem for lined prestressed containments, is inadequate to ensure that this type of failure could not occur. It concludes that, in view of the inadequacy of the current state-of-the-art to predict leakage from lined structures, the logical remedy is to eliminate all possibility of bursting failure by making provision for venting of containments. (orig.)

  16. A Study on the Evaluation of Field Application of High-Fluidity Concrete Containing High Volume Fly Ash

    Directory of Open Access Journals (Sweden)

    Yun-Wang Choi

    2015-01-01

    Full Text Available In the recent concrete industry, high-fluidity concrete is being widely used for the pouring of dense reinforced concrete. Normally, in the case of high-fluidity concrete, it includes high binder contents, so it is necessary to replace part of the cement through admixtures such as fly ash to procure economic feasibility and durability. This study shows the mechanical properties and field applicability of high-fluidity concrete using mass of fly ash as alternative materials of cement. The high-fluidity concrete mixed with 50% fly ash was measured to manufacture concrete that applies low water/binder ratio to measure the mechanical characteristics as compressive strength and elastic modulus. Also, in order to evaluate the field applicability, high-fluidity concrete containing high volume fly ash was evaluated for fluidity, compressive strength, heat of hydration, and drying shrinkage of concrete.

  17. Optimization and influence of parameter affecting the compressive strength of geopolymer concrete containing recycled concrete aggregate: using full factorial design approach

    Science.gov (United States)

    Krishnan, Thulasirajan; Purushothaman, Revathi

    2017-07-01

    There are several parameters that influence the properties of geopolymer concrete, which contains recycled concrete aggregate as the coarse aggregate. In the present study, the vital parameters affecting the compressive strength of geopolymer concrete containing recycled concrete aggregate are analyzedby varying four parameters with two levels using full factorial design in statistical software Minitab® 17. The objective of the present work is to gain an idea on the optimization, main parameter effects, their interactions and the predicted response of the model generated using factorial design. The parameters such as molarity of sodium hydroxide (8M and 12M), curing time (6hrs and 24 hrs), curing temperature (60°C and 90°C) and percentage of recycled concrete aggregate (0% and 100%) are considered. The results show that the curing time, molarity of sodium hydroxide and curing temperature were the orderly significant parameters and the percentage of Recycled concrete aggregate (RCA) was statistically insignificant in the production of geopolymer concrete. Thus, it may be noticeable that the RCA content had negligible effect on the compressive strength of geopolymer concrete. The expected responses from the generated model showed a satisfactory and rational agreement to the experimental data with the R2 value of 97.70%. Thus, geopolymer concrete comprising recycled concrete aggregate can solve the major social and environmental concerns such as the depletion of the naturally available aggregate sources and disposal of construction and demolition waste into the landfill.

  18. Concrete

    DEFF Research Database (Denmark)

    2015-01-01

    Concrete is a component of coherent transition between a concrete base and a wooden construction. The structure is based on a quantity of investigations of the design possibilities that arise when combining digital fabrication tools and material capacities. Through tangible experiments the project...... specific for this to happen. And the knowledge and intention behind the drawing becomes specialised through the understanding of the fabrication processes and their affect on the materials.The structure Concrete is a result of a multi-angled kerf series in ash wood and a concrete base. The ash wood is cut...... using a 5-axis CNC router with a thin saw blade attached. The programming of the machining results in variations of kerfs that lets the ash wood twist into unique shapes.The shapes of the revolving ash ribbons continue into the concrete creating a cohesive shape. The form for the concrete itself is made...

  19. The Aesthetical quality of SSA-containing mortar and concrete

    DEFF Research Database (Denmark)

    Kappel, Annemette; Kirkelund, Gunvor Marie; Ottosen, Lisbeth M.

    2014-01-01

    that gives a characteristic red colour. The process of grinding SSA has shown to improve the compressive strength of SSA- containing mortar (Donatello et al. 2010). Thus, in this study SSA was grinded in 6 different intervals ranging from 0 – 10 min, and then added to the mortar mix replacing 20% of cement....... The experiment revealed that the colour of the SSA-containing mortar intensified as the time interval of the grinding process increased. Each of the 6 steps within the time interval provided an additional colour tone and generated a colour scale consisting of mortar samples ranging from greyish to a more...

  20. Specific problems concerning aircraft impact on nuclear containment vessels

    International Nuclear Information System (INIS)

    Fuzier, J.P.; Cheyrezy, M.H.; Dufour, C.J.

    1977-01-01

    Due to the high population density, in Belgium PWR power plants are designed against aircraft impacts (BOEING 707 crashing at 360 km per hour and STARFIGHTER F 104 G crashing at 540 km per hour). A double wall is used for the containment shield. The lack of relevant data and specifications for such a loading on the non-prestressed external wall led the authors to determine the suitable safety criteria, the most appropriate materials to be used and the corresponding limit state design through dynamic and plastic analysis. (Auth.)

  1. A remote inspection system for use inside reactor containment vessels

    International Nuclear Information System (INIS)

    Aoki, Toshihiko; Kashiwai, Jun-ichi; Yamamoto, Ikuo; Fukada, Koichi; Yamanaka, Yoshinobu.

    1985-01-01

    The harsh environment in the reactor-containment vesels of pressurized-water reactor nuclear-power plants precludes the possibility of direct circuit inspection; a remote-inspection system is essential. A robot for performing this task must not only be able to withstand the harsh conditions but must also be small and maneuverable enough to function effectively within complex and confined spaces. The article describes a monorail-type remote-inspection robot developed by Mitsubishi Electric to meet these needs, which is now under trial production and testing. (author)

  2. Installation method for the steel container and vessel of the nuclear heating reactor

    International Nuclear Information System (INIS)

    Chen Liying; Guo Jilin; Liu Wei

    2000-01-01

    The Nuclear Heating Reactor (NHR) has the advantages of inherent safety and better economics, integrated arrangement, full power natural circulation and dual vessel structure. However, the large thin container presents a new and difficult problem. The characteristics of the dual vessel installation method are analyzed with system engineering theory. Since there is no foreign or domestic experience, a new method was developed for the dual vessel installation for the 5 MW NHR. The result shows that the installation method is safe and reliable. The research on the dual vessel installation method has important significance for the design, manufacture and installation of the NHR dual vessel, as well as the industrialization and standardization of the NHR

  3. Axisymmetric analysis of a 1:6-scale reinforced concrete containment building using a distributed cracking model for the concrete

    International Nuclear Information System (INIS)

    Weatherby, J.R.

    1987-09-01

    Results of axisymmetric structural analyses of a 1:6 scale model of a reinforced concrete nuclear containment building are presented. Both a finite element shell analysis and a simplified membrane analysis were made to predict the structural response and ultimate pressure capacity of the model. Analytical results indicate that the model will fail at an internal pressure of 187 psig when the stress level in the hoop reinforcement at the midsection of the cylinder exceeds the ultimate strength of the bar splices. 5 refs., 34 figs., 6 tabs

  4. Strength and deformational characteristics of three-way reinforced concrete containment models subjected to lateral forces

    International Nuclear Information System (INIS)

    Aoyagi, Y.; Yamada, K.; Takahashi, T.

    1981-01-01

    With a view to investigating the earthquake resistance characteristics of reinforced concrete containments two cylindrical models with three-way system of bars were made and loaded laterally up to failure combined with or without internal pressures, simulating the conditions in which containments were subjected to earthquake forces at a simultaneous LOCA or at normal operation. The main conclusions obtained withing the limit of the experiments are as follows. (1) Stresses in reinforcements in three-way reinforced concrete plate elements can reasonably be estimated by the equations proposed by Baumann. It is, however, necessary to take into consideration the contributions of concrete between cracks to the deformation in order to accurately estimate the average strains in the plate elements, applying such a formula as CEB as reformed by the authors. (2) The strength capacity of three-way reinforced concrete containments against lateral forces combined with internal pressure is somewhat inferior to that of orthogonally reinforced one if compared on the condition that the volumetric reinforcement ratios are the same for the two cases of reinforcement arrangements. However, three-way reinforcement improves initial shear rigidity as well as ultimate horizontal deformability for lateral forces. (3) The ability for three-way reinforced concrete containment to absorb strain energy in the range of large deformations is superior to that of orthogonally reinforced one. The equivalent viscous damping coefficient for the former is markedly larger than that for the latter, especially at the increased deformational stages. These experimental evidences suggent that three-way system of reinforcement may constitute one of the prospective measures to improve the earthquake resistance of reinforced concrete containments. (orig./HP)

  5. Pre-operational proof and leakage rate testing requirements for concrete containment structures for CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1994-02-01

    This Standard provides the requirements for pre-operational proof tests and leakage rate tests of concrete containment structures of a containment system designed as Class Containment components. 1 fig

  6. Absorption Characteristics of Cement Combination Concrete Containing Portland Cement, fly ash, and Metakaolin

    Directory of Open Access Journals (Sweden)

    Folagbade S.O.

    2016-03-01

    Full Text Available The resistance to water penetration of cement combination concretes containing Portland cement (PC, fly ash (FA, and metakaolin (MK have been investigated at different water/cement (w/c ratios, 28-day strengths, and depths of water penetration using their material costs and embodied carbon-dioxide (eCO2 contents. Results revealed that, at equal w/c ratio, eCO2 content reduced with increasing content of FA and MK. MK contributed to the 28-day strengths more than FA. Compared with PC, FA reduced cost and increased the depth of water penetration, MK increased cost and reduced the depth of water penetration, and their ternary combinations become beneficial. At equal strengths and levels of resistance to water penetration, most of the cement combination concretes are more environmentally compatible and costlier than PC concrete. Only MK binary cement concretes with 10%MK content or more and ternary cement concretes at a total replacement level of 55% with 10%MK content or more have higher resistance to water penetration than PC concrete.

  7. Foamed concrete containing rice husk ash as sand replacement: an experimental study on compressive strength

    Science.gov (United States)

    Rum, R. H. M.; Jaini, Z. M.; Boon, K. H.; Khairaddin, S. A. A.; Rahman, N. A.

    2017-11-01

    This study presents the utilization of rice husk ash (RHA) as sand replacement in foamed concrete. The study focuses on the effect of RHA on the compressive strength of foamed concrete. RHA contains high pozzolanic material that reacts with cementitious to enhance the strength and durability of foamed concrete. RHA also acts as filler causing the foamed concrete to become denser while retaining its unique low density. A total 243 cube specimens was prepared for the compression test. Two sets of mix design were employed at water-cement (W/C) ratio of 0.55, 0.60 and cement-sand ratio of 0.50, 0.33. The results revealed that the presence of RHA as sand replacement resulted in an increase in the compressive strength of foamed concrete. Moreover, 30% to 40% RHA was the optimum content level, contributing to the compressive strength of 18.1 MPa to 22.4 MPa. The W/C ratio and superplasticiser dosage play small roles in improving workability. In contrast, density governs the compressive strength of foamed concrete.

  8. Utilization of crushed radioactive concrete for mortar to fill waste container void space

    International Nuclear Information System (INIS)

    Ishikura, Takeshi; Ohnishi, Kazuhiko; Oguri, Daiichiro; Ueki, Hiroyuki

    2004-01-01

    Minimizing the volume of radioactive waste generated during dismantling of nuclear power plants is a matter of great importance. In Japan waste forms buried in a shallow burial disposal facility as low level radioactive waste must be solidified by cement or other materials with adequate strength and must provide no harmful opening. The authors have developed an improved method to minimize radioactive waste volume by utilizing radioactive concrete for fine aggregate for mortars to fill void space in waste containers. Tests were performed with pre-placed concrete waste and with filling mortar using recycled fine aggregate produced from concrete. It was estimated that the improved method substantially increases the waste fill ratio in waste containers, thereby decreasing the total volume of disposal waste. (author)

  9. Compressive and flexural strength of concrete containing palm oil biomass clinker and polypropylene fibres

    Science.gov (United States)

    Ibrahim, M. H. Wan; Mangi, Sajjad Ali; Burhanudin, M. K.; Ridzuan, M. B.; Jamaluddin, N.; Shahidan, S.; Wong, YH; Faisal, SK; Fadzil, M. A.; Ramadhansyah, P. J.; Ayop, S. S.; Othman, N. H.

    2017-11-01

    This paper presents the effects of using palm oil biomass (POB) clinker with polypropylene (PP) fibres in concrete on its compressive and flexural strength performances. Due to infrastructural development works, the use of concrete in the construction industry has been increased. Simultaneously, it raises the demand natural sand, which causes depletion of natural resources. While considering the environmental and economic benefits, the utilization of industrial waste by-products in concrete will be the alternative solution of the problem. Among the waste products, one of such waste by-product is the palm oil biomass clinker, which is a waste product from burning processes of palm oil fibres. Therefore, it is important to utilize palm oil biomass clinker as partial replacement of fine aggregates in concrete. Considering the facts, an experimental study was conducted to find out the potential usage of palm oil fibres in concrete. In this study, total 48 number of specimens were cast to evaluate the compressive and flexural strength performances. Polypropylene fibre was added in concrete at the rate of 0.2%, 0.4% and 0.6%, and sand was replaced at a constant rate of 10% with palm oil biomass clinker. The flexural strength of concrete was noticed in the range of 2.25 MPa and 2.29 MPa, whereas, the higher value of flexural strength was recorded with 0.4% polypropylene fibre addition. Hence, these results show that the strength performances of concrete containing POB clinker could be improved with the addition of polypropylene fibre.

  10. Assessment of aggregates- cement paste border in concretes containing silica fume and fly ash

    Directory of Open Access Journals (Sweden)

    Ali Sademomtazi

    2017-12-01

    Full Text Available The bond between aggregate and cement paste, called the interfacial transition zone (ITZ is an important parameter that effect on the mechanical properties and durability of concrete. Transition zone microstructure and porosity (pores of cement paste or concrete are affected by the type and properties of materials used which evaluated in this research. On the other hand, the use of efficient, low-cost and reliable method is particularly important for evaluating of concrete performance against the chloride ion penetration and its relationships with transition zone as a suitable index to assess the durability. So far, various methods to approach the electrical Indices are presented. In this research, the effect of pozzolanic materials fly ash (10%, 20% and 30% and silica fume (5% and 10% as substitute of cement by weight in binary and ternary mixtures on the fresh and hardened concrete properties were investigated. To determine mechanical properties, the compressive strength, splitting tensile strength and modulus of elasticity tests were performed. Also, water penetration depth, porosity, water sorptivity, specific electrical resistivity, rapid chloride penetration test (RCPT and rapid chloride migration test (RCMT tests were applied to evaluate concrete durability. To examine the border of aggregate and cement paste morphology of concrete specimens, scanning electron microscope images (SEM was used. The fresh concrete results showed that the presence of silica fume in binary and ternary mixtures reduced workability and air content but fly ash increased them. Adding silica fume to mixtures of containing flay ash while increasing mechanical strength reduced the porosity and pores to 18%. The presence of pozzolanic materials in addition to increasing bond quality and uniformity of aggregate-cement matrix border a considerably positive effect on the transport properties of concrete.

  11. Statistical and Detailed Analysis on Fiber Reinforced Self-Compacting Concrete Containing Admixtures- A State of Art of Review

    Science.gov (United States)

    Athiyamaan, V.; Mohan Ganesh, G.

    2017-11-01

    Self-Compacting Concrete is one of the special concretes that have ability to flow and consolidate on its own weight, completely fill the formwork even in the presence of dense reinforcement; whilst maintaining its homogeneity throughout the formwork without any requirement for vibration. Researchers all over the world are developing high performance concrete by adding various Fibers, admixtures in different proportions. Various different kinds Fibers like glass, steel, carbon, Poly propylene and aramid Fibers provide improvement in concrete properties like tensile strength, fatigue characteristic, durability, shrinkage, impact, erosion resistance and serviceability of concrete[6]. It includes fundamental study on fiber reinforced self-compacting concrete with admixtures; its rheological properties, mechanical properties and overview study on design methodology statistical approaches regarding optimizing the concrete performances. The study has been classified into seven basic chapters: introduction, phenomenal study on material properties review on self-compacting concrete, overview on fiber reinforced self-compacting concrete containing admixtures, review on design and analysis of experiment; a statistical approach, summary of existing works on FRSCC and statistical modeling, literature review and, conclusion. It is so eminent to know the resent studies that had been done on polymer based binder materials (fly ash, metakaolin, GGBS, etc.), fiber reinforced concrete and SCC; to do an effective research on fiber reinforced self-compacting concrete containing admixtures. The key aim of the study is to sort-out the research gap and to gain a complete knowledge on polymer based Self compacting fiber reinforced concrete.

  12. Long-Term Behaviors of the OPC Concrete with Fly-ash and Type V Concrete Applied on Reactor Containment Building

    International Nuclear Information System (INIS)

    Yoon, Eui Sik; Lee, Hee Taik; Paek, Yong Lak; Park, Young Soo

    2010-01-01

    The prestressed concrete has been used extensively in the construction of Reactor Containment Buildings (RCBs) in Korea in order to strengthen the RCBs and at the same time, prevent the release of radiation due to the Design Basis Accident and Design Basis Earthquake. It is well known that the prestressed concrete loses its prestressing force over the age, and the shrinkage and creep of the concrete significantly contributes to these long term prestressing losses. In this study, an evaluations of long term behaviors of the concrete such as creep and shrinkage have been performed for two types of concretes : Ordinary Portland Cement containing fly-ash used for the Shin- Kori 1 and 2 NPP and Type V cement used for the Ul- Chin 5 and 6 NPP

  13. Long-Term Behaviors of the OPC Concrete with Fly-ash and Type V Concrete Applied on Reactor Containment Building

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Eui Sik; Lee, Hee Taik; Paek, Yong Lak [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Park, Young Soo [Korea Hydro and Nuclear Power Co., Busan (Korea, Republic of)

    2010-10-15

    The prestressed concrete has been used extensively in the construction of Reactor Containment Buildings (RCBs) in Korea in order to strengthen the RCBs and at the same time, prevent the release of radiation due to the Design Basis Accident and Design Basis Earthquake. It is well known that the prestressed concrete loses its prestressing force over the age, and the shrinkage and creep of the concrete significantly contributes to these long term prestressing losses. In this study, an evaluations of long term behaviors of the concrete such as creep and shrinkage have been performed for two types of concretes : Ordinary Portland Cement containing fly-ash used for the Shin- Kori 1 and 2 NPP and Type V cement used for the Ul- Chin 5 and 6 NPP

  14. Research status and needs for shear tests on large-scale reinforced concrete containment elements

    International Nuclear Information System (INIS)

    Oesterle, R.G.; Russell, H.G.

    1982-01-01

    Reinforced concrete containments at nuclear power plants are designed to resist forces caused by internal pressure, gravity, and severe earthquakes. The size, shape, and possible stress states in containments produce unique problems for design and construction. A lack of experimental data on the capacity of reinforced concrete to transfer shear stresses while subjected to biaxial tension has led to cumbersome if not impractical design criteria. Research programs recently conducted at the Construction Technology Laboratories and at Cornell University indicate that design criteria for tangential, peripheral, and radial shear are conservative. This paper discusses results from recent research and presents tentative changes for shear design provisions of the current United States code for containment structures. Areas where information is still lacking to fully verify new design provisions are discussed. Needs for further experimental research on large-scale specimens to develop economical, practical, and reliable design criteria for resisting shear forces in containment are identified. (orig.)

  15. A Constraint Programming Model for Fast Optimal Stowage of Container Vessel Bays

    DEFF Research Database (Denmark)

    Delgado-Ortegon, Alberto; Jensen, Rune Møller; Janstrup, Kira

    2012-01-01

    Container vessel stowage planning is a hard combinatorial optimization problem with both high economic and environmental impact. We have developed an approach that often is able to generate near-optimal plans for large container vessels within a few minutes. It decomposes the problem into a master...... planning phase that distributes the containers to bay sections and a slot planning phase that assigns containers of each bay section to slots. In this paper, we focus on the slot planning phase of this approach and present a constraint programming and integer programming model for stowing a set...... of containers in a single bay section. This so-called slot planning problem is NP-hard and often involves stowing several hundred containers. Using state-of-the-art constraint solvers and modeling techniques, however, we were able to solve 90% of 236 real instances from our industrial collaborator to optimality...

  16. Applicability of JIS SPV 50 steel to primary containment vessel of nuclear power station

    International Nuclear Information System (INIS)

    Iida, Kunihiro; Ishikawa, Koji; Sakai, Keiichi; Onozuka, Masakazu; Sato, Makoto.

    1979-01-01

    The space within reactor containment vessels must be expanded in order to improve the reliability of nuclear power plants, accordingly the adoption of large reactor containment vessels is investigated. SGV 42 and 49 steels in JIS G 3118 have been used for containment vessels so far, but stress relief annealing is required when the thickness exceeds 38 mm. The time has come when the use of thicker conventional plates without stress relieving or the use of high strength steel must be examined in detail. In this study, the tests of confirming material properties were carried out on SPV 50 in JIS G 3115, Steels for pressure vessels, aiming at the method of fabrication without stress relieving. The highest and lowest temperatures in use were set at 171 deg and -8 deg C, respectively. The chemical composition and the mechanical properties of the plates tested, the method of welding, the results of tensile test on the parent metal and the welds, the required lowest preheating temperature, the fracture toughness at low temperature and the brittle fracture causing test are reported. The parent metal and the welded joints of SPV 50 have the properties suitable to reactor containment vessels, namely the sufficient fracture toughness to guarantee the prevention of unstable fracture when the method of welding without stress relieving is adopted. (Kako, I.)

  17. Current state of knowledge on the behavior of steel liners in concrete containments subjected to overpressurization loads

    International Nuclear Information System (INIS)

    von Riesemann, W.A.; Parks, M.B.

    1993-01-01

    In the United States, concrete containment buildings for commercial nuclear power plants have steel liners that act as the intemal pressure boundary. The liner abuts the concrete, acting as the interior concrete form. The liner is attached to the concrete by either studs or by a continuous structural shape (such as a T-section or channel) that is either continuously or intermittently welded to the liner. Studs are commonly used in reinforced concrete containments, while prestressed containments utilize a structural element as the anchorage. The practice in some countries follows the US practice, while in other countries the containment does not have a steel liner. In this latter case, there is a true double containment, and the annular region between the two containments is vented. This paper will review the practice of design of the liner system prior to the consideration of severe accident loads (overpressurization loads beyond the design conditions)

  18. Scenario based optimization of a container vessel with respect to its projected operating conditions

    Directory of Open Access Journals (Sweden)

    Jonas Wagner

    2014-06-01

    Full Text Available In this paper the scenario based optimization of the bulbous bow of the KRISO Container Ship (KCS is presented. The optimization of the parametrically modeled vessel is based on a statistically developed operational profile generated from noon-to-noon reports of a comparable 3600 TEU container vessel and specific development functions representing the growth of global economy during the vessels service time. In order to consider uncertainties, statistical fluctuations are added. An analysis of these data lead to a number of most probable upcoming operating conditions (OC the vessel will stay in the future. According to their respective likeliness an objective function for the evaluation of the optimal design variant of the vessel is derived and implemented within the parametrical optimization workbench FRIENDSHIP Framework. In the following this evaluation is done with respect to vessel's calculated effective power based on the usage of potential flow code. The evaluation shows, that the usage of scenarios within the optimization process has a strong influence on the hull form.

  19. Scenario based optimization of a container vessel with respect to its projected operating conditions

    Science.gov (United States)

    Wagner, Jonas; Binkowski, Eva; Bronsart, Robert

    2014-06-01

    In this paper the scenario based optimization of the bulbous bow of the KRISO Container Ship (KCS) is presented. The optimization of the parametrically modeled vessel is based on a statistically developed operational profile generated from noon-to-noon reports of a comparable 3600 TEU container vessel and specific development functions representing the growth of global economy during the vessels service time. In order to consider uncertainties, statistical fluctuations are added. An analysis of these data lead to a number of most probable upcoming operating conditions (OC) the vessel will stay in the future. According to their respective likeliness an objective function for the evaluation of the optimal design variant of the vessel is derived and implemented within the parametrical optimization workbench FRIENDSHIP Framework. In the following this evaluation is done with respect to vessel's calculated effective power based on the usage of potential flow code. The evaluation shows, that the usage of scenarios within the optimization process has a strong influence on the hull form.

  20. Applicability of JIS SPV 50 steel to primary containment vessels of nuclear power stations

    International Nuclear Information System (INIS)

    Iida, K.; Ishikawa, K.; Satoh, M.; Soya, I.

    1980-01-01

    The fracture toughness of JIS SPV 50 steel and its weldment has been examined in order to verify the applicability of these materials to primary containment vessels of nuclear power stations. Test results were evaluated using elastic plastic fracture mechanics through the COD and the J integral concepts for non ductile fracture initiation characteristics. Linear fracture mechanics was employed for propagation arrest characteristics. Results showed that the materials tested here have a sufficient fracture toughness to prevent nonductile fracture and that this steel is a suitable material for use in construction of primary containment vessels of nuclear power stations. (author)

  1. Minimizing Lid Overstows in Master Stowage Plans for Container Vessels is NP-Complete

    DEFF Research Database (Denmark)

    Ajspur, Mai Lise; Jensen, Rune Møller; Guilbert, Nicolas

    Container vessel stowage is a particularly hard combinatorial problem within the shipping industry. The currently most successful approaches decompose the problem hierarchically and first generate a master plan that handle highlevel constraints and objectives such as balance and stress moments...... that it is an NP -complete problem to generate master plans that minimize the number of these lid overstows. Since any efficient approach to container vessel stowage most likely must include a master plan, the implication of this result is that future research must focus and developing good heuristics...

  2. Experimental study of a laboratory concrete material representative of containment buildings: desorption isotherms and permeability determination

    International Nuclear Information System (INIS)

    Semete, P.; Fevrier, B.; Delorme, J.; Sanahuja, J.; Desgree, P.; Le Pape, Y.

    2015-01-01

    The isotherm sorption curve is a first order parameter for the calculations of concrete drying and/or creep using Finite Element Analysis. An experimental campaign was undertaken by EDF MMC in order to characterize the first desorption isotherm at room temperature of a laboratory material representative of concrete containment buildings. Long term drying tests were carried out on cement paste and on three samples geometries on concrete (with radial and axial one-dimensional drying on thin disks and multi-dimensional drying on Representative Elementary Volumes). The measurements results (porosity, densities and mass loss curves) are provided and the isotherms obtained for the four different configurations are compared. Several analyses of the results are proposed including the assessment of a criterion for the determination of the moisture content final balance (estimation of the asymptotic mass loss) and the back-analysis of equivalent permeability. (authors)

  3. Concrete

    OpenAIRE

    Kruse Aagaard, Anders

    2015-01-01

    Concrete is a component of coherent transition between a concrete base and a wooden construction. The structure is based on a quantity of investigations of the design possibilities that arise when combining digital fabrication tools and material capacities.Through tangible experiments the project discusses materiality and digitally controlled fabrications tools as direct expansions of the architect’s digital drawing and workflow. The project sees this expansion as an opportunity to connect th...

  4. Non-linear analysis up to rupture of a model of a multi-cavity prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Rebora, B.; Uffer, F.; Zimmermann, T.

    1977-01-01

    Within the frame of a German-Swiss agreement concerning the project of a high-temperature nuclear plant (HHT), the Swiss Federal Institute for Reactor Research (EIR, in Wuerlingen) has developed an integrated variant of an helium-cooled high temperature reactor of 3x500 Mwe. A test on a model (1:20) of this prestressed concrete nuclear vessel with multiple cavities has been carried out under the supervision of 'Bonnard et Gardel ingenieurs-conseils SA (BG). The aim of this analysis is to determine the mechanism of ruin and ultimate load of the structure. In addition, comparison with the results of the test emphasizes the mathematical model with a view to its utilisation for the analysis of any prestressed concrete nuclear vessel. The principal interest of this paper is to show the accuracy of non-linear analysis of a complex massive structure with the test results and the evolution of the behaviour of the structure from the apparition of the first crack up to the ruin by rupture of the steel wires. (Auth.)

  5. Strengths and Failure Characteristics of Self-Compacting Concrete Containing Recycled Waste Glass Aggregate

    Directory of Open Access Journals (Sweden)

    Rahman Khaleel AL-Bawi

    2017-01-01

    Full Text Available The effects of different proportions of green-colored waste glass (WG cullet on the mechanical and fracture properties of self-compacting concrete (SCC were experimentally investigated. Waste bottles were collected, washed, crushed, and sieved to prepare the cullet used in this study. Cullet was incorporated at different percentages (0%, 20%, 40%, 60%, 80%, and 100% by weight instead of natural fine aggregate (NFA and/or natural coarse aggregate (NCA. Three SCC series were designed with a constant slump flow of 700±30 mm, total binder content of 570 kg/m3 and at water-to-binder (w/b ratio of 0.35. Moreover, fly ash (FA was used in concrete mixtures at 20% of total binder content. Mechanical aspects such as compressive, splitting tensile, and net flexural strengths and modulus of elasticity of SCC were investigated and experimentally computed at 28 days of age. Moreover, failure characteristics of the concretes were also monitored via three-point bending test on the notched beams. The findings revealed that the mechanical properties as well as fracture parameters were adversely influenced by incorporating of WG cullet. However, highest reduction of compressive strength did not exceed 43% recorded at 100% WG replacement level. Concretes containing WG showed less brittle behavior than reference concrete at any content.

  6. Monitoring of prestressed concrete pressure vessels. 1. An overview of concrete embedment strain instrumentation and calibration test results for selected concrete embedment strain meters

    International Nuclear Information System (INIS)

    Naus, D.J.; Hurtt, C.C.

    1978-01-01

    The report presents results of calibration tests on strain meters. The approach was divided into two phases: (1) an overview of meter performance criteria for PCPV applications and techniques for strain measurements in concrete and (2) procurement of commercially available gages and their evaluation to assess the reliability of manufacturer-supplied calibration factors. Calibration test results for gages embedded in 15.2-cm-diam by 54-cm cylindrical concrete specimens indicated that calibration factors should be determined (verified) by embedding samples of the gages in test specimens fabricated using a representative mix and that further research should be conducted on other measurement techniques based on inductance, capacitance, semiconductors, and fluidic principles

  7. Design and construction of a large reinforced concrete containment model to be tested to failure

    International Nuclear Information System (INIS)

    Ucciferro, J.J.; Horschel, D.S.

    1987-01-01

    The US Nuclear Regulatory Commission is investigating the performance of LWR containments subjected to severe accidents. This work is being performed by the Containment Integrity Division at Sandia National Laboratories (Sandia). The latest research effort involves the testing of a 1/6-scale reinforced concrete containment model. The containment, which was designed and constructed by United Engineers and Constructors, is the largest and most complex model of its kind. The design and construction of the containment model are the subject of this paper. The objective of the containment model tests is to generate data that can be used to qualify methods for reliably predicting the response of LWR containment buildings to severe accident loads. The data recorded during testing include deformations and leakage past sealing surfaces, as well as strains and displacements of the containment shell

  8. Shell finite element of reinforced concrete for internal pressure analysis of nuclear containment building

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hong Pyo, E-mail: hplee@kepri.re.k [Nuclear Power Laboratory, Korea Electric Power Research Institute, 103-16 Munji-Dong, Yuseong-Gu, Daejeon 305-380 (Korea, Republic of)

    2011-02-15

    Research highlights: Finite element program with 9-node degenerated shell element was developed. The developed program was mainly forced to analyze nuclear containment building. Concrete material model is adapted Niwa and Yamada failure criteria. The performance of program developed is verified through various numerical examples. The numerical analysis results similar to the experimental data. - Abstract: This paper describes a 9-node degenerated shell finite element (FE), an analysis program developed for ultimate pressure capacity evaluation and nonlinear analysis of a nuclear containment building. The shell FE developed adopts the Reissner-Mindlin (RM) assumptions to consider the degenerated shell solidification technique and the degree of transverse shear strain occurring in the structure. The material model of the concrete determines the level of the concrete stress and strain by using the equivalent stress-equivalent strain relationship. When a crack occurs in the concrete, the material behavior is expressed through the tension stiffening model that takes adhesive stress into account and through the shear transfer mechanism and compressive strength reduction model of the crack plane. In addition, the failure envelope proposed by Niwa is adopted as the crack occurrence criteria for the compression-tension region, and the failure envelope proposed by Yamada is used for the tension-tension region. The performance of the program developed is verified through various numerical examples. The analysis based on the application of the shell FE developed from the results of verified examples produced results similar to the experiment or other analysis results.

  9. Shell finite element of reinforced concrete for internal pressure analysis of nuclear containment building

    International Nuclear Information System (INIS)

    Lee, Hong Pyo

    2011-01-01

    Research highlights: → Finite element program with 9-node degenerated shell element was developed. → The developed program was mainly forced to analyze nuclear containment building. → Concrete material model is adapted Niwa and Yamada failure criteria. → The performance of program developed is verified through various numerical examples. → The numerical analysis results similar to the experimental data. - Abstract: This paper describes a 9-node degenerated shell finite element (FE), an analysis program developed for ultimate pressure capacity evaluation and nonlinear analysis of a nuclear containment building. The shell FE developed adopts the Reissner-Mindlin (RM) assumptions to consider the degenerated shell solidification technique and the degree of transverse shear strain occurring in the structure. The material model of the concrete determines the level of the concrete stress and strain by using the equivalent stress-equivalent strain relationship. When a crack occurs in the concrete, the material behavior is expressed through the tension stiffening model that takes adhesive stress into account and through the shear transfer mechanism and compressive strength reduction model of the crack plane. In addition, the failure envelope proposed by Niwa is adopted as the crack occurrence criteria for the compression-tension region, and the failure envelope proposed by Yamada is used for the tension-tension region. The performance of the program developed is verified through various numerical examples. The analysis based on the application of the shell FE developed from the results of verified examples produced results similar to the experiment or other analysis results.

  10. Development of ultrasonic testing technique to inspect containment liners embedded in concrete on nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Ishida, H.; Kurozumi, Y. [Inst. of Nuclear Safety System, Incorporated, Mihama, Fukui (Japan); Kaneshima, Y. [The Kansai Electric Power Company, Inc., Mihama, Fukui (Japan)

    2004-07-01

    The purpose of this study is development of ultrasonic testing technique to inspect containment liners embedded in concrete on nuclear power plants. Integrity of containment liners on nuclear power plants can be secured by suitable present operation and maintenance. Furthermore, non-destructive testing technique to inspect embedded liners will ensure the integrity of the containment further. In order to develop the non-destructive testing technique, ultrasonic transducers were made newly and ultrasonic testing data acquisition and evaluation were carried out by using a mock-up. We adopted the surface shear horizontal (SH) wave, low frequency (0.3-0.5MHz), to be able to detect an echo from a defect against attenuation of ultrasonic waves due to long propagation in the liners and dispersion into concrete. We made transducers with three large active elements (40mm x 40mm) in a line which were equivalent to a 120mm width active element. Artificial hollows, {phi}200mm - 19mm depth (1/2thickness) and {phi}200mm - 9.5mm depth (1/4thickness), were made on a surface of a mock-up: carbon steel plate, 38mm thickness, 2,000mm length, 1000mm width. The surfaces of the plate were covered with concrete in order to simulate liners embedded in concrete. As a result of the examinations, the surface SH transducers could detect clearly the echo from the hollows at a distance of 1500mm. We evaluate that the newly made surface SH transducers with three elements have ability of detection of defects such as corrosion on the liners embedded in concrete. (author)

  11. Electrical and mechanical properties of asphalt concrete containing conductive fibers and fillers

    NARCIS (Netherlands)

    Wang, H.; Yang, Jun; Liao, Hui; Chen, Xianhua

    2016-01-01

    Electrically conductive asphalt concrete has the potential to satisfy multifunctional applications. Designing such asphalt concrete needs to balance the electrical and mechanical performance of asphalt concrete. The objective of this study is to design electrically conductive asphalt concrete

  12. Ultimate internal pressure capacity of a reinforced concrete Mark III containment

    International Nuclear Information System (INIS)

    McGaughy, J.P. Jr.; Lin, F.T.; Sen, S.K.

    1983-01-01

    The static ultimate capacity of a Mark III BWR pressure suppression type containment has been investigated with a view to determine its capability to withstand the internal pressure associated with a postulated hydrogen burn. The reinforced concrete containment consists of a right circular cylinder covered by a hemispherical dome and supported on a flat circular foundation mat. A 1/4'' thick welded steel liner plate covers the inside surface of the containment shell. The cylinder is a 3.5 ft. thick shell with an inside radius of 62.0 feet. The thickness of the dome is 3.5 feet. Reinforcement in the shell is comprised of multi-layers of circumferential, meridional and diagonal rebars. Major containment penetrations consists of a circular equipment hatch and two personnel airlock assemblies. The containment ultimate capacity is determined by performing a non-linear analysis using the proprietary finite element computer code 'FINEL'. The code has the capability of modelling concrete cracking in tension and redistribution forces and moments to account for such phenomenon. For analysis purposes, the finite element model included the containment dome and the upper portion of the containment cylinder with appropriate boundary conditions applied at the model cut off region. This portion of the containment structure is selected because the segment of the cylinder that is included in the model has the least amount of hopp reinforcement, and when the general yield state is reached, the hoop reinforcement will be the limiting element. The containment structure has been treated as an axisymmetric shell using axisymmetric quadrilateral finite elements in the radial plane to model the liner plate and concrete. The reinforcing steel have been idealized by finite elements with unidirectional stiffness. (orig./RW)

  13. A photoelastic study of the effects of an impulsive seismic wave on a nuclear containment vessel

    International Nuclear Information System (INIS)

    Burger, C.P.

    1981-01-01

    A dynamic photoelastic study of the progressive movement of a dilatational P-wave into a model of a nuclear containment vessel,is studied. The reflections at the dome abutments are observed and the strong flexural wave that deforms the dome itself is studied with photoelasticity and with dynamic strain gage procedures. (E.G.) [pt

  14. Concrete containment tests: Phase 2, Structural elements with liner plates: Interim report

    International Nuclear Information System (INIS)

    Hanson, N.W.; Roller, J.J.; Schultz, D.M.; Julien, J.T.; Weinmann, T.L.

    1987-08-01

    The tests described in this report are part of Phase 2 of the Electric Power Research Institute (EPRI) program. The overall objective of the EPRI program is to provide a test-verified analytical method of estimating capacities of concrete reactor containment buildings under internal overpressurization from postulated degraded core accidents. The Phase 2 testing included seven large-scale specimens representing structural elements from reinforced and prestressed concrete reactor containment buildings. Six of the seven test specimens were square wall elements. Of these six specimens, four were used for biaxial tension tests to determine strength, deformation, and leak-rate characteristics of full-scale wall elements representing prestressed concrete containment design. The remaining two square wall elements were used for thermal buckling tests to determine whether buckling of the steel liner plate would occur between anchorages when subjected to a sudden extreme temperature differential. The last of the seven test specimens for Phase 2 represented the region where the wall and the basemat intersect in a prestressed concrete containment building. A multi-directional loading scheme was used to produce high bending moments and shear in the wall/basemat junction region. The objective of this test was to determine if there is potential for liner plate tearing in the junction region. Results presented include observed behavior and extensive measurements of deformations and strains as a function of applied load. The data are being used to confirm analytical models for predicting strength and deformation of containment structures in a separate parallel analytical investigation sponsored by EPRI

  15. Some Properties of Carbon Fiber Reinforced Magnetic Reactive Powder Concrete Containing Nano Silica

    Directory of Open Access Journals (Sweden)

    Zain El-Abdin Raouf

    2016-08-01

    Full Text Available This study involves the design of 24 mixtures of fiber reinforced magnetic reactive powder concrete containing nano silica. Tap water was used for 12 of these mixtures, while magnetic water was used for the others. The nano silica (NS with ratios (1, 1.5, 2, 2.5 and 3 % by weight of cement, were used for all the mixtures. The results have shown that the mixture containing 2.5% NS gives the highest compressive strength at age 7 days. Many different other tests were carried out, the results have shown that the carbon fiber reinforced magnetic reactive powder concrete containing 2.5% NS (CFRMRPCCNS had higher compressive strength, modulus of rupture, splitting tension, stress in compression and strain in compression than the corresponding values for the carbon fiber reinforced nonmagnetic reactive powder concrete containing the same ratio of NS (CFRNRPCCNS. The percentage increase in these values for CFRMRPCCNS were (22.37, 17.96, 19.44, 6.44 and 25.8 % at 28 days respectively, as compared with the corresponding CFRNRPCCNS mixtures.

  16. Scenario based optimization of a container vessel with respect to its projected operating conditions

    Directory of Open Access Journals (Sweden)

    Wagner Jonas

    2014-06-01

    Full Text Available In this paper the scenario based optimization of the bulbous bow of the KRISO Container Ship (KCS is presented. The optimization of the parametrically modeled vessel is based on a statistically developed operational profile generated from noon-to-noon reports of a comparable 3600 TEU container vessel and specific development functions representing the growth of global economy during the vessels service time. In order to consider uncertainties, statistical fluctuations are added. An analysis of these data lead to a number of most probable upcoming operating conditions (OC the vessel will stay in the future. According to their respective likeliness an objective function for the evaluation of the optimal design variant of the vessel is derived and implemented within the parametrical optimization workbench FRIENDSHIP Framework. In the following this evaluation is done with respect to vessel’s calculated effective power based on the usage of potential flow code. The evaluation shows, that the usage of scenarios within the optimization process has a strong influence on the hull form.

  17. Examination of leakage aspects through concrete - steel interfaces at and around containment penetration assemblies

    International Nuclear Information System (INIS)

    Chakrabarti, S.K.; Sai, A.S.R.; Basu, P.C.

    1994-01-01

    Penetration assemblies are parts required to be provided in the containment wall/dome to permit piping, mechanical devices, equipments, electrical cables, personnel movements etc. Integrity of arrangements with respect to leak tightness at or around these penetration assemblies, is of utmost importance for achieving safe functioning of containment. Considering the feasibilities in controlling leakages along different possible paths, it has been found necessary to examine in detail the leakage possibilities at concrete - steel interfaces at and around penetration assemblies. The present paper addresses this issue with respect to the important related aspects like constructional details, testing conditions, normal operating conditions, and the accidental situation associated with containment structures. (author)

  18. EFFECT OF SEA WATER ON THE STRENGTH OF POROUS CONCRETE CONTAINING PORTLAND COMPOSITE CEMENT AND MICROFILAMENT POLYPROPYLENE FIBER

    OpenAIRE

    TJARONGE, M.W

    2011-01-01

    The aim of this research is to study the influence of sea water on the strength of porous concrete containing Portland Composite cement and micro monofilament polypropylene fibre. The specimens of porous concrete were immersed in the sea water up to 28 days. The compressive strength test and flexural strength test were carried out at 3, 7 and 28 days in order to investigate the strength development. The test result indicated that the strength of porous concrete can develop in t...

  19. The moisture conditions of nuclear reactor concrete containment walls - an example for a BWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, L.O.; Johansson, P. [Lund Institute of Technology, Laboratory of Building Materials, PO Box 118, 221 00 Lund (Sweden)

    2006-07-01

    A method is presented on how to quantify the moisture conditions of nuclear concrete containment walls. The method is based on first quantifying the boundary conditions at the outer and inner surfaces and then describing the moisture fixation and moisture transport within the concrete wall. The temperature and humidity conditions of the outdoor air and of the air close to the wall surfaces are monitored for a period of time and the vapour contents in the different points are compared. From the differences between the vapour contents the sources of moisture are identified and quantified. The previous and future climatic conditions are then predicted. An example is given for the conditions in the containment walls at Barsebaeck nuclear power plant, where moisture measurements have been performed in situ and on samples taken from the walls. (authors)

  20. The mechanical properties of brick containing recycled concrete aggregate and polyethylene terephthalate waste as sand replacement

    Science.gov (United States)

    Sheikh Khalid, Faisal; Bazilah Azmi, Nurul; Natasya Mazenan, Puteri; Shahidan, Shahiron; Ali, Noorwirdawati

    2018-03-01

    This research focuses on the performance of composite sand cement brick containing recycle concrete aggregate and waste polyethylene terephthalate. This study aims to determine the mechanical properties such as compressive strength and water absorption of composite brick containing recycled concrete aggregate (RCA) and polyethylene terephthalate (PET) waste. The bricks specimens were prepared by using 100% natural sand, they were then replaced by RCA at 25%, 50% and 75% with proportions of PET consists of 0.5%, 1.0% and 1.5% by weight of natural sand. Based on the results of compressive strength, only RCA 25% with 0.5% PET achieve lower strength than normal bricks while others showed a high strength. However, all design mix reaches strength more than 7N/mm2 as expected. Besides that, the most favorable mix design that achieves high compressive strength is 75% of RCA with 0.5% PET.

  1. Nonlinear finite element analysis of nuclear reinforced prestressed concrete containments up to ultimate load capacity

    International Nuclear Information System (INIS)

    Gupta, A.; Singh, R.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1996-01-01

    For safety evaluation of nuclear structures a finite element code ULCA (Ultimate Load Capacity Assessment) has been developed. Eight/nine noded isoparametric quadrilateral plate/shell element with reinforcement as a through thickness discrete but integral smeared layer of the element is presented to analyze reinforced and prestressed concrete structures. Various constitutive models such as crushing, cracking in tension, tension stiffening and rebar yielding are studied and effect of these parameters on the reserve strength of structures is brought out through a number of benchmark tests. A global model is used to analyze the prestressed concrete containment wall of a typical 220 MWe Pressurized Heavy Water Reactor (PHWR) up to its ultimate capacity. This demonstrates the adequacy of Indian PHWR containment design to withstand severe accident loads

  2. Compilation of three-dimensional coordinates and specific data of the instrumentation of the prestressed concrete pressure vessel/high temperature helium test rig

    International Nuclear Information System (INIS)

    Klausinger, D.

    1977-04-01

    The positions of the thermoelements, strain gauges of various types, and of Gloetzl instruments installed by SGAE in the model vessel of the Common Project Prestressed Concrete Pressure Vessel/High Temperature Helium Test Rig are defined in cylindrical coordinates. The specific data of the instruments are given like configuration of multiple instruments; type, group and number of the instrument; number of cable and of channel; calibration factors; resistances of instruments and cables. (author)

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

    Science.gov (United States)

    Philipose, K.; Shenton, B.

    2011-04-01

    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 the early 1970s and is currently in a decommissioned state. The structure at present is under surveillance and monitoring. In the year 2000, a field investigation was conducted as part of a condition assessment and corrosion was detected in some of the grouted post-tension cable strands. However, no further work was done at that time to determine the cause, nature, impact and extent of the corrosion. An investigation of the Gentilly-1 containment building is currently underway to assess the condition of grouted post-tensioning cables and reinforced concrete. At two selected locations, concrete and steel reinforcements were removed from the containment building wall to expose horizontal cables. Individual cable strands and reinforcement bars were instrumented and measurements were taken in-situ before removing them for forensic examination and destructive testing to determine the impact of ageing and corrosion. Concrete samples were also removed and tested in a laboratory. The purpose of the field investigation and laboratory testing, using this structure as a test bed, was also to collect material ageing data and to develop potential Nondestructive Examination (NDE) methods to monitor Containment Building Integrity. The paper describes the field work conducted and the test results obtained for concrete, reinforcement and post-tensioning cables.

  4. Inorganic material candidates to replace a metallic or non-metallic concrete containment liner

    Energy Technology Data Exchange (ETDEWEB)

    Seni, C [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Mills, R H [Toronto Univ., ON (Canada)

    1994-12-31

    Internal liners for concrete containments are generally organic or metals. They have to be able to inhibit radioactive leakage into the environment, but both types have shortcomings. The results of research to develop a better liner are published in this paper. The best material found was fibre-reinforced mortar. Of the fibres considered, steel, kevlar and glass were the best, each showing advantages and disadvantages. 1 ref., 9 tabs., 12 figs.

  5. Inorganic material candidates to replace a metallic or non-metallic concrete containment liner

    International Nuclear Information System (INIS)

    Seni, C.; Mills, R.H.

    1994-01-01

    Internal liners for concrete containments are generally organic or metals. They have to be able to inhibit radioactive leakage into the environment, but both types have shortcomings. The results of research to develop a better liner are published in this paper. The best material found was fibre-reinforced mortar. Of the fibres considered, steel, kevlar and glass were the best, each showing advantages and disadvantages. 1 ref., 9 tabs., 12 figs

  6. Design of the containment structure in prestressed concrete for the Embalse-Cordoba Nuclear Power Plant

    International Nuclear Information System (INIS)

    Godoy, A.R.; Marinelli, C.A.; Gruenbaum, C.E.

    1978-01-01

    The design of a typical prestressed concrete containment structure for a 600 MW Candu - PHW Reactor, presently under construction at Embalse - Cordoba, Argentina is briefly described. The structural behaviour , adcpted prestressing system and tendon pattern are described. Afterwards the evaluation of the prestressing forces as well as the losses assessment and the prestressing sequence are discussed. Finally, some conclusions are drawn in the light of the experience gained at different stages of the construction. (Author)

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

    Directory of Open Access Journals (Sweden)

    Shenton B.

    2011-04-01

    Full Text Available 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 the early 1970s and is currently in a decommissioned state. The structure at present is under surveillance and monitoring. In the year 2000, a field investigation was conducted as part of a condition assessment and corrosion was detected in some of the grouted post-tension cable strands. However, no further work was done at that time to determine the cause, nature, impact and extent of the corrosion. An investigation of the Gentilly-1 containment building is currently underway to assess the condition of grouted post-tensioning cables and reinforced concrete. At two selected locations, concrete and steel reinforcements were removed from the containment building wall to expose horizontal cables. Individual cable strands and reinforcement bars were instrumented and measurements were taken in-situ before removing them for forensic examination and destructive testing to determine the impact of ageing and corrosion. Concrete samples were also removed and tested in a laboratory. The purpose of the field investigation and laboratory testing, using this structure as a test bed, was also to collect material ageing data and to develop potential Nondestructive Examination (NDE methods to monitor Containment Building Integrity. The paper describes the field work conducted and the test results obtained for concrete, reinforcement and post-tensioning cables.

  8. Delayed behaviour of concrete in nuclear power plant containment: analysis and modelling

    International Nuclear Information System (INIS)

    Granger, L.

    1995-02-01

    The containment of French nuclear power plant of the 1300 and 1400 MWe PWR type are made of prestressed concrete and their delayed behaviour is systematically monitored by a very complete instrumentation. In an accidental phase, the tightness of the 1.2 m thick structure, dimensioned to withstand an internal absolute pressure of 0.5 MPa depends mainly on the residual prestress of concrete. But surveillance devices reveal substantial differences from one site to another, from which the regulation calculation models cannot make satisfactory allowance. For the purpose of improving the management of the population of power stations, EDF in 1992 initiated a large study aimed at predicting the true creep behaviour of the containments already built. This study, more material oriented, includes numerous shrinkage and creep tests on reconstructed concrete in laboratory as well as on cement paste and aggregate. The main results are presented in part one. In the second part, we consider the different delayed strains of concrete one by one. A precise analysis of the physico-chemical phenomena at the origin of the delayed behaviours, leads us to propose a practical modelling of concrete in an overall equivalent continuous material approach. Secondly, the few parameters of the model are determined on the experimental results. In order to do so, two particular finite element programs in CESAR-LCPC have been developed. The first one permits to take into account the non linear diffusion of humidity in concrete as a function of temperature. The diffusion coefficient D(C) (C = water content) is fitted on the loss of weight tests as a function of time. The second step is a creep calculation; first, the program reads back the temperature and humidity results of the previous computations and then calculates the different delayed strains in time. For basic creep, we have chosen a viscoelastic model function of temperature and humidity. The numerical scheme uses the principle of

  9. NFAP calculation of the response of a 1/6 scale reinforced concrete containment model

    International Nuclear Information System (INIS)

    Costantino, C.J.; Pepper, S.; Reich, M.

    1989-01-01

    The details associated with the NFAP calculation of the pressure response of the 1/6th scale model containment structure are discussed in this paper. Comparisons are presented of some of the primary items of interest with those determined from the experiment. It was found from this comparison that the hoop response of the containment wall was adequately predicted by the NFAP finite element calculation, including the response in the high pressure, high strain range at which cracking of the concrete and yielding of the hoop reinforcement occurred. In the vertical or meridional direction, it was found that the model was significantly softer than predicted by the finite element calculation; that is, the vertical strains in the test were three to four times larger than computed in the NFAP calculation. These differences were noted even at low strain levels at which the concrete would not be expected to be cracked under tensile loadings. Simplified calculations for the containment indicate that the vertical stiffness of the wall is similar to that which would be determined by assuming the concrete fully cracked. Thus, the experiment indicates an anomalous behavior in the vertical direction

  10. Mechanical Properties of High Performance Concrete Containing Waste Plastic as Aggregate

    Directory of Open Access Journals (Sweden)

    Abdulkader Ismail Al-Hadithi

    2015-08-01

    Full Text Available The world's population growth and the increasing demand for new infrastructure facilities and buildings , present us with the vision of a higher resources consumption, specially in the form of more durable concrete such as High Performance Concrete (HPC . Moreover , the growth of the world pollution by plastic waste has been tremendous. The aim of this research is to investigate the change in mechanical properties of HPC with added waste plastics in concrete. For this purpose 2.5%, 5% and 7.5% in volume of natural fine aggregate in the HPC mixes were replaced by an equal volume of Polyethylene Terephthalate (PET waste , got by shredded PET bottles. The mechanical properties (compressive, splitting tensile, and flexural strength evaluated at the ages of (7 ,28, 56 and 91 days while the static modulus of elasticity tested at (28 and 91 days . The results indicated that HPC containing PET-aggregate presented lower compressive strength and static elasticity . The splitting strength displayed an arising trend at the initial stages, however, they have a tendency to decrease after a while. On the other hand, flexural strength results gave better modulus of rapture at all ages of curing , as compared with reference concrete specimens.

  11. Behaviours of reinforced concrete containment models under thermal gradient and internal pressure

    International Nuclear Information System (INIS)

    Aoyagi, Y.; Ohnuma, H.; Yoshioka, Y.; Okada, K.; Ueda, M.

    1979-01-01

    The provisions for design concepts in Japanese Technical Standard of Concrete Containments for Nuclear Power Plants require to take account of thermal effects into design. The provisions also propose that the thermal effects could be relieved according to the degree of crack formation and creep of concrete, and may be neglected in estimating the ultimate strength capacity in extreme environmental loading conditions. This experimental study was carried out to clarify the above provisions by investigating the crack and deformation behaviours of two identical reinforced cylindrical models with dome and basement (wall outer diameter 160 cm, and wall thickness 10 cm). One of these models was hydraulically pressurized up to failure at room temperature and the other was subjected to similar internal pressure combined with the thermal gradient of approximately 40 to 50 0 C across the wall. Initial visual cracks were recognized when the stress induced by the thermal gradient reached at about 85% of bending strength of concrete used. The thermal stress of reinforcement calculated with the methods proposed by the authors using an average flexural rigidity considering the contribution of concrete showed good agreement with test results. The method based on the fully cracked section, however, was recognized to underestimate the measured stress. These cracks considerably reduced the initial deformation caused by subsequent internal pressure. (orig.)

  12. The surrounding concrete structure of the containment as a safety component

    International Nuclear Information System (INIS)

    Alex, H.; Kuntze, W.M.

    1978-01-01

    This paper will briefly discuss the containments of the various types of reactors in the Federal Republic of Germany and will try to show the importance of the surrounding concrete structures with respect to safety. It will be seen that the surrounding concrete structures serve in any case - as protection against external events - as secondary shielding and must therefore be considered as a passive safety feature. The design requirements for the surrounding concrete structures with respect to protection against external events and to physical protection generally supplement each other. Reference will be made to possible alternatives, which might result from studies of underground siting of nuclear power plants. Whether or not this type of construction can lead to additional safety can only be judged when the results of all these studies - some of which are still under way - are evaluated. The concluding part of this paper will deal with the responsibilities of the civil engineering supervisory authorities and the nuclear licensing authorities with respect to the surrounding concrete structures. (orig.) [de

  13. Material properties for reactor pressure vessels and containment shells under dynamic loading

    International Nuclear Information System (INIS)

    Albertini, C.

    1997-01-01

    The effects of high strain rate, dynamic biaxial loading and deformation mode (tension, shear) on the mechanical properties of AISI 316 austenitic stainless steel in as-received and pre-damaged (creep, LCF) conditions are reported. This research was conducted to assess the performances of the containment shell of fast breeder reactors. The results of this research have been utilized to prepare similar investigations for SA 537 Class 1 ferritic steel used for the containment shell of LWR. The first results of these investigations are reported. A programme to study the mechanical properties of plain concrete with real size aggregate at high strain rate is described. (orig.)

  14. Mouse lung contains endothelial progenitors with high capacity to form blood and lymphatic vessels

    Directory of Open Access Journals (Sweden)

    Barleon Bernhard

    2010-07-01

    Full Text Available Abstract Background Postnatal endothelial progenitor cells (EPCs have been successfully isolated from whole bone marrow, blood and the walls of conduit vessels. They can, therefore, be classified into circulating and resident progenitor cells. The differentiation capacity of resident lung endothelial progenitor cells from mouse has not been evaluated. Results In an attempt to isolate differentiated mature endothelial cells from mouse lung we found that the lung contains EPCs with a high vasculogenic capacity and capability of de novo vasculogenesis for blood and lymph vessels. Mouse lung microvascular endothelial cells (MLMVECs were isolated by selection of CD31+ cells. Whereas the majority of the CD31+ cells did not divide, some scattered cells started to proliferate giving rise to large colonies (> 3000 cells/colony. These highly dividing cells possess the capacity to integrate into various types of vessels including blood and lymph vessels unveiling the existence of local microvascular endothelial progenitor cells (LMEPCs in adult mouse lung. EPCs could be amplified > passage 30 and still expressed panendothelial markers as well as the progenitor cell antigens, but not antigens for immune cells and hematopoietic stem cells. A high percentage of these cells are also positive for Lyve1, Prox1, podoplanin and VEGFR-3 indicating that a considerabe fraction of the cells are committed to develop lymphatic endothelium. Clonogenic highly proliferating cells from limiting dilution assays were also bipotent. Combined in vitro and in vivo spheroid and matrigel assays revealed that these EPCs exhibit vasculogenic capacity by forming functional blood and lymph vessels. Conclusion The lung contains large numbers of EPCs that display commitment for both types of vessels, suggesting that lung blood and lymphatic endothelial cells are derived from a single progenitor cell.

  15. Structural model testing for prestressed concrete pressure vessels: a study of grouted vs nongrouted posttensioned prestressing tendon systems

    International Nuclear Information System (INIS)

    Naus, D.J.

    1979-04-01

    Nongrouted tendons are predominantly used in this country as the prestressing system for prestressed concrete pressure vessels (PCPVs) because they are more easily surveyed to detect reductions in prestressing level and distress such as results from corrosion. Grouted tendon systems, however, offer advantages which may make them cost-effective for PCPV applications. Literature was reviewed to (1) provide insight on the behavior of grouted tendon system, (2) establish performance histories for structures utilizing grouted tendons, (3) examine corrosion protection procedures for prestressing tendons, (4) identify arguments for and against using grouted tendons, and (5) aid in the development of the experimental investigation. The experimental investigation was divided into four phases: (1) grouted-nongrouted tendon behavior, (2) evaluation of selected new material systems, (3) bench-scale corrosion studies, and (4) preliminary evaluation of acoustic emission techniques for monitoring grouted tendons in PCPVs. The groutability of large tendon systems was also investigated

  16. Preliminary calculation with code CONTEMPT-LT for spray cooling tests with JAERI model containment vessel

    International Nuclear Information System (INIS)

    Tanaka, Mitsugu

    1978-01-01

    LWR plants have a containment spray system to reduce the escape of radioactive material to the environment in a loss-of-coolant accident (LOCA) by washing out fission products, especially radioiodine, and condensing the steam to lower the pressure. For carrying out the containment spray tests, pressure and temperature behaviour of the JAERI Model Containment Vessel in spray cooling has been calculated with computer program CONTEMPT-LT. The following could be studied quantitatively: (1) pressure and temperature raise rates for steam addition rate and (2) pressure fall rate for spray flow rate and spray heat transfer efficiency. (auth.)

  17. Probabilistic evaluation of concrete containment capacity for beyond design basis internal pressures

    International Nuclear Information System (INIS)

    Tang, H.T.; Dameron, R.A.; Rashid, Y.R.

    1995-01-01

    For beyond design basis internal pressure loading, experimental studies have demonstrated that the most probable failure mode governing the ultimate functional capacity of concrete containments is leak rather than break. Based on leak rates measured in experiments, a prediction formula for leak rate as functions of containment liner size and internal pressure has been postulated. The determination of liner tear is cast in a probabilistic framework. In calculating leakage, particular attention is paid to the evaluation of leakage versus rupture and the loading rates that may be required to leapfrog over a leakage mode. (orig.)

  18. Review of inservice inspections of greased tendons in prestressed-concrete containments

    International Nuclear Information System (INIS)

    Dougan, J.R.; Ashar, H.

    1983-01-01

    Prestressed-concrete containments in the United States using greased prestressing tendons are inspected periodically to ensure structural integrity and to identify and correct problem areas before they become critical. An analysis of the available utility inspection data and an evaluation of the current and proposed guidelines were conducted to provide a measure of the reliability of the inspection process. Comments from utility and industry personnel were factored into the analysis. The results indicated that the majority of the few incidences of problems or abnormalities which occurred were minor in nature and did not threaten the structural integrity of the containment

  19. Improvement of impact-resistance of a nuclear containment building using fiber reinforced concrete

    International Nuclear Information System (INIS)

    Jeon, Se-Jin; Jin, Byeong-Moo

    2016-01-01

    Highlights: • Impact-resistance of a structure can be improved by fiber reinforced concrete (FRC). • Material modeling of FRC is incorporated into finite element analysis of a structure. • A new index for impact-resistance is proposed based on plastic dissipation energy. • A nuclear power plant made of FRC shows improved resistance against aircraft crashes. - Abstract: Since the act of terrorism that occurred in the USA on September 11, 2001, the protection of nuclear power plants against large commercial aircraft crashes has been an emerging issue. Besides the verification of the safety of nuclear power plants in operation or in design, efficient methods for improving the impact-resistance of these structures have been investigated. Fiber reinforced concrete (FRC) has been generally accepted as an effective material for this purpose. In particular, FRC has been developed to improve the tensile behavior of concrete such as tensile strength, ductility and toughness. One of the main fields of application of FRC can be found in blast-protective or blast-resistant concrete structures. It is expected, therefore, that safety-related structures in a nuclear power plant can also be effectively protected from external blast, aircraft crash, etc. by applying FRC. In order to analytically verify the effect on structural behavior of applying FRC, the particular material properties of FRC should be incorporated into the material modeling of a structural analysis program. This study investigates the mathematical modeling of FRC, which represents various aspects of material behavior. Two numerical examples are provided to show the improved impact-resistance of a nuclear containment building that is expected when applying FRC in comparison with ordinary concrete. The analysis results show that the displacement decreases by 43–67% while the impact-resistance increases by 40–82%, depending on a fiber type.

  20. Improvement of impact-resistance of a nuclear containment building using fiber reinforced concrete

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Se-Jin, E-mail: conc@ajou.ac.kr [Ajou University, 206, World cup-ro, Yeongtong-gu, Suwon-si, Gyeonggi-do 16499 (Korea, Republic of); Jin, Byeong-Moo [DAEWOO E& C, Institute of Construction Technology, 20, Suil-ro 123beon-gil, Jangan-gu, Suwon-si, Gyeonggi-do 16297 (Korea, Republic of)

    2016-08-01

    Highlights: • Impact-resistance of a structure can be improved by fiber reinforced concrete (FRC). • Material modeling of FRC is incorporated into finite element analysis of a structure. • A new index for impact-resistance is proposed based on plastic dissipation energy. • A nuclear power plant made of FRC shows improved resistance against aircraft crashes. - Abstract: Since the act of terrorism that occurred in the USA on September 11, 2001, the protection of nuclear power plants against large commercial aircraft crashes has been an emerging issue. Besides the verification of the safety of nuclear power plants in operation or in design, efficient methods for improving the impact-resistance of these structures have been investigated. Fiber reinforced concrete (FRC) has been generally accepted as an effective material for this purpose. In particular, FRC has been developed to improve the tensile behavior of concrete such as tensile strength, ductility and toughness. One of the main fields of application of FRC can be found in blast-protective or blast-resistant concrete structures. It is expected, therefore, that safety-related structures in a nuclear power plant can also be effectively protected from external blast, aircraft crash, etc. by applying FRC. In order to analytically verify the effect on structural behavior of applying FRC, the particular material properties of FRC should be incorporated into the material modeling of a structural analysis program. This study investigates the mathematical modeling of FRC, which represents various aspects of material behavior. Two numerical examples are provided to show the improved impact-resistance of a nuclear containment building that is expected when applying FRC in comparison with ordinary concrete. The analysis results show that the displacement decreases by 43–67% while the impact-resistance increases by 40–82%, depending on a fiber type.

  1. Properties of concrete containing different type of waste materials as aggregate replacement exposed to elevated temperature – A review

    Science.gov (United States)

    Ghadzali, N. S.; Ibrahim, M. H. W.; Sani, M. S. H. Mohd; Jamaludin, N.; Desa, M. S. M.; Misri, Z.

    2018-04-01

    Concrete is the chief material of construction and it is non-combustible in nature. However, the exposure to the high temperature such as fire can lead to change in the concrete properties. Due to the higher temperature, several changes in terms of mechanical properties were observed in concrete such as compressive strength, modulus of elasticity, tensile strength and durability of concrete will decrease significantly at high temperature. The exceptional fire-proof achievement of concrete is might be due to the constituent materials of concrete such as its aggregates. The extensive use of aggregate in concrete will leads to depletion of natural resources. Hence, the use of waste and other recycled and by-product material as aggregates replacements becomes a leading research. This review has been made on the utilization of waste materials in concrete and critically evaluates its effects on the concrete performances during the fire exposure. Therefore, the objective of this paper is to review the previous search work regarding the concrete containing waste material as aggregates replacement when exposed to elevated temperature and come up with different design recommendations to improve the fire resistance of structures.

  2. Experimental investigation on the properties of concrete containing post-consumer plastic waste as coarse aggregate replacement

    Directory of Open Access Journals (Sweden)

    Zasiah TAFHEEM

    2018-03-01

    Full Text Available The consumption of various forms of plastic has been increased in recent days due to the boost in industrialization and other human activities. Most of the plastic wastes are abandoned and require large landfill area for storage. More importantly, the low biodegradability of plastic poses a serious threat to environment protection issue. Various methods have been followed for the disposal of plastic in an attempt to reduce the negative impact of the plastic on the environment. Recently, various types of plastic have been incorporated in concrete to minimize the exposure of plastic to the environment. The aim of this study is to investigate the properties of concrete containing polyethylene terephthalate (PET, and high density polyethylene (HDPE plastic that were used as partial replacement of coarse aggregate (CA. In this study, four compositions of stone aggregate(S: plastic waste ratios have been used by volume basis: 100% S: 0% Plastic (control concrete, 90% S: 10% PET, 90% S: 10% HDPE, and 90% S: 5% PET+5% HDPE. The effects of waste plastic addition on the mechanical properties of concrete are presented in this paper. Test results reveal that minimum reduction in compressive strength has been found 35% in case of 10% PET plastic replaced concrete whereas splitting tensile strength for 10% PET replaced concrete has been increased by 21% while compared to control concrete. In addition, fresh unit weight of concrete containing plastic waste has been decreased by 4% in comparison to control concrete.

  3. Study on the application of thickened welds without post weld heat treatment for containment vessels

    International Nuclear Information System (INIS)

    Takeuchi, T.; Fukaya, T.; Sato, M.; Takano, G.

    1978-01-01

    As material for containment vessels, SGV49 steel plates are mainly used. However, those used for this purpose are limited in thickness to smaller than 38 mm. This is because the present standard requires welds thicker than 38 mm to be subjected to post weld heat treatment but operation on the site is practically difficult. In the case of 3-loop containment vessels of pressurized water type reactors, use of 38 mm SGV49 brings an increase in their height and this is disadvantageous from a seismic viewpoint. Therefore, use of 45 mm-thick steel material has become necessary in order to increase design internal pressure and reduce the height of the vessels. To investigate the propriety of the use of 45 mm-thick SGV49 for this purpose without post weld heat treatment we investigated the basic performances of base metal and welded joints. We also conducted large-scale embrittlement fracture tests (CT test, deep notch test, wide plate tensile test and ESSO test) in order to examine whether welds not subjected to post weld heat treatment are safe against embrittlement fracture under the operating conditions of the vessels. The results proved that the welds of SGV49 steel plates are safe enough under the operating conditions. (author)

  4. Development of improved SGV480 steel plate for containment vessel in PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, Norioki [Advanced Nuclear Equipment Research Inst., Tokyo (Japan); Morikage, Yasushi; Okayama, Yutaka; Higashikubo, Tomohiro

    2001-01-01

    When a nuclear containment vessel made of steel plate at PWR plants in Japan is produced, SGV480 steel plate made by annealing method according to JIS G3118 is usually used in main. And, when thickness of welding portion of the vessel is larger than 38 mm, as heat treatment after welding is regulated to carry out according to the ministerial ordinance, it is difficult in actual to carry out the heat treatment of the actual welded portions. In a leading plant, approval of welding using a special method without heat treatment less than 47.25 mm of SGV480 carbon steel plate for JIS G3118 middle and ordinary pressure vessel was carried out to supply it for actual use. And, it is required for protection of welding fracture to carry out pre-heat treatment before welding. Because of increasing plate thickness requiring for lower temperature and more seismic resistance in construction condition, in order to produce a containment vessel without heat treatment after welding, more toughness is required for using material and welded portion. Therefore, a new SGV480 steel plate was developed by using TMCP method of modern steel manufacturing technology, to establish lower carbon equivalence and finer texture with upgrading of both toughness and weldability, without heat treatment after welding and pre-heat treatment before welding, at the Shin-Nippon Steel Co, Ltd. and Kawasaki Steel, Co. Ltd., respectively. (G.K.)

  5. Influence of Steel Fibers on the Structural Performance of a Prestressed Concrete Containment Building

    International Nuclear Information System (INIS)

    Choun, Youngsun; Hahm, Daegi; Park, Junhee

    2013-01-01

    A large number of previous experimental investigations indicate that the use of steel fibers in conventional reinforced concrete (RC) can enhance the structural and functional performance of prestressed concrete containment buildings (PCCBs) in nuclear power plants. A prevention of through-wall cracks and an increase of the post-cracking ductility will improve the ultimate internal pressure capacity, and a high shear resistance under cyclic loadings will increase the seismic resisting capacity. In this study, the effects of steel fiber reinforced concrete (SFRC) on the ultimate pressure and seismic capacities of a PCCB are investigated. The effects of steel fibers on the ultimate pressure and shear resisting capacities of a PCCB are investigated. It is revealed that both of the ultimate pressure capacity and the shear resisting capacity of a PCCB can be greatly enhanced by introducing steel fibers in a conventional RC. Estimation results indicate that the ultimate pressure capacity and maximum lateral displacement of a PCCB can be improved by 16% and 64%, respectively, if a conventional RC contains hooked steel fibers in a volume fraction of 1.0%

  6. Inspection of a large concrete block containing embedded defects using ground penetrating radar

    Science.gov (United States)

    Eisenmann, David; Margetan, Frank J.; Koester, Lucas; Clayton, Dwight

    2016-02-01

    Ground penetrating radar (GPR), also known as impulse response radar, was used to examine a thick concrete block containing reinforcing steel bars (rebar) and embedded defects. The block was located at the University of Minnesota, measured approximately 7 feet tall by 7 feet wide by 40 inches deep, and was intended to simulate certain aspects of a concrete containment wall at a nuclear power plant. This paper describes the measurements that were made and various analyses of the data. We begin with a description of the block itself and the GPR equipment and methods used in our inspections. The methods include the application of synthetic aperture focusing techniques (SAFT). We then present and discuss GPR images of the block's interior made using 1600-MHz, 900-MHz, and 400-MHz antennas operating in pulse/echo mode. A number of the embedded defects can be seen, and we discuss how their relative detectability can be quantified by comparison to the response from nearby rebar. We next discuss through-transmission measurements made using pairs of 1600-MHz and 900-MHz antennas, and the analysis of that data to deduce the average electromagnetic (EM) wave speed and attenuation of the concrete. Through the 40-inch thickness, attenuation rises approximately linearly with frequency at a rate near 0.7 dB/inch/GHz. However, there is evidence that EM properties vary with depth in the block. We conclude with a brief summary and a discussion of possible future work.

  7. Investigation of Properties of Asphalt Concrete Containing Boron Waste as Mineral Filler

    Directory of Open Access Journals (Sweden)

    Cahit GÜRER

    2016-05-01

    Full Text Available During the manufacture of compounds in the boron mining industry a large quantity of waste boron is produced which has detrimental effects on the environment. Large areas have to be allocated for the disposal of this waste. Today with an increase in infrastructure construction, more efficient use of the existing sources of raw materials has become an obligation and this involves the recycling of various waste materials. Road construction requires a significant amount of raw materials and it is possible that substantial amounts of boron-containing waste materials can be recycled in these applications. This study investigates the usability of boron wastes as filler in asphalt concrete. For this purpose, asphalt concrete samples were produced using mineral fillers containing 4%, 5%, 6%, 7% and 8% boron waste as well as a 6% limestone filler (6%L as the control sample. The Marshall Design, mechanical immersion and Marshall Stability test after a freeze-thaw cycle and indirect tensile stiffness modulus (ITSM test were performed for each of the series. The results of this experimental study showed that boron waste can be used in medium and low trafficked asphalt concrete pavements wearing courses as filler.

  8. Influence of Steel Fibers on the Structural Performance of a Prestressed Concrete Containment Building

    Energy Technology Data Exchange (ETDEWEB)

    Choun, Youngsun; Hahm, Daegi; Park, Junhee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    A large number of previous experimental investigations indicate that the use of steel fibers in conventional reinforced concrete (RC) can enhance the structural and functional performance of prestressed concrete containment buildings (PCCBs) in nuclear power plants. A prevention of through-wall cracks and an increase of the post-cracking ductility will improve the ultimate internal pressure capacity, and a high shear resistance under cyclic loadings will increase the seismic resisting capacity. In this study, the effects of steel fiber reinforced concrete (SFRC) on the ultimate pressure and seismic capacities of a PCCB are investigated. The effects of steel fibers on the ultimate pressure and shear resisting capacities of a PCCB are investigated. It is revealed that both of the ultimate pressure capacity and the shear resisting capacity of a PCCB can be greatly enhanced by introducing steel fibers in a conventional RC. Estimation results indicate that the ultimate pressure capacity and maximum lateral displacement of a PCCB can be improved by 16% and 64%, respectively, if a conventional RC contains hooked steel fibers in a volume fraction of 1.0%.

  9. Analysis of initial prestress force of spatial tendon prestressed concrete containment structures

    International Nuclear Information System (INIS)

    Shiau, H.-S.

    1975-01-01

    A theoretical investigation is presented of the initial stage of prestressed tendon and prestressed concrete before and after jacking force of tendon anchorage released. A method is developed that is applicable to any kind of spatial tendon considering frictional loss due to length and curvature effects. A triple integral equation of one independent variable and jacking force is derived to represent an exact solution of tendon force along the whole tendon which may have reverse curvatures. In order to analyze the stress response of concrete due to this prestress force by using existing finite element computer program or any other kind of computer program, a systematic method is suggested to obtain tendon force components, which are represented by a series of equations of one independent variable, in any coordinate system as external force applied on the concrete. The resulting systems of the equations are then solved by numerical mathematic and computer techniques. Two numerical examples are represented. The first example is, dome prestress analysis of containment building by the proposed method and Kalnins' computer program for shell of revolution. Results are discussed. The second example is picked from prestress analysis for personnel air lock of containment building by using proposed method and FELAP finite element Computer program. It includes two different tendon arrangements around the opening. The results of these two different arrangements are compared and discussed

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

    International Nuclear Information System (INIS)

    Stiefel, M.

    1983-01-01

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

  11. Design criteria for the structural analysis of shipping cask containment vessels

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

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

  12. Current state of knowledge on the behavior of steel liners in concrete containments subjected to overpressurization loads

    International Nuclear Information System (INIS)

    Riesemann, W.A. von; Parks, M.B.

    1995-01-01

    In the US, concrete containment buildings for commercial nuclear power plants have steel liners that act as the internal pressure boundary. The liner abuts the concrete, acting as the interior concrete form. The liner is attached to the concrete by either studs or by a continuous structural shape (such as a T-section or channel) that is either continuously or intermittently welded to the liner. Studs are commonly used in reinforced concrete containments, while prestressed containments utilize a structural element as the anchorage. The practice in some countries follows the US practice, while in other countries the containment does not have a steel liner. In this latter case, there is a true double containment, and the annular region between the two containments is vented.This paper will review the practice of design of the liner system prior to the consideration of severe accident loads (overpressurization loads beyond the design conditions).An overpressurization test of a 1:6 scale reinforced concrete containment at Sandia National Laboratories resulted in a failure mechanism in the liner that was not fully anticipated. Post-test analyses and experiments have been conducted to understand the failure better. This work and the activities that followed the test are reviewed. Areas in which additional research should be conducted are given. (orig.)

  13. Design, analysis and construction of the prestressed concrete containment of the nuclear power station Gundremmingen

    International Nuclear Information System (INIS)

    Mueller, W.F.; Ick, U.

    1977-01-01

    Kraftwerk Union AG is presently constructing at Gundremmingen (Bavaria) on the River Danube a BWR twin-plant (KRB Units B and C) with a capacity of 2x1300 MWe. Owing to the wall thickness/diameter ratio the containment can be calculated as a thin-walled shell. Areas of discontinuity are subjected to three-dimensional investigations. For the design of the concrete structure different fracture safety margins are defined for the load conditions occurring in operation in the event of a loss-of-coolant accident and as a result of an aircraft or an earthquake. From this results that in the cross sectional areas without discontinuities of the prestressed outer cylinder no resultant tensions occur. For the steel liner different limits of strain are permitted for the various load conditions, bearing in mind that the integrity of the liner must remain ensured at any time. In order to keep the stresses resulting from the constraint of the containment outer cylinder in the foundation slab low, the cylindrical wall is placed on bearings. The suppression pool top slab is constrained at the containment outer cylinder and at the containment inner cylindrical wall. The inlets of the vent pipes are integrated in the slab in a way resulting in a double slab. The liner consists of 8 mm thick steel plate and is anchored in the concrete via steel sections. Mechanical equipment anchoring in the concrete is provided by welding anchor plates into the liner after the section concerned has been completed. The carcass work on the reactor building is scheduled to be completed within

  14. Experiences in development, qualification, and use of concrete high-integrity containers in commercial disposal of radioactive wastes

    International Nuclear Information System (INIS)

    Schmitt, R.C.; Reno, H.W.

    1985-01-01

    Disposal of EPICOR prefilters as commercial radioactive wastes is being accomplished by using a first-of-a-kind, reinforced concrete, high-integrity container in lieu of prior in situ solidification of resins before disposal of prefilters. Experiences in developing, testing, certifying, and using high-integrity containers are an untold story worthy of review for the benefit of the nuclear industry at large. The lessons learned in gaining regulatory acceptance of the concrete HIC are discussed

  15. Experiences in development, qualification, and use of concrete high-integrity containers in commercial disposal of radioactive wastes

    International Nuclear Information System (INIS)

    Schmitt, R.C.; Reno, H.W.

    1985-01-01

    Disposal of EPICOR prefilters as commercial radioactive wastes is being accomplished by using a first-of-a-kind, reinforced concrete, high-integrity container (HIC) in lieu of prior in situ solidification of resins before disposal of prefilters. Experiences in developing, testing, certifying, and using high-integrity containers are an untold story worthy of review for the benefit of the nuclear industry at large. The lessons learned in gaining regulatory acceptance of the concrete HIC are discussed. 6 refs., 1 tab

  16. Axisymmetric global structural analysis of BARC prestressed concrete containment model for beyond design pressure

    International Nuclear Information System (INIS)

    Singh, Tarvinder; Singh, R.K.; Ghosh, A.K.

    2008-10-01

    In order to check the adequacy of the Indian Pressurized Heavy Water Reactor (PHWR) containment structure to withstand severe accident induced internal pressure load, the ultimate load capacity assessment is required. Reactor Safety Division (RSD) of Bhabha Atomic Research Centre (BARC) has initiated an experimental program at BARC Tarapur Containment Test Facility to evaluate the ultimate load capacity of Indian PHWR containment. For this study, BARC Containment Model (BARCOM), which is 1:4 scale representation of Tarapur Atomic Power Station (TAPS) unit-3 and 4 540 MWe PHWR Inner Containment of Pre-stressed Concrete has been constructed. The model includes all the important major design features of the prototype containment and simulates Main Air Lock (MAL), Steam Generator (SG), Emergency Air Lock (EAL) and Fueling Machine Air Lock (FMAL) openings. The design pressure (Pd) of BARCOM is 1.44kg/cm 2 (g), which is same as the prototype. The pretest analysis of BARCOM has been performed with finite element axi-symmetric modeling. The objective of this simulation was to understand the behavior of containment model under internal pressure and find out the various failure modes and critical locations important for instrumentation during the experiment. The structural response of the containment model is assessed in terms of wall and dome displacement; cracking of concrete, longitudinal and hoop strains and stresses. Another objective of the analysis was to predict the various failure modes of BARCOM with regard to the concrete cracking, reinforcement yielding and tendon inelastic behavior along with the estimation of the ultimate load capacity of the containment model. It is noted that the BARCOM has an ultimate load capacity factor of 3.54 Pd. However, further analysis is needed to quantify the factor of safety with detail 3D model, which should account for the local structural behavior due to various openings. Meanwhile, this preliminary simplified analysis helps to

  17. Use of variational principles for solution of infinitely redundant continuum problem with special reference to containment vessels

    International Nuclear Information System (INIS)

    Stefanou, G.D.

    1982-01-01

    The calculation of time-deepndent stresses in concrete structures operating at elevated temperatures is discussed. The method described is of a direct formulation technique and it is based on the principles of the calculus of variation. The paper mainly deals with the application of the method to a large and infinitely redundant continuum problem. The analytical procedure of the variational principle is also described and the mathematical expressions are developed for uniaxial and biaxial stress problems. The solution for the biaxial state of stress is carried out by a two-dimensional finite element stiffness analysis. A step-by-step method developed by the author using two-dimensional finite element stiffness analysis is also described in APPENDIX 3. Both methods are then applied to a real problem for which experimental data exist from Ref. (1) Predicted analytical values obtained by both methods are compared with experimental results. The method is suitable for predicting the distribution of stress in the end slabs of containment vessels. These slabs are perforated to permit fuel loading by the charging machine. (author)

  18. Pore structure modification of cement concretes by impregnation with sulfur-containing compounds

    Directory of Open Access Journals (Sweden)

    YANAKHMETOV Marat Rafisovich

    2015-02-01

    Full Text Available The authors study how the impregnation with sulfur-containing compounds changes the concrete pore structure and how it influences on the water absorption and watertightness. The results of this research indicate that impregnation of cement concrete with water-based solution of polysulphide modifies pore structure of cement concrete in such a way that it decreases total and effective porosity, reduces water absorption and increases watertightness. The proposed impregnation based on mineral helps to protect for a long time the most vulnerable parts of buildings – basements, foundations, as well as places on the facades of buildings exposed to rain, snow and groundwater. Application of the new product in the construction industry can increase the durability of materials, preventing the destruction processes caused by weathering, remove excess moisture in damp basements. The surfaces treated by protective compounds acquire antisoiling properties for a long time, and due to reduced thermal conductivity the cost of heating buildings is decreased. The effectiveness of the actions and the relatively low cost of proposed hydrophobizator makes it possible to spread widely the proposed protection method for building structures.

  19. Mechanical Properties of High Strength Concrete Containing Coal Bottom Ash and Oil-Palm Boiler Clinker as Fine Aggregates

    Directory of Open Access Journals (Sweden)

    Soofinajafi Mahmood

    2016-01-01

    Full Text Available This research aims to utilize Coal Furnace Bottom ash (CBA and Oil-Palm Boiler Clinker (OPBC as fine aggregate in concrete mix proportions. They are solid wastes from power plant and Oil Palm industry, respectively. Since these by-products do not have any primary use and are pure waste, an opportunity to use them as aggregate in concrete industry not only is economical but also will be an environmental friendly opportunity leading towards a more sustainable production chain. CBA and OPBC sands had similar grading to normal sand but have lower density and higher water absorption. In a high strength concrete, normal sand was replaced up to 25% with either CBA or OPBC. Test results showed that although water absorption of these wastes was more than normal sand but the slump value of concrete containing each of these wastes showed that these concretes had good workability. All mixes containing these wastes had slightly lower compressive strength at early ages and equivalent or higher compressive strength at later ages compared to control mix. The 28-day compressive strength of these concretes was in the range of 69–76 MPa which can be categorized as high strength concrete. In general, the performance of OPBC was better than CBA at 25% replacement level. However, it is recommended that at least 12.5% of total volume of fine aggregate in a high strength concrete is used of CBA or OPBC.

  20. Device for the simultaneous operation of the closing valve of a vessel and the closing valve of a transport container

    International Nuclear Information System (INIS)

    Tellier, Claude; Surriray, Michel.

    1982-01-01

    This device includes mechanisms for unlatching the closing valve of the vessel and securing it to the closing valve of the transport container and other mechanisms for vertically raising the assembly of valves, pivoting it and bringing it into a vertical position in a bulge provided in the bottom of the transport container. For example the first containment is a nuclear reactor vessel and the transport container is used for carrying an item from the vessel to an external area (for instance, a defective pump to the repair area) and for the return transport operation [fr

  1. Investigating the Properties of Asphalt Concrete Containing Glass Fibers and Nanoclay

    Directory of Open Access Journals (Sweden)

    Hasan Taherkhani

    2016-06-01

    Full Text Available The performance of asphaltic pavements during their service life is highly dependent on the mechanical properties of the asphaltic layers. Therefore, in order to extend their service life, scientists and engineers are constantly trying to improve the mechanical properties of the asphaltic mixtures. One common method of improving the performance of asphaltic mixtures is using different types of additives. This research investigated the effects of reinforcement by randomly distributed glass fibers and the simultaneous addition of nanoclayon some engineering properties of asphalt concrete have been investigated. The properties of a typical asphalt concrete reinforced by different percentages of glass fibers were compared with those containing both the fibers and nanoclay. Engineering properties, including Marshall stability, flow, Marshall quotient, volumetric properties and indirect tensile strength were studied. Glass fibers were used in different percentages of 0.2, 0.4 and 0.6% (by weight of total mixture, and nanoclay was used in 2, 4 and 6% (by the weight of bitumen. It was found that the addition of fibers proved to be more effective than the nanoclay in increasing the indirect tensile strength. However, nanoclay improved the resistance of the mixture against permanent deformation better than the glass fibers. The results also showed that the mixture reinforced by 0.2% of glass fiber and containing 6% nanoclay possessed the highest Marshall quotient, and the mixture containing 0.6% glass fibers and 2% nanoclay possessedthe highest indirect tensile strength.

  2. Self-consolidating concretes containing waste PET bottles as sand replacement

    Science.gov (United States)

    Khalid, Faisal Sheikh; Azmi, Nurul Bazilah; Mazenan, Puteri Natasya; Shahidan, Shahiron; Othman, Nor hazurina; Guntor, Nickholas Anting Anak

    2018-02-01

    This study evaluates the effect of self-consolidating concrete (SCC) containing waste polyethylene terephthalate (PET) granules on the fresh, mechanical and water absorption properties. Fine aggregates were replaced from 0% to 8% by PET granules. The fresh properties of SCC containing PET granules were determined using slump flow and V-funnel flow time tests. The compressive and splitting tensile strength were evaluated. The results indicated that utilization of waste PET granules in production of SCC could be an effective way for recycling purpose. The maximum amount of PET replacement should be limited to 5%. Exceeding 5% of PET content may result in an increase of V-funnel flow time to overpass the limiting value, decrease in strength. The production of high performance SCC containing 5% PET granules satisfies all the requirements for SCC with satisfactory outputs.

  3. Analytical predictions for the performance of a reinforced concrete containment model subject to overpressurization

    International Nuclear Information System (INIS)

    Weatherby, J.R.; Clauss, D.B.

    1987-01-01

    Under the sponsorship of the US Nuclear Regulatory Commission, Sandia National Laboratories is investigating methods for predicting the structural performance of nuclear reactor containment buildings under hypothesized severe accident conditions. As part of this program, a 1/6th-scale reinforced concrete containment model will be pressurized to failure in early 1987. Data generated by the test will be compared to analytical predictions of the structural response in order to assess the accuracy and reliability of the analytical techniques. As part of the pretest analysis effort, Sandia has conducted a number of analyses of the containment structure using the ABAQUS general purpose finite element code. This paper describes results from a nonlinear axisymmetric shell analysis as well as the material models and failure criteria used in conjunction with the analysis

  4. Creep deformation and crack growth in a low alloy steel welded pressure vessel containing defects

    International Nuclear Information System (INIS)

    Coleman, M.C.

    1982-01-01

    A full-size pressure vessel was tested for effects of welding residual stresses on creep deformation and crack growth. The vessel, based on 1/2 Cr 1/2 Mo 1/4 V main steam pipe, contained four 2CrMo manual metal arc welds, two in the as-welded condition and two stress-relieved. All the welds contained pre-existing defects machined in the heat affected zones. Testing was carried out at two internal steam pressures, 250 and 350 bar, and 565 0 C. Cracked and uncracked areas of the vessel were monitored continuously. Results are presented for the continuous creep deformation observed in both the hoop and axial directions of the welds throughout the 11,400 h of testing, as well as the intermittent strain data obtained during inspections. Crack growth observations are described based on nondestructive examination. The residual stresses measured are also given for both the as-welded and stress relieved weldments. Results obtained are discussed in terms of the effects of welding residual stress on the hoop and axial deformations observed in the welds. Similarly, the effects of residual stress on creep crack growth are considered together with compositional and microstructural implications. 9 figures, 5 tables

  5. Large scale model experimental analysis of concrete containment of nuclear power plant strengthened with externally wrapped carbon fiber sheets

    International Nuclear Information System (INIS)

    Yang Tao; Chen Xiaobing; Yue Qingrui

    2005-01-01

    Concrete containment of Nuclear Power Station is the last shield structure in case of nuclear leakage during an accident. The experiment model in this paper is a 1/10 large-scale model of a real-sized prestressed reinforced concrete containment. The model containment was loaded by hydraulic pressure which simulated the design pressure during the accident. Hundreds of sensors and advanced data-collect systems were used in the test. The containment was first loaded to the damage pressure then strengthened with externally wrapping Carbon fiber sheet around the outer surface of containment structure. Experimental results indicate that CFRP system can greatly increase the capacity of concrete containment to endure the inner pressure. CFRP system can also effectively confine the deformation and the cracks caused by loading. (authors)

  6. Evaluation charts of thermal stresses in cylindrical vessels induced by thermal stratification of contained fluid

    International Nuclear Information System (INIS)

    Furuhashi, Ichiro; Kawasaki, Nobuchika; Kasahara, Naoto

    2008-01-01

    Temperature and thermal stress in cylindrical vessels were analysed for the thermal stratification of contained fluid. Two kinds of temperature analysis results were obtained such as the exact temperature solution of eigenfunction series and the simple approximate one by the temperature profile method. Furthermore, thermal stress shell solutions were obtained for the simple approximate temperatures. Through comparison with FEM analyses, these solutions were proved to be adequate. The simple temperature solution is described by one parameter that is the temperature decay coefficient. The thermal stress shell solutions are described by two parameters. One is the ratio between the temperature decay coefficient and the load decay coefficient. Another is the nondimensional width of stratification. These solutions are so described by few parameters that those are suitable for the simplified thermal stress evaluation charts. These charts enable quick and accurate thermal stress evaluations of cylindrical vessel of this problem compared with conventional methods. (author)

  7. Evaluation charts of thermal stresses in cylindrical vessels induced by thermal stratification of contained fluid

    International Nuclear Information System (INIS)

    Furuhashi, Ichiro; Kawasaki, Nobuchika; Kasahara, Naoto

    2007-01-01

    Temperature and thermal stress in cylindrical vessels were analysed for the thermal stratification of contained fluid. Two kinds of temperature analysis results were obtained such as the exact temperature solution of eigen-function series and the simple approximate one by the temperature profile method. Furthermore, shell solutions of thermal stress were obtained for the simple approximate temperatures. Through comparison with FEM analyses, these solutions were proved to be adequate. The simple temperature solution is described by one parameter that is the temperature decay factor. The shell solutions of thermal stress are described by two parameters. One is the ratio between the temperature decay factor and the local decay factor. Another is the non-dimensional width of stratification. These solution are so described by few parameters that those are suitable for the simplified thermal stress evaluation charts. These charts enable quick and accurate thermal stress evaluations of cylindrical vessel of this problem compared with conventional methods. (author)

  8. Evaluation of Shear Resisting Capacity of a Prestressed Concrete Containment Building with Steel or Polyamide Fiber Reinforcement

    International Nuclear Information System (INIS)

    Choun, Youngsun; Park, Junhee

    2014-01-01

    Conventional reinforced concrete (RC) members generally show a rapid deterioration in shear resisting mechanisms under a reversed cyclic load. However, the use of high-performance fiber-reinforced cement composites provides excellent damage tolerance under large displacement reversals compared with regular concrete. Previous experimental studies have indicated that the use of fibers in conventional RC can enhance the structural and functional performance of prestressed concrete containment buildings (PCCBs) in nuclear power plants. This study evaluates the shear resisting capacity for a PCCB constructed using steel fiber reinforced concrete (SFRC) or polyamide fiber reinforced concrete (PFRC). The effects of steel and polyamide fibers on the shear performance of a PCCB were investigated. It was revealed that steel fibers are more effective to enhance the shear resisting capacity of a PCCB than polyamide fibers. The ductility and energy dissipation increase significantly in fiber reinforced PCCBs

  9. Evaluation of Shear Resisting Capacity of a Prestressed Concrete Containment Building with Steel or Polyamide Fiber Reinforcement

    Energy Technology Data Exchange (ETDEWEB)

    Choun, Youngsun; Park, Junhee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Conventional reinforced concrete (RC) members generally show a rapid deterioration in shear resisting mechanisms under a reversed cyclic load. However, the use of high-performance fiber-reinforced cement composites provides excellent damage tolerance under large displacement reversals compared with regular concrete. Previous experimental studies have indicated that the use of fibers in conventional RC can enhance the structural and functional performance of prestressed concrete containment buildings (PCCBs) in nuclear power plants. This study evaluates the shear resisting capacity for a PCCB constructed using steel fiber reinforced concrete (SFRC) or polyamide fiber reinforced concrete (PFRC). The effects of steel and polyamide fibers on the shear performance of a PCCB were investigated. It was revealed that steel fibers are more effective to enhance the shear resisting capacity of a PCCB than polyamide fibers. The ductility and energy dissipation increase significantly in fiber reinforced PCCBs.

  10. Characterisation of concrete containers for radioactive waste in the engineering tranches system at the Yugoslav R.A waste storing center

    International Nuclear Information System (INIS)

    Plecas, I.; Peric, A.; Drljaca, J.; Kostadinovic, A.

    1987-10-01

    Low and intermediate level radioactive waste represents 90% of total R.A. waste. It is conditioned into special concrete containers. Since these concrete containers are to protect safely the radioactive waste for 300 years, the selection of materials and precise control of their physical and mechanical properties is very important. In this paper results obtained with some concrete compositions are described. (author)

  11. A simple evaluation of containment integrity against ex-vessel steam explosion

    International Nuclear Information System (INIS)

    Nishiura, Hiroshi

    2000-01-01

    The guideline for consideration to severe accidents on containment design for next-generation LWR was published in 1999. In order to verify the validity of future containment designs, we have developed a method of assessing for the containment integrity against ex-vessel steam explosion. First, we conducted a simple evaluation on an Advanced PWR. The strength of the reactor cavity wall was assumed to be equivalent to the total strain energy which would accumulate by the time one reinforcing bar element would first reach the failure strain in FEM analyses. As a result, the strength was evaluated to be about 72 MJ. The explosion energy was assumed to be a function of the mass of the dropping melted core and the conversion ratio. Assuming the conversion ratio of 1%, it was estimated that the explosion energy would amount to about 1 MJ if the melt mass corresponds to the break of one instrumentation guide tube penetration, and about 40 MJ if the mass corresponds to the simultaneous break of all penetrations. Therefore, it is expected that the explosion energy would be less than the wall strength; thus, the containment integrity would be maintained even if an ex-vessel steam explosion were to occur. (author)

  12. Principles of design and construction for the top caps of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

    Hughes, A.N.; Bellwood, G.N.; Paton, A.A.

    1976-01-01

    The building of the top cap poses problems because of the number of penetrations to be cast therein. The fuel and control system routes need to be tightly specified and controlled so that during station life misalignments do not occur which interfere with the fuelling and control operations. The paper outlines the route requirements and illustrates how these affect the tolerances and movements which can be allowed at various stages of construction. Development work is discussed to show the necessity of resolving the different priorities of design, programme and overall pressure vessel construction requirements, so that the reactor build is not inhibited by the special demands of the top cap, and the integration of the monitoring and survey systems during the top cap build are explained. (author)

  13. DHCVIM - a direct heating containment vessel interactions module: applications to Sandia National Laboratories Surtsey experiments

    International Nuclear Information System (INIS)

    Ginsberg, T.; Tutu, N.K.

    1987-01-01

    Direct containment heating is the mechanism of severe nuclear reactor accident containment loading that results from transfer of thermal and chemical energy from high-temperature, finely divided, molten core material to the containment atmosphere. The direct heating containment vessel interactions module (DHCVIM) has been developed at Brookhaven National Laboratory to model the mechanisms of containment loading resulting from the direct heating accident sequence. The calculational procedure is being used at present to model the Sandia National Laboratories one-tenth-scale Surtsey direct containment heating experiments. The objective of the code is to provide a test bed for detailed modeling of various aspects of the thermal, chemical, and hydrodynamic interactions that are expected to occur in three regions of a containment building: reactor cavity, intermediate subcompartments, and containment dome. Major emphasis is placed on the description of reactor cavity dynamics. This paper summarizes the modeling principles that are incorporated in DHCVIM and presents a prediction of the Surtsey Test DCH-2 that was made prior to execution of the experiment

  14. Mechanical properties of concrete containing a high volume of tire-rubber particles.

    Science.gov (United States)

    Khaloo, Ali R; Dehestani, M; Rahmatabadi, P

    2008-12-01

    Due to the increasingly serious environmental problems presented by waste tires, the feasibility of using elastic and flexible tire-rubber particles as aggregate in concrete is investigated in this study. Tire-rubber particles composed of tire chips, crumb rubber, and a combination of tire chips and crumb rubber, were used to replace mineral aggregates in concrete. These particles were used to replace 12.5%, 25%, 37.5%, and 50% of the total mineral aggregate's volume in concrete. Cylindrical shape concrete specimens 15 cm in diameter and 30 cm in height were fabricated and cured. The fresh rubberized concrete exhibited lower unit weight and acceptable workability compared to plain concrete. The results of a uniaxial compressive strain control test conducted on hardened concrete specimens indicate large reductions in the strength and tangential modulus of elasticity. A significant decrease in the brittle behavior of concrete with increasing rubber content is also demonstrated using nonlinearity indices. The maximum toughness index, indicating the post failure strength of concrete, occurs in concretes with 25% rubber content. Unlike plain concrete, the failure state in rubberized concrete occurs gently and uniformly, and does not cause any separation in the specimen. Crack width and its propagation velocity in rubberized concrete are lower than those of plain concrete. Ultrasonic analysis reveals large reductions in the ultrasonic modulus and high sound absorption for tire-rubber concrete.

  15. Reliability assessment and probability based design of reinforced concrete containments and shear walls

    International Nuclear Information System (INIS)

    Hwang, H.; Reich, M.; Ellingwood, B.; Shinozuka, M.

    1986-03-01

    This report summarizes work completed under the program entitled, ''Probability-Based Load Combinations for Design of Category I Structures.'' Under this program, the probabilistic models for various static and dynamic loads were formulated. The randomness and uncertainties in material strengths and structural resistance were established. Several limit states of concrete containments and shear walls were identified and analytically formulated. Furthermore, the reliability analysis methods for estimating limit state probabilities were established. These reliability analysis methods can be used to evaluate the safety levels of nuclear structures under various combinations of static and dynamic loads. They can also be used to generate analytically the fragility data for PRA studies. In addition to the development of reliability analysis methods, probability-based design criteria for concrete containments and shear wall structures have also been developed. The proposed design criteria are in the load and resistance factor design (LRFD) format. The load and resistance factors are determined for several limit states and target limit state probabilities. Thus, the proposed design criteria are risk-consistent and have a well-established rationale. 73 refs., 18 figs., 16 tabs

  16. Study on the application of 50 mm thick welded joints without PWHT for containment vessels

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Nozomu; Sakai, Yoshiyuki; Hayashi, Kazutoshi; Higashikubo, Tomohiro (Mitsubishi Heavy Industries. Ltd., Kobe Shipyard and Machinery Works (Japan)); Iida, Kunihiro (Shibaura Inst. of Tech., Dept. of Mechanical Engineering, Tokyo (Japan)); Satou, Masanobu (Mitsubishi Heavy Industries. Ltd., Tkasago Research and Development Center (Japan))

    1992-01-01

    In order to investigate the propriety of the use of 50 mm thick SGV480 carbon steel which is equivalent to ASTM A516 Gr. 70 without post weld heat treatment for containment vessels, the authors have certified the basic properties of base metal and welded joints of 50 mm thick SGV480 steel plates. The results showed that fracture thoughness of welded joints is high without PWHT and the steel is safe enough without PWHT against embrittlement fracture under the operating conditions. (orig.).

  17. Study on the application of 50 mm thick welded joints without PWHT for containment vessels

    International Nuclear Information System (INIS)

    Watanabe, Nozomu; Sakai, Yoshiyuki; Hayashi, Kazutoshi; Higashikubo, Tomohiro; Iida, Kunihiro; Satou, Masanobu

    1992-01-01

    In order to investigate the propriety of the use of 50 mm thick SGV480 carbon steel which is equivalent to ASTM A516 Gr. 70 without post weld heat treatment for containment vessels, the authors have certified the basic properties of base metal and welded joints of 50 mm thick SGV480 steel plates. The results showed that fracture thoughness of welded joints is high without PWHT and the steel is safe enough without PWHT against embrittlement fracture under the operating conditions. (orig.)

  18. Hull Girder Fatigue Damage Estimations of a Large Container Vessel by Spectral Analysis

    DEFF Research Database (Denmark)

    Andersen, Ingrid Marie Vincent; Jensen, Jørgen Juncher

    2013-01-01

    This paper deals with fatigue damage estimation from the analysis of full-scale stress measurements in the hull of a large container vessel (9,400 TEU) covering several months of operation. For onboard decision support and hull monitoring sys-tems, there is a need for a fast reliable method...... for esti-mation of fatigue damage in the ship hull. The objective of the study is to investigate whether the higher frequency contributions from the hydroelastic responses (springing and whipping) can satisfactory be included in the fatigue damage estimation by only a few parameters derived from the stress...

  19. Evaluation of Ultimate Pressure Capacity of a Prestressed Concrete Containment Building with Steel or Polyamide Fiber Reinforcement

    Energy Technology Data Exchange (ETDEWEB)

    Choun, Youngsun; Hahm, Daegi [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Fiber reinforced concrete (FRC) includes thousands of small fibers that are distributed randomly in the concrete. Fibers resist the growth of cracks in concrete through their bridging at the cracks. Therefore, FRC fails in tension only when the fibers break or are pulled out of the cement matrix. For this reason, the addition of fibers in concrete mixing increases the tensile toughness of concrete and enhances the post-cracking behavior. A prevention of through-wall cracks and an increase of the post-cracking ductility will improve the ultimate internal pressure capacity of a prestressed concrete containment building (PCCB). In this study, the effects of steel or polyamide fiber reinforcement on the ultimate pressure capacity of a PCCB are evaluated. When R-SFRC contains hooked steel fibers in a volume fraction of 1.0%, the ultimate pressure capacity of a PCCB can be improved by 17%. When R-PFRC contains polyamide fibers in a volume fraction of 1.5%, the ultimate pressure capacity of a PCCB can be enhanced by 10%. Further studies are needed to determine the strain limits acceptable for PCCBs reinforced with fibers.

  20. Evaluation of Ultimate Pressure Capacity of a Prestressed Concrete Containment Building with Steel or Polyamide Fiber Reinforcement

    International Nuclear Information System (INIS)

    Choun, Youngsun; Hahm, Daegi

    2014-01-01

    Fiber reinforced concrete (FRC) includes thousands of small fibers that are distributed randomly in the concrete. Fibers resist the growth of cracks in concrete through their bridging at the cracks. Therefore, FRC fails in tension only when the fibers break or are pulled out of the cement matrix. For this reason, the addition of fibers in concrete mixing increases the tensile toughness of concrete and enhances the post-cracking behavior. A prevention of through-wall cracks and an increase of the post-cracking ductility will improve the ultimate internal pressure capacity of a prestressed concrete containment building (PCCB). In this study, the effects of steel or polyamide fiber reinforcement on the ultimate pressure capacity of a PCCB are evaluated. When R-SFRC contains hooked steel fibers in a volume fraction of 1.0%, the ultimate pressure capacity of a PCCB can be improved by 17%. When R-PFRC contains polyamide fibers in a volume fraction of 1.5%, the ultimate pressure capacity of a PCCB can be enhanced by 10%. Further studies are needed to determine the strain limits acceptable for PCCBs reinforced with fibers

  1. Method for calculating the duration of vacuum drying of a metal-concrete container for spent nuclear fuel

    Science.gov (United States)

    Karyakin, Yu. E.; Nekhozhin, M. A.; Pletnev, A. A.

    2013-07-01

    A method for calculating the quantity of moisture in a metal-concrete container in the process of its charging with spent nuclear fuel is proposed. A computing method and results obtained by it for conservative estimation of the time of vacuum drying of a container charged with spent nuclear fuel by technologies with quantization and without quantization of the lower fuel element cluster are presented. It has been shown that the absence of quantization in loading spent fuel increases several times the time of vacuum drying of the metal-concrete container.

  2. Further development of KAVERN and code development on gas generation from the containment basement during concrete decomposition

    International Nuclear Information System (INIS)

    Schwarzott, W.; Artnik, J.; Hassmann, K.; Kemner, H.; Stuckenberg, X.

    1983-04-01

    The events during the melt/concrete interaction, e.g. the shape of the cavity and the mass and energy of gases released to the containment atmosphere can be analysed by the computer code KAVERN. In case of basaltic conrete sump water contacts the melt surface after 7 hours. Overpressurization of the containment is calculated to occur after appr. 5 days. For different paths out of the reactor cavity to the containment atmosphere STROMI calculates the mass flow of the gases released during melt concrete interaction. Results show max. temperatures up to 1200 0 C which is well above the self ignition temperature of H 2 . (orig.) [de

  3. Dynamic analysis of steel-concrete structure of TVO power plant containment building

    International Nuclear Information System (INIS)

    Hakala, M.; Karjunen, T.

    1996-08-01

    The report presents results from a study concerning the ability of the containment to withstand the loads caused by steams explosions which are possible during a severe accident at TVO plant (BWR). In the first phase, the suitability of the engineering mechanics code (FLAC) for modelling the dynamic response of damaging steel-concrete structures was tested by post-calculating a small scale test. As a result, a new dynamic material model taking account the fracture orientation was developed. In containment calculations both the developed and the best generally accepted material model were used. The loads against the containment were obtained from a simple model for steam explosions, which allowed the impulse of the pressure load to be fixed by tuning a few parameters. The ability of the containment to withstand the pressure pulses was analysed with loads of 5, 1 0, 20, 40, 60, and 80 kPa s impulse. As a results, the area and magnitude of permanent damage together with time histories of displacement and stress at critical points are presented. The estimations on the consequences of the observed structural damages as far as the containment leak tightness and stability are concerned and presented as conclusions. (9 refs.)

  4. Over-pressure test on BARCOM pre-stressed concrete containment

    Energy Technology Data Exchange (ETDEWEB)

    Parmar, R.M.; Singh, Tarvinder; Thangamani, I.; Trivedi, Neha; Singh, Ram Kumar, E-mail: rksingh@barc.gov.in

    2014-04-01

    Bhabha Atomic Research Centre (BARC), Trombay has organized an International Round Robin Analysis program to carry out the ultimate load capacity assessment of BARC Containment (BARCOM) test model. The test model located in BARC facilities Tarapur; is a 1:4 scale representation of 540 MWe Pressurized Heavy Water Reactor (PHWR) pre-stressed concrete inner containment structure of Tarapur Atomic Power Station (TAPS) unit 3 and 4. There are a large number of sensors installed in BARCOM that include vibratory wire strain gauges of embedded and spot-welded type, surface mounted electrical resistance strain gauges, dial gauges, earth pressure cells, tilt meters and high resolution digital camera systems for structural response, crack monitoring and fracture parameter measurement to evaluate the local and global behavior of the containment test model. The model has been tested pneumatically during the low pressure tests (LPTs) followed by proof test (PT) and integrated leakage rate test (ILRT) during commissioning. Further the over pressure test (OPT) has been carried out to establish the failure mode of BARCOM Test-Model. The over-pressure test will be completed shortly to reach the functional failure of the test model. Pre-test evaluation of BARCOM was carried out with the results obtained from the registered international round robin participants in January 2009 followed by the post-test assessment in February 2011. The test results along with the various failure modes related to the structural members – concrete, rebars and tendons identified in terms of prescribed milestones are presented in this paper along with the comparison of the pre-test predictions submitted by the registered participants of the Round Robin Analysis for BARCOM test model.

  5. Calculation of the process of vacuum drying of a metal-concrete container with spent nuclear fuel

    Science.gov (United States)

    Karyakin, Yu. E.; Lavrent'ev, S. A.; Pavlyukevich, N. V.; Pletnev, A. A.; Fedorovich, E. D.

    2012-01-01

    An algorithm and results of calculation of the process of vacuum drying of a metal-concrete container intended for long-term "dry" storage of spent nuclear fuel are presented. A calculated substantiation of the initial amount of moisture in the container is given.

  6. Preliminary Investigation of Acoustical Properties of Concrete Containing Oil Palm Shell as an Aggregate Replacement

    Science.gov (United States)

    Zanariah, J.; Zaiton, H.; Musli Nizam, Y.; Khairulzan, Y.; Dianah, M.; Nadirah, D.; Hanifi, O. Mohd

    2018-03-01

    Research has been so far focused extensively on mechanical properties of oil palm shell (OPS) concrete but less on sound properties. Thus, the objective of this study is to investigate whether concrete containing OPS can be applied in the field of road noise barrier. The acoustic properties of the samples were determined by using an impedance tube connected to a sound source. The noise reduction coefficient (NRC) and weighted sound absorption coefficient (αw) which is more commonly use in the road traffic noise barrier field were calculated according to BS EN ISO 11654:1997. Compressive strengths of samples were also determined by using compressive test. The results presented that the compressive strength of the OPS composites decreased as increased in w/c wit minimum of 20.44 N/mm2 at 28 days for w/c = 0.6 but still satisfactory for structural use. The sound absorption coefficient demonstrated that they were decreased as the w/c are higher with typical curve of two peaks at 315Hz and 1000Hz. All samples were then can be classified as class E as 0.5< αw < 0.25 and should be classified as L due to favourable deviation higher than 0.25 for 250 Hz.

  7. Comparison between continuous and localized methods to evaluate the flow rate through containment concrete structures

    Energy Technology Data Exchange (ETDEWEB)

    Jason, L., E-mail: ludovic.jason@cea.fr [Atomic Energy Commission (CEA), DEN, DANS, DM2S, SEMT, Mechanics and System Simulation Laboratory (LM2S), F-91191 Gif sur Yvette (France); LaMSID, UMR CNRS-EDF-CEA 8193, F-92141 Clamart (France); Masson, B. [Electricité de France (EDF), SEPTEN, F-69628 Villeurbanne (France)

    2014-10-01

    Highlights: • The contribution focuses on the gas transfer through reinforced concrete structures. • A continuous approach with a damage–permeability law is investigated. • It is significant, for this case, only when the damage variable crosses the section. • In this case, two localized approaches are compared. • It helps at evaluating a “reference” crack opening for engineering laws. - Abstract: In this contribution, different techniques are compared to evaluate the gas flow rate through a representative section of a reinforced and prestressed concrete containment structure. A continuous approach is first applied which is based on the evaluation of the gas permeability as a function of the damage variable. The calculations show that the flow rate becomes significant only when the damage variable crosses the section. But in this situation, the continuous approach is no longer fully valid. That is why localized approaches, based on a fine description of the crack openings, are then investigated. A comparison between classical simplified laws (Poiseuille flow) and a more refined model which takes into account the evolution of the crack opening in the depth of the section enables to define the validity domain of the simplified laws and especially the definition of the associated “reference opening”.

  8. Radioactive material-containing vessel and method of manufacturing the same

    International Nuclear Information System (INIS)

    Kanazawa, Hiroshi; Wada, Katsuyoshi; Ota, Shigeo; Nishioka, Eiji; Okuno, Michinori.

    1995-01-01

    In a vessel for containing radioactive materials having an outer wall with a structure of interposing a lead layer, as a shielding material between inner and outer cylinders made of steel plates, the inner cylinder and the lead layer are in close contact by way of a thin layer of a lead/tin type soldering material and to such an extent that the boundary layer is not detected by supersonic inspection. In addition, flux is coated to the steel plate, which forms the inner cylinder, on the surface being in contact with the lead layer, then a thin layer of the soldering material such as lead or tin is formed, to cast the lead between the inner and the outer cylinders. Then, since the inner cylinder and the lead layer are thermally joined tightly, heat generated at the inside can effectively be released to the outside, so that it is effective as a high-performance cask for transporting a large amount of radioactive materials such as spent nuclear fuels having high temperature afterheat. In addition, a containing vessel with good contact between the inner cylinder and the lead can be manufactured at a low cost only applying a simple primer treatment on the surface of the inner cylinder in addition to an existent lead casting method. (N.H.)

  9. Round Robin Posttest analysis of a 1/10-scale Steel Containment Vessel Model Test

    International Nuclear Information System (INIS)

    Komine, Kuniaki; Konno, Mutsuo

    1999-01-01

    NUPEC and U.S. Nuclear Regulatory Commission (USNRC) have been jointly sponsoring 'Structural Behavior Test' at Sandia National Laboratory (SNL) in Cooperative Containment Research Program'. As one of the test, a test of a mixed scaled SCV model with 1/10 in the geometry and 1/4 in the shell thickness. Round Robin analyses of a 1/10-scale Steel Containment Vessel (SCV) Model Test were carried out to obtain an adequate analytical method among seven organizations belonged to five countries in the world. As one of sponsor, Nuclear Power Engineering Corporation (NUPEC) filled the important role of a posttest analysis of SCV model. This paper describes NUPEC's analytical results in the round robin posttest analysis. (author)

  10. Round Robin Posttest analysis of a 1/10-scale Steel Containment Vessel Model Test

    Energy Technology Data Exchange (ETDEWEB)

    Komine, Kuniaki [Nuclear Power Engineering Corp., Tokyo (Japan); Konno, Mutsuo

    1999-07-01

    NUPEC and U.S. Nuclear Regulatory Commission (USNRC) have been jointly sponsoring 'Structural Behavior Test' at Sandia National Laboratory (SNL) in Cooperative Containment Research Program'. As one of the test, a test of a mixed scaled SCV model with 1/10 in the geometry and 1/4 in the shell thickness. Round Robin analyses of a 1/10-scale Steel Containment Vessel (SCV) Model Test were carried out to obtain an adequate analytical method among seven organizations belonged to five countries in the world. As one of sponsor, Nuclear Power Engineering Corporation (NUPEC) filled the important role of a posttest analysis of SCV model. This paper describes NUPEC's analytical results in the round robin posttest analysis. (author)

  11. Concrete benchmark experiment: ex-vessel LWR surveillance dosimetry; Experience ``Benchmark beton`` pour la dosimetrie hors cuve dans les reacteurs a eau legere

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, H.; D`Hondt, P.; Oeyen, J.; Risch, P.; Bioux, P.

    1993-09-01

    The analysis of DOEL-1 in-vessel and ex-vessel neutron dosimetry, using the DOT 3.5 Sn code coupled with the VITAMIN-C cross-section library, showed the same C/E values for different detectors at the surveillance capsule and the ex-vessel cavity positions. These results seem to be in contradiction with those obtained in several Benchmark experiments (PCA, PSF, VENUS...) when using the same computational tools. Indeed a strong decreasing radial trend of the C/E was observed, partly explained by the overestimation of the iron inelastic scattering. The flat trend seen in DOEL-1 could be explained by compensating errors in the calculation such as the backscattering due to the concrete walls outside the cavity. The `Concrete Benchmark` experiment has been designed to judge the ability of this calculation methods to treat the backscattering. This paper describes the `Concrete Benchmark` experiment, the measured and computed neutron dosimetry results and their comparison. This preliminary analysis seems to indicate an overestimation of the backscattering effect in the calculations. (authors). 5 figs., 1 tab., 7 refs.

  12. Significance of fluid-structure interaction phenomena for containment response to ex-vessel steam explosions

    Energy Technology Data Exchange (ETDEWEB)

    Almstroem, H.; Sundel, T. [National Defence Research Establishment, Stockholm (Sweden); Frid, W.; Engelbrektson, A.

    1998-01-01

    When studying the structural response of a containment building to ex-vessel steam explosion loads, a two-step procedure is often used. In the first step of this procedure the structures are treated as rigid and the pressure-time history generated by the explosion at the rigid wall is calculated. In the second step the calculated pressure is applied to the structures. The obvious weakness of the two-step procedure is that it does not correspond to the real dynamic behaviour of the fluid-structure system. The purpose of this paper is to identify and evaluate the relevant fluid-structure interaction phenomena. This is achieved through direct treatment of the explosion process and the structural response. The predictions of a direct and two-step treatment are compared for a BWR Mark II containment design, consisting of two concentric walls interacting with water masses in the central and annular pools. It is shown that the two-step approach leads to unrealistic energy transfer in the containment system studied, and to significant overestimation of the deflection of the containment wall. As regards the pedestal wall, the direct method analysis shows that the flexibility of this wall affects the pressure-time history considerably. Three load types have been identified for this wall namely shock load, water blow as a result of water cavitation, and hydrodynamic load. Reloading impulse due to cavitation phenomena plays an important role as it amounts to about 40% of the total impulse load. Investigation of the generality of the cavitation phenomena in the context of ex-vessel steam explosion loads was outside the scope of this work. (author)

  13. Significance of fluid-structure interaction phenomena for containment response to ex-vessel steam explosions

    Energy Technology Data Exchange (ETDEWEB)

    Almstroem, H.; Sundel, T. (Nat. Defence Res. Establ., Tumba (Sweden)); Frid, W. (Swedish Nuclear Power Inspectorate, SE-10658, Stockholm (Sweden)); Engelbrektson, A. (VBB/SWECO, Box 34044, SE-10026, Stockholm (Sweden))

    1999-05-01

    When studying the structural response of a containment building to ex-vessel steam explosion loads, a two-step procedure is often used. In the first step of this procedure the structures are treated as rigid and the pressure-time history generated by the explosion, at the rigid wall, is calculated. In the second step the calculated pressure is applied to the structures. The obvious weakness of the two-step procedure is that it does not correspond to the real dynamic behaviour of the fluid-structure system. The purpose of this paper is to identify and evaluate the relevant fluid-structure interaction phenomena. This is achieved through direct treatment of the explosion process and the structural response. The predictions of a direct and two-step treatment are compared for a BWR Mark II containment design, consisting of two concentric walls interacting with water masses in the central and annular pools. It is shown that the two-step approach leads to unrealistic energy transfer in the containment system studied and to significant overestimation of the deflection of the containment wall. As regards the pedestal wall, the direct method analysis shows that the flexibility of this wall affects the pressure-time history considerably. Three load types have been identified for this wall namely shock load, water blow as a result of water cavitation, and hydrodynamic load. Reloading impulse due to cavitation phenomena plays an important role as it amounts to [approx]40% of the total impulse load. Investigation of the generality of the cavitation phenomena in the context of ex-vessel steam explosion loads was outside the scope of this work. (orig.) 5 refs.

  14. Significance of fluid-structure interaction phenomena for containment response to ex-vessel steam explosions

    International Nuclear Information System (INIS)

    Almstroem, H.; Sundel, T.; Frid, W.; Engelbrektson, A.

    1999-01-01

    When studying the structural response of a containment building to ex-vessel steam explosion loads, a two-step procedure is often used. In the first step of this procedure the structures are treated as rigid and the pressure-time history generated by the explosion, at the rigid wall, is calculated. In the second step the calculated pressure is applied to the structures. The obvious weakness of the two-step procedure is that it does not correspond to the real dynamic behaviour of the fluid-structure system. The purpose of this paper is to identify and evaluate the relevant fluid-structure interaction phenomena. This is achieved through direct treatment of the explosion process and the structural response. The predictions of a direct and two-step treatment are compared for a BWR Mark II containment design, consisting of two concentric walls interacting with water masses in the central and annular pools. It is shown that the two-step approach leads to unrealistic energy transfer in the containment system studied and to significant overestimation of the deflection of the containment wall. As regards the pedestal wall, the direct method analysis shows that the flexibility of this wall affects the pressure-time history considerably. Three load types have been identified for this wall namely shock load, water blow as a result of water cavitation, and hydrodynamic load. Reloading impulse due to cavitation phenomena plays an important role as it amounts to ∼40% of the total impulse load. Investigation of the generality of the cavitation phenomena in the context of ex-vessel steam explosion loads was outside the scope of this work. (orig.)

  15. EDF reactor building containment: Monitoring of the pre-stressed concrete structure

    International Nuclear Information System (INIS)

    Badez, N.

    2009-01-01

    The concrete containments of the EDF PWR are pre-stressed, and are monitored to observe the ageing effects on the structure, in particular the evolutions of creep, shrinkage, pre-stress loss, and air leakage tightness. Monitoring devices are installed during construction period, and measurements are checked, stored on a data base, and analysed during all the plant operating life time. The topic of the presentation is to present each part of the EDF monitoring organisation. A continuous monitoring makes it possible to produce periodical comprehensive reports about the mechanical analysis of the structure, the strain stabilisation,... Periodical tests (each 10 years) are planned. They consist to submit the containment to an internal air pressure at the accidental pressure level. The monitoring system gives the strain values in order to check their linearity and reversibility with decreasing pressure. At the same time, the containment tightness is checked with a specific instrumentation to verify that leak rate is lower than the required level. A general view of instrumentation implemented on the containment (sensors, data acquisition), and a data analysis are presented

  16. Durability Properties of Self Compacting Concrete containing Fly ash, Lime powder and Metakaolin

    Directory of Open Access Journals (Sweden)

    Rizwan Ahmad Khan

    2016-01-01

    Full Text Available This paper investigates the durability properties of Self-compacting concrete (SCC, with different amounts of fly ash (FA, lime powder (LP and metakaolin (MK. A total of 6 mixes were prepared that have a constant water-binder ratio (w/b of 0.41 and superplasticizer dosage of 1% by weight of cement. In addition to compressive strength, the durability properties of SCC mixes were determined by means of Initial surface absorption test (ISAT and Capillary suction test. The test results indicated that the durability properties of the mixes appeared to be very dependent on the type and amount of the mineral admixture used; the mixes containing MK were found to have considerably higher permeability resistance. Good co-relation between strength and absorption were achieved.

  17. Discrimination of high-Z materials in concrete-filled containers using muon scattering tomography

    Science.gov (United States)

    Frazão, L.; Velthuis, J.; Thomay, C.; Steer, C.

    2016-07-01

    An analysis method of identifying materials using muon scattering tomography is presented, which uses previous knowledge of the position of high-Z objects inside a container and distinguishes them from similar materials. In particular, simulations were performed in order to distinguish a block of Uranium from blocks of Lead and Tungsten of the same size, inside a concrete-filled drum. The results show that, knowing the shape and position from previous analysis, it is possible to distinguish 5 × 5 × 5 cm3 blocks of these materials with about 4h of muon exposure, down to 2 × 2 × 2 cm3 blocks with 70h of data using multivariate analysis (MVA). MVA uses several variables, but it does not benefit the discrimination over a simpler method using only the scatter angles. This indicates that the majority of discrimination is provided by the angular information. Momentum information is shown to provide no benefits in material discrimination.

  18. Concrete works for Hamaoka No. 1 nuclear power plant

    International Nuclear Information System (INIS)

    Horiuchi, Minoru; Sugihara, Kazuo; Iwasawa, Jiro.

    1975-01-01

    Various aspects of concrete works performed for the reactor building of Hamaoka No.1 plant are reviewed. Control building and waste disposal building were all together combined with the reactor building in order to improve safety against earthquakes. Special consideration was given for the quality control of concrete works by establishing quality control committee, making quality control manual and by performing daily examination and monthly report. The quality and various materials of concrete used are described. The composition of concrete used for various parts of the building is also listed. Detailed description is made regarding the concrete placing for foundation mat, under a containment vessel, and the construction of air gaps and the placing of shielding concrete around the containment vessel. Curves representing the temperature history of concrete at various points are presented. As for testing, the items of test, methods of measurement, and the results of these test and measurement are presented in detail. (Aoki, K.)

  19. Experimental Investigation of Thermal Conductivity of Concrete Containing Micro-Encapsulated Phase Change Materials

    DEFF Research Database (Denmark)

    Pomianowski, Michal Zbigniew; Heiselberg, Per; Jensen, Rasmus Lund

    2011-01-01

    in this article utilizes integration of the concrete and the microencapsulated Phase Change Material (PCM). PCM has the ability to absorb and release significant amounts of heat at a specific temperature range. As a consequence of admixing PCM to the concrete, new thermal properties like thermal conductivity...... and specific heat capacity have to be defined. This paper presents results from the measurements of the thermal conductivity of various microencapsulated PCM-concrete and PCM-cement-paste mixes. It was discovered that increase of the amount of PCM decreases the thermal conductivity of the concrete PCM mixture....... Finally, a theoretical calculation methodology of thermal conductivity for PCM-concrete mixes is developed....

  20. Two, three and four buttressed PWR containment vessels. A comparative study

    International Nuclear Information System (INIS)

    Dufour, C.J.; Cheyrezy, M.H.; Thorsen, N.E.

    1977-01-01

    The purpose of the paper is to analyse the advantages and drawbacks of different arrangements of hood tendons and buttresses for a PWR containment vessel. It is shown that the solution with two buttresses and full hoop tendons (through 360 0 ) gives: acceptable secondary stresses and strains; a lower total cost of the prestressing; a shorter time schedule for prestressing operations; a smaller 'forbidden area' for the penetration lay out. This arrangement has been used for the first time (to our knowledge) in PWR containment design for the nuclear plant of DOEL III, Belgium. The comparative study has been carried out for a containment vessel with a 85 cm thick wall and an interior diameter of 42.50 m. The Guaranteed Ultimate Tensile Strength of the tendons is 9000 KN. The minimum required prestressing force is 10 300 KN per linear meter of height. The buttresses are 4.0 m long and 1.70 m thick on their vertical centerline. The following arrangements have been studied: 4 buttresses and 3/4 hoop tendons; 3 buttresses and 2/3 hoop tendons; 3 buttresses and full hoop tendons; 2 buttresses and full hoop tendons. The ovalisation and other secondary effects result mainly from three different causes: the average prestressing force is not constant around a hoop; the buttress itself constitutes a sudden thickening and creates local disturbances even under an axisymmetric loading: the anchored tendons are straight over their first 6 meters and they initiate local stress perturbations in the vicinity of the junction of the buttress and the wall. The determination of these secondary effects has been performed by a plane stress finite element analysis