WorldWideScience

Sample records for primary nuclear containment

  1. 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)

  2. Bellefonte primary containment structure

    International Nuclear Information System (INIS)

    Olyniec, J.H.

    1981-01-01

    Construction of the reactor building primary containment structure at the Bellefonte Nuclear Plant involved several specialized construction techniques. This two unit plant is one of the nine nuclear units at six different sites now under construction by the Tennessee Valley Authority (TVA). The post-Tensioned, cast-in-place interior steel lined containment structure is unique within TVA. Problems during construction were identified at weekly planning meetings, and options were discussed. Close coordination between craft supervisors and on-site engineering personnel drew together ''hands-on''experience and technical background. Details of the construction techniques, problems, and solutions are presented

  3. 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.)

  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. Leak rates and structural integrity tests for Laguna Verde Nuclear Power Plant primary containment. Regulatory experience

    International Nuclear Information System (INIS)

    Mamani Alegria, Yuri Raul; Salgado Gonzalez, Julio Ricardo

    1996-01-01

    In the Appendix A General Design Criteria for Nuclear Power Plants of the US Code of Federal Regulations title 10 part 50 (10CFR50) is established the Criterion 1 Quality standards and records which requires that structures, systems and components important to safety should be tested to quality standards according with the importance of the safety function to be performed. This regulation has been adopted by the Mexican Regulatory Body (CNSNS) for their nuclear power plants. (author)

  6. Development of an automated remote inspection system for the interior of the primary containment vessel of a nuclear power plant

    International Nuclear Information System (INIS)

    Senoo, Makoto; Yoshida, Tomiharu; Omote, Tatsuyuki; Tanaka, Keiji; Koga, Kazunori

    1996-01-01

    An automated remote inspection system has been developed for the interior of the primary containment vessel of a nuclear power plant. This system consists of an inspection robot and an operator's console. The inspection robot travels along a monorail provided in the interior of the primary containment vessel. The operator's console is located in the central control room of the power plant. We have made efforts to downsize the robot and automate the inspection and monitoring machinery. As for downsizing the robot, a 152 mm wide, 290 mm high cross-sectional area and 15 kg weight can be realized using commercially available small sensors and rearranging the parts in those sensors. As for automating the inspection and monitoring, several monitoring functions are developed using image processing, frequency analysis and other techniques applied to signals from sensors such as an ITV camera, an infrared camera and a microphone, which are mounted on the robot. Endurance tests show resistance of the robot to radiational and thermal conditions is adequate for actual use in actual power plants. (author)

  7. Nuclear reactor container

    International Nuclear Information System (INIS)

    Ishiyama, Takenori.

    1989-01-01

    This invention concerns a nuclear reactor container in which heat is removed from a container by external water injection. Heat is removed from the container by immersing the lower portion of the container into water and scattering spary water from above. Thus, the container can be cooled by the spray water falling down along the outer wall of the container to condensate and cool vapors filled in the container upon occurrence of accidents. Further, since the inside of the container can be cooled also during usual operation, it can also serve as a dry well cooler. Accordingly, heat is removed from the reactor container upon occurrence of accidents by the automatic operation of a spray device corresponding to the change of the internal temperature and the pressure in the reactor container. Further, since all of these devices are disposed out of container, maintenance is also facilitated. (I.S.)

  8. Component nuclear containment structure

    International Nuclear Information System (INIS)

    Harstead, G.A.

    1979-01-01

    The invention described is intended for use primarily as a nuclear containment structure. Such structures are required to surround the nuclear steam supply system and to contain the effects of breaks in the nuclear steam supply system, or i.e. loss of coolant accidents. Nuclear containment structures are required to withstand internal pressure and temperatures which result from loss of coolant accidents, and to provide for radiation shielding during operation and during the loss of coolant accident, as well as to resist all other applied loads, such as earthquakes. The nuclear containment structure described herein is a composite nuclear containment structure, and is one which structurally combines two previous systems; namely, a steel vessel, and a lined concrete structure. The steel vessel provides strength to resist internal pressure and accommodate temperature increases, the lined concrete structure provides resistance to internal pressure by having a liner which will prevent leakage, and which is in contact with the concrete structure which provides the strength to resist the pressure

  9. Nuclear reactor containment device

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu.

    1980-01-01

    Purpose: To reduce the volume of a containment shell and decrease the size of a containment equipment for BWR type reactors by connecting the containment shell and a suppression pool with slanted vent tubes to thereby shorten the vent tubes. Constitution: A pressure vessel containing a reactor core is installed at the center of a building and a containment vessel for the nuclear reactor that contains the pressure vessel forms a cabin. To a building situated below the containment shell, is provided a suppression chamber in which cooling water is charged to form a suppression pool. The suppression pool is communicated with vent tubes that pass through the partition wall of the containment vessel. The vent tubes are slanted and their lower openings are immersed in coolants. Therefore, if accident is resulted and fluid at high temperature and high pressure is jetted from the pressure vessel, the jetting fluid is injected and condensated in the cooling water. (Moriyama, K.)

  10. Nuclear reactor container

    International Nuclear Information System (INIS)

    Yamaki, Rika; Kawabe, Ryuhei.

    1989-01-01

    A venturi scrubber is connected to a nuclear reactor container. Gases containing radioactive aerosols in the container are introduced into the venturi scrubber in the form of a high speed stream under the pressure of the container. The radioactive aerosols are captured by inertia collision due to the velocity difference between the high speed gas stream and water droplets. In the case of the present invention, since the high pressure of the reactor container generated upon accident is utilized, compressor, etc. is no more required, thereby enabling to reduce the size of the aerosol removing device. Further, since no external power is used, the radioactive aerosols can be removed with no starting failure upon accidents. (T.M.)

  11. Nuclear reactor container

    International Nuclear Information System (INIS)

    Hosaka, Seiichi.

    1988-01-01

    Cables coverd with non-halogen covering material are used as electric wire cables wired for supplying electric power to a reactor recycling pump. Silicone rubber having specified molecular formula is used for the non-halogen covering material. As a result, formation of chlorine in a nuclear reactor container can be eliminated and increase in the deposited salts to SUS pipeways, etc. can be prevented, to avoid the occurrence of stress corrosion cracks. (H.T.)

  12. Nuclear reactor container

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1989-01-01

    Aerosol filters considered so far for nuclear reactor containers in conventional BWR type nuclear power plants make the facility larger and involve a risk of clogging. In view of the above, in the present invention, the diameter of a flow channel of gases entering from a bent pipe to a suppression pool is made smaller thereby decreasing the diameter of gas bubbles in the supperssional pool. Since this reduces the force of surface tension, the diameter of resulted gas bubbles is made remarkably smaller as compared with the case where the gases are released from the lower end of the bent pipe. Since the absorption velocity of bubble-entrained aerosols into water is in proportion to the square of the bubble diameter, the absorption efficiency can be increased remarkably by reducing the diameter of the gas bubbles. Accordingly, it is possible to improve the efficiency of eliminating radioactivity of released gases. (K.M.)

  13. Analysis of a Multi-Venturi filter for the venting of the primary container of a nuclear reactor

    International Nuclear Information System (INIS)

    Reyes G, A. A.; Sainz M, E.; Ortiz V, J.

    2017-09-01

    Since the Chernobyl nuclear accident, European nuclear power plants have opted to install filters in the containment vent pipes, whose function is to help mitigate the consequences of a severe accident, by means of the controlled depressurization of the containment passively through of a containment filtering vent system. These systems are designed to relieve the internal pressure of containment by deliberately opening pressure relief devices, either a valve or rupture disk during a severe accident and being channeled to the filtering unit. In this work, the hydraulic response of a liquid gas washing filtration system is evaluated, since this information is necessary to estimate the effect of the increase of the containment pressure on the venting capacity of the vent pipes. Through CFD simulation, using the programs with open source license CaeLinux-2014 and OpenFoam, the hydrodynamic characteristics of the Multi-Venturi system were obtained for the washing of the gases coming from the containment, which could be included in the general model of the vent pipe. Representative models of the venturi tubes of each concentric sector that are part of the washing system were generated and by parametric calculations the average mass expense established by each venturi was estimated, according to its dimensions and depth to which is located inside the tank. In the same way, the pressure and mass expense required to activate each concentric sector was calculated according to the pressure and mass load from the containment, in order to estimate the maximum expenditure that is established through the filter. The velocity profiles and the characteristic pressure at which each sector operates were also calculated, as well as the local and global discharge pressure drop. (Author)

  14. Nuclear reactor container

    International Nuclear Information System (INIS)

    Moriyama, Takeo; Ochiai, Kanehiro; Niino, Tsuyoshi; Kodama, Toyokazu; Hirako, Shizuka.

    1988-01-01

    Purpose: To obtain structures suitable to a container structures for nuclear power plants used in those districts where earthquakes occur frequently, in which no local stresses are caused to the fundamental base portions and the workability for the fundamental structures is improved. Constitution: Basic stabilizers are attached to a nuclear reactor container (PCV) and a basic concrete recess for receiving a basic stabilizer is disposed in basic concretes. A top stabilizer is joined and fixed to a top stabilizer receiving plate at the inside of an outer shielding wall. On the other hand, a PCV top recess for conducting the load of PCV to the top stabilizer is attached to the top of the PCV. By disposing stabilizer structures allowing miner displacements at the two points, that is, the top and the lowermost portion of the PCV, no local stress concentrations can be generated to the extension on the axial direction of components due to the inner pressure of the PCV and to the horizontal load applied to the upper portion of the PCV upon earthquakes. (Yoshino, Y.)

  15. Nuclear reactor container

    International Nuclear Information System (INIS)

    Fukui, Tooru; Murase, Michio; Kataoka, Yoshiyuki; Hidaka, Masataka; Sumita, Isao; Tominaga, Kenji.

    1992-01-01

    In a nuclear reactor container, a chamber in communication with a wet well of a pressure suppression chamber is disposed and situated to such a position that the temperature is lower than a chamber containing pool water upon occurrence of loss of coolant accident. In addition, the inner surface of the pressure suppression chamber is constituted with steel walls in contact with pool water, and an outer circumferential pool is disposed at the outer circumferential surface thereof. Further, a circulation channel is disposed, and a water intake port is disposed at a position higher than an exit to the pool water, and a water discharge port is opened in the pool water at a position lower than the exit to the pool water. With such a constitution, the allowable temperature of the pressure suppression pool water can be elevated to a saturated steam temperature corresponding to the resistant pressure of the container, so that the temperature difference between the pressure suppression pool and the outer side thereof is increased by so much, to improve thermal radiation performance. Accordingly, it can be utilized as a pressure suppression means for a plant of greater power. Further, thermal conduction efficiency from the pool water region of the pressure suppression chamber to the outer circumferential pool water is improved, or thermal radiation area is enlarged due to the circulation channel, to improve the heat radiation performance. (N.H.)

  16. Nuclear reactor container

    International Nuclear Information System (INIS)

    Shioiri, Akio.

    1992-01-01

    In a nuclear reactor container, a vent tube communication port is disposed to a pressure suppression pool at a position higher than the pool water therein for communication with an upper dry well, and the upper end opening of a dry well communication pipe is disposed at a position higher than the communication port. When condensate return pipeline is ruptured in the upper dry well, water in a water source pool is injected to the pressure vessel and partially discharged out of the ruptured port and a depressurization valve connected to the pressure vessel to the inside of the upper dry well. The discharged water stays in the upper dry well and, when the water level reaches the height of the vent tube communication port, it flows into the pressure suppression pool. Even in a state that the entire amount of water in the water source pool is supplied, since water does not reach the upper opening port of the dry well communication pipe, water does not flow into a lower dry well. Accordingly, the motor of a control rod drives disposed in the lower dry well can be prevented from submerging. The reactor core can be cooled more reliably, to improve the reliability of the pressure suppression function. (N.H.)

  17. Composite containment for nuclear power

    International Nuclear Information System (INIS)

    Harstead, G.A.; Soeoet, O.

    1977-01-01

    Fundamentally, a nuclear reactor containment structure provides three major functions; namely, (1), to withstand loads due to pressure and temperature increase due to Design Basis Accident (DBA) (2), to withstand environmental loads such as seismic, tornado and normal loads, and (3) act as a radiation shield. Conventional design practise is to employ either a steel vessel and concrete shield building or a steel lined concrete structure. This paper deals with a new concept in which a steel liner is employed which carries much of the primary membrane loads. This type of structure is similar in some aspects to the previously described systems: a) A mat, lined with a thin plate on its top surface, is similar to concrete containment. b) A cylinder and hemispherical dome, made up of steel plate and concrete, is about 2.5 feet thick (the minimum required for radiation shielding). Although the steel plate and concrete are in contact, as in concrete containment, the steel plate in composite containment is much thicker than the liner. There are two main advantages over present practise; namely reduction of materials and therefore reduced capital cost and even more significantly a shortened construction schedule which will permit more flexibility in overall plant construction schedule and will benefit the cash flow situation. (Auth.)

  18. Nuclear power. [Contains glossary

    Energy Technology Data Exchange (ETDEWEB)

    Patterson, W.C.

    1983-01-01

    Lay language brings an understanding of nuclear technology and nuclear politics to the non-specialist reader. The author notes that there has been little change in the technology during the four decades of the nuclear age, but mankind has still to learn how to live with it. Part One explains how reactors work, identifies different reactor types, and describes the fuel cycle. Part two follows research developments during the pre-Manhatten Project days, the war effort, and the decision to pursue commercial nuclear power. He traces the development of policies to secure fission materials and international efforts to prevent the proliferation of weapons grade material and the safe handling of radioactive wastes on a global as well as national scale. There are four appendices, including an annotated reference to other publications. 9 figures.

  19. Passive containment system for a nuclear reactor

    International Nuclear Information System (INIS)

    Kleimola, F.W.

    1976-01-01

    A containment system is described that provides complete protection entirely by passive means for the loss of coolant accident in a nuclear power plant and wherein all stored energy released in the coolant blowdown is contained and absorbed while the nuclear fuel is continuously maintained submerged in liquid. The primary containment vessel is restored to a high subatmospheric pressure within a few minutes after accident initiation and the decay heat is safely transferred to the environment while radiolytic hydrogen is contained by passive means

  20. Nuclear reactor containing facility

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Murase, Michio.

    1994-01-01

    In a reactor containing facility, a condensation means is disposed above the water level of a cooling water pool to condensate steams of the cooling water pool, and return the condensated water to the cooling water pool. Upon occurrence of a pipeline rupture accident, steams generated by after-heat of a reactor core are caused to flow into a bent tube, blown from the exit of the bent tube into a suppression pool and condensated in a suppression pool water, thereby suppressing the pressure in the reactor container. Cooling water in the cooling water pool is boiled by heat conduction due to the condensation of steams, then the steams are exhausted to the outside of the reactor container to remove the heat of the reactor container to the outside of the reactor. In addition, since cooling water is supplied to the cooling water pool quasi-permanently by gravity as a natural force, the reactor container can be cooled by the cooling water pool for a long period of time. Since the condensation means is constituted with a closed loop and interrupted from the outside, radioactive materials are never released to the outside. (N.H.)

  1. Nuclear reactor container

    International Nuclear Information System (INIS)

    Fujimoto, Kiyoshi; Kataoka, Yoshiyuki; Murase, Michio; Fujii, Tadashi; Susuki, Akira.

    1994-01-01

    A wet well space above a pressure suppression pool is divided into a first wet well on the side in contact with the pressure suppression pool and a second wet well on the side not in contact with the pool. Cooling water is contained in the second wet well and it is in communication with the first wet well by pipelines. Since steams flown into the second well are condensed in the cooling water, they continuously transfer from the first wet well to the second wet well, thereby capable of eliminating the effects of incondensible gases in the first wet well. With such procedures, the effect of the incondensible gases can be eliminated even without cooling from the outside of the reactor. Heat accumulation can be increased in a container of any material, so that thermal load on cooling circuits for removing after-heat can be mitigated. (T.M.)

  2. Nuclear steam system containment

    International Nuclear Information System (INIS)

    Jabsen, F.

    1980-01-01

    An improved containment used for radiation shielding and pressure suppression comprising a dry well includes a pressure vessel, a plurality of concentric wall means, said plurality of concentric wall means defining at least three annular regions about said dry well. A first annular region provides the containment used for radiation shielding, a second annular region is substantially dry, a third annular region provides a wet well for relieving fluid pressure released from the pressure vessel into the dry well. Pipe connection means extend in the wet well from the dry well, a pool of liquid is disposed to partially fill said third annular region, the upper end portion of the second and third annular regions having an enclosure, and a plurality of baffle plates extending vertically downward from said enclosure in said third annular region into said pool of liquid so as to circumferentially divide the upper portion of said third annular region into a plurality of circumferential upper portions

  3. Shipping container for nuclear fuels

    International Nuclear Information System (INIS)

    Housholder, W.R.; Greer, N.L.

    1976-01-01

    A container for nuclear materials is described wherein a specially and uniquely constructed pressure vessel and gamma shield assembly for holding the nuclear materials is provided in a housing, and wherein a positioning means extends between the housing and the assembly for spacing the same, insulation in the housing essentially filling the space between the assembly and housing, the insulation comprising beads, globules or the like of water encapsulated in plastic and which, in one important embodiment, contains neutron absorbing matter

  4. Container for nuclear fuel powders

    International Nuclear Information System (INIS)

    Etheredge, B.F.; Larson, R.I.

    1982-01-01

    A critically safe container is disclosed for the storage and rapid discharge of enriched nuclear fuel material in powder form is disclosed. The container has a hollow, slab-shaped container body that has one critically safe dimension. A powder inlet is provided on one side wall of the body adjacent to a corner thereof and a powder discharge port is provided at another corner of the body approximately diagonal the powder inlet. Gas plenum for moving the powder during discharge are located along the side walls of the container adjacent the discharge port

  5. Glass containing radioactive nuclear waste

    International Nuclear Information System (INIS)

    Boatner, L.A.; Sales, B.C.

    1985-01-01

    Lead-iron phosphate glasses containing a high level of Fe 2 O 3 for use as a storage medium for high-level-radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90 C, with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10 2 to 10 3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe 2 O 3 in forming the lead-iron phosphate glass is critical. Lead-iron phosphate nuclear waste glass can be prepared at temperatures as low as 800 C, since they exhibit very low melt viscosities in the 800 to 1050 C temperature range. These waste-loaded glasses do not readily devitrify at temperatures as high as 550 C and are not adversely affected by large doses of gamma radiation in H 2 O at 135 C. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear waste forms. (author)

  6. Nuclear fuel element and container

    International Nuclear Information System (INIS)

    Grubb, W.T.; King, L.H.

    1981-01-01

    The invention is based on the discovery that a substantial reduction in metal embrittlement or stress corrosion cracking from fuel pellet-cladding interaction can be achieved by the use of a copper layer or liner in proximity to the nuclear fuel, and an intermediate zirconium oxide barrier layer between the copper layer and the zirconium cladding substrate. The intermediate zirconia layer is a good copper diffusion barrier; also, if the zirconium cladding surface is modified prior to oxidation, copper can be deposited by electroless plating. A nuclear fuel element is described which comprises a central core of fuel material and an elongated container using the system outlined above. The method for making the container is again described. It comprises roughening or etching the surface of the zirconium or zirconium alloy container, oxidizing the resulting container, activating the oxidized surface to allow for the metallic coating of such surfaces by electroless deposition and further coating the activated-oxidized surface of the zirconium or zirconium alloy container with copper, iron or nickel or an alloy thereof. (U.K.)

  7. Double containment shell for nuclear power plants

    International Nuclear Information System (INIS)

    Sykora, D.

    1977-01-01

    A double containment shell is proposed for nuclear power plants, especially those equipped with pressurized water reactors. The shell offers increased environmental protection from primary circuit accidents. The inner shell is built of steel or concrete while the outer shell is always built of concrete. The space between the two shells is filled with water and is provided with several manholes and with stiffeners designed for compensation for load due to the water hydrostatic pressure. Water serves the airtight separation of the containment shell inside from the environment and the absorption of heat released in a primary circuit accident. In case the inner shell is made of concrete, it is provided with heat-removal tubes in-built in its walls ensuring rapid heat transfer from the inside of the containment to the water in the interwall space. (Z.M.)

  8. Nuclear power plant containment construction

    International Nuclear Information System (INIS)

    Schabert, H.P.; Danisch, R.; Strickroth, E.

    1975-01-01

    The Nuclear Power Plant Containment Construction includes the spherical steel safety enclosure for the reactor and the equipment associated with the reactor and requiring this type of enclosure. This steel enclosure is externally structurally protected against accident by a concrete construction providing a foundation for the steel enclosure and having a cylindrical wall and a hemispherical dome, these parts being dimensioned to form an annular space surrounding the spherical steel enclosure, the latter and the concrete construction heretofore being concentrically arranged with respect to each other. In the disclosed construction the two parts are arranged with their vertical axis horizontally offset from each other so that opposite to the offsetting direction of the concrete construction a relatively large space is formed in the now asymmetrical annular space in which reactor auxiliary equipment not requiring enclosure by the steel containment vessel or safety enclosure, may be located outside of the steel containment vessel and inside of the concrete construction where it is structurally protected by the latter

  9. Present status of nuclear containments in Korea

    International Nuclear Information System (INIS)

    Park, Jihong; Hong, Jaekeun; Lee, Byunghoon; Son, Youngho

    2007-01-01

    Since the first nuclear power plant in Korea, Kori unit no.1, was started in commercial service in 1978, 20 units including Kori unit no.1 have been operated and maintained until now in Korea. Recently several units were started to be constructed and also, additionally more than 4 units were planned to be constructed in the near future. The importance of nuclear containments has been always one of the hottest issues for the safety and protection of nuclear power plants until now. At the beginning of nuclear power plants construction in Korea, several typed nuclear containment systems were adopted. For those reasons, various codes, standards, and inspection technologies are applied to nuclear containment systems differently. In this study, the status of inservice inspection performed for the safety and maintenance of nuclear containments in Korea was researched. Overall nuclear containment systems and inspections performed up to recently in Korea including trends, inspection items, periods, and regulations were described briefly. (author)

  10. Containment for low temperature district nuclear-heating reactor

    International Nuclear Information System (INIS)

    He Shuyan; Dong Duo

    1992-03-01

    Integral arrangement is adopted for Low Temperature District Nuclear-heating Reactor. Primary heat exchangers, control rod drives and spent fuel elements are put in the reactor pressure vessel together with reactor core. Primary coolant flows through reactor core and primary heat exchangers in natural circulation. Primary coolant pipes penetrating the wall of reactor pressure vessel are all of small diameters. The reactor vessel constitutes the main part of pressure boundary of primary coolant. Therefore the small sized metallic containment closed to the wall of reactor vessel can be used for the reactor. Design principles and functions of the containment are as same as the containment for PWR. But the adoption of small sized containment brings about some benefits such as short period of manufacturing, relatively low cost, and easy for sealing. Loss of primary coolant accident would not be happened during the rupture accident of primary coolant pressure boundary inside the containment owing to its intrinsic safety

  11. Containments for consolidated nuclear steam systems

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1978-01-01

    A containment system for a consolidated nuclear steam system incorporating a nuclear core, steam generator and reactor coolant pumps within a single pressure vessel is described which is designed to provide radiation shielding and pressure suppression. Design details, including those for the dry well and wet well of the containment, are given. (UK)

  12. Ultimate load model test for Sizewell 'B' primary containment

    International Nuclear Information System (INIS)

    Crowder, R.

    1988-01-01

    This paper considers the factors influencing the adoption of an ultimate load factor for the Sizewell 'B' PWR primary containment structure. As part of the validation process for the ultimate load analysis method, a proposal has been made by Nuclear Design Associates to build and test a 1/10th scale model of the containment structure, which would proceed following the granting of section 2 consent for Sizewell 'B'. The modelling principles, construction method and test proposals are examined in some detail. The proposal is currently being considered by the CEGB's Project Management Team. (author)

  13. Analysis of a Multi-Venturi filter for the venting of the primary container of a nuclear reactor; Analisis de un filtro multiventuri para el venteo del contenedor primario de un reactor nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Reyes G, A. A.; Sainz M, E.; Ortiz V, J., E-mail: alejandroantonioreyess@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2017-09-15

    Since the Chernobyl nuclear accident, European nuclear power plants have opted to install filters in the containment vent pipes, whose function is to help mitigate the consequences of a severe accident, by means of the controlled depressurization of the containment passively through of a containment filtering vent system. These systems are designed to relieve the internal pressure of containment by deliberately opening pressure relief devices, either a valve or rupture disk during a severe accident and being channeled to the filtering unit. In this work, the hydraulic response of a liquid gas washing filtration system is evaluated, since this information is necessary to estimate the effect of the increase of the containment pressure on the venting capacity of the vent pipes. Through CFD simulation, using the programs with open source license CaeLinux-2014 and OpenFoam, the hydrodynamic characteristics of the Multi-Venturi system were obtained for the washing of the gases coming from the containment, which could be included in the general model of the vent pipe. Representative models of the venturi tubes of each concentric sector that are part of the washing system were generated and by parametric calculations the average mass expense established by each venturi was estimated, according to its dimensions and depth to which is located inside the tank. In the same way, the pressure and mass expense required to activate each concentric sector was calculated according to the pressure and mass load from the containment, in order to estimate the maximum expenditure that is established through the filter. The velocity profiles and the characteristic pressure at which each sector operates were also calculated, as well as the local and global discharge pressure drop. (Author)

  14. Aircraft transporting container for nuclear fuel

    International Nuclear Information System (INIS)

    Kurakami, Jun-ichi; Kubo, Minoru.

    1991-01-01

    The present invention concerns an air craft transporting container for nuclear fuels. A sealing container that seals a nuclear fuel container and constitutes a sealed boundary for the transporting container is incorporated in an inner container. Shock absorbers are filled for absorbing impact shock energy in the gap between the inner container and the sealing container. The inner container is incorporated with wooden impact shock absorbers being filled so that it is situated in a substantially central portion of an external container. Partitioning cylinders are disposed coaxially in the cylindrical layer filled with wooden impact shock absorbers at an intermediate portion between the outer and the inner containers. Further, a plurality of longitudinally intersecting partitioning disks are disposed each at a predetermined distance in right and left cylindrical wooden impact shock absorbing layers which are in contact with the end face of the inner container. Accordingly, the impact shock energy can be absorbed by the wooden impact shock absorbers efficiently by a plurality of the partitioning disks and the partitioning cylinders. (I.N.)

  15. Nuclear power plant with a containment

    International Nuclear Information System (INIS)

    Barthelmes, C.P.

    1982-01-01

    In nuclear power plants there is usually a containment incorporating components bearing activity. If in the cladding free hydrogen develops, controlled oxidation must be ensured by means of a recombination device, in order to prevent oxyhydrogen explosions. For this purpose, a permanent thoroughmixing of the gases in the containment is required. This can be achieved by vertical shafts reaching to at least half the height of the containment and provided with heating devices to initiate the gas circulation by the stack effect. These heating devices mainly serve as thermal recombinator. (orig.) [de

  16. Predicting the Lifetimes of Nuclear Waste Containers

    Science.gov (United States)

    King, Fraser

    2014-03-01

    As for many aspects of the disposal of nuclear waste, the greatest challenge we have in the study of container materials is the prediction of the long-term performance over periods of tens to hundreds of thousands of years. Various methods have been used for predicting the lifetime of containers for the disposal of high-level waste or spent fuel in deep geological repositories. Both mechanical and corrosion-related failure mechanisms need to be considered, although until recently the interactions of mechanical and corrosion degradation modes have not been considered in detail. Failure from mechanical degradation modes has tended to be treated through suitable container design. In comparison, the inevitable loss of container integrity due to corrosion has been treated by developing specific corrosion models. The most important aspect, however, is to be able to justify the long-term predictions by demonstrating a mechanistic understanding of the various degradation modes.

  17. 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)

  18. 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

  19. Nanoporous Glasses for Nuclear Waste Containment

    OpenAIRE

    Woignier, Thierry; Primera, Juan; Reynes, Jerôme

    2016-01-01

    Research is in progress to incorporate nuclear waste in new matrices with high structural stability, resistance to thermal shock, and high chemical durability. Interactions with water are important for materials used as a containment matrix for the radio nuclides. It is indispensable to improve their chemical durability to limit the possible release of radioactive chemical species, if the glass structure is attacked by corrosion. By associating high structural stability and high chemical dura...

  20. Standardized analyses of nuclear shipping containers

    International Nuclear Information System (INIS)

    Parks, C.V.; Hermann, O.W.; Petrie, L.M.; Hoffman, T.J.; Tang, J.S.; Landers, N.F.; Turner, W.D.

    1983-01-01

    This paper describes improved capabilities for analyses of nuclear fuel shipping containers within SCALE -- a modular code system for Standardized Computer Analyses for Licensing Evaluation. Criticality analysis improvements include the new KENO V, a code which contains an enhanced geometry package and a new control module which uses KENO V and allows a criticality search on optimum pitch (maximum k-effective) to be performed. The SAS2 sequence is a new shielding analysis module which couples fuel burnup, source term generation, and radial cask shielding. The SAS5 shielding sequence allows a multidimensional Monte Carlo analysis of a shipping cask with code generated biasing of the particle histories. The thermal analysis sequence (HTAS1) provides an easy-to-use tool for evaluating a shipping cask response to the accident capability of the SCALE system to provide the cask designer or evaluator with a computational system that provides the automated procedures and easy-to-understand input that leads to standarization

  1. Long term integrity of reactor pressure vessel and primary containment vessel after the severe accidents in Fukushima Daiichi Nuclear Power Station. Leaching property of spent oxide fuel segment and corrosion property of a carbon steel under artificial seawater immersion

    International Nuclear Information System (INIS)

    2014-06-01

    Primary containment vessel (PCV), reactor pressure vessel and pedestal in Fukushima Daiichi Nuclear power station units 1 through 3 have been exposed to severe thermal, chemical and mechanical conditions due to core meltdown events and seawater injections for emergent core cooling. These components will be immersed in diluted seawater with dissolved fission products under irradiation until the end of debris removal. Fresh water injected into the cores contacts with debris to cool, dissolves or erodes their constituents, mixed with retained water, and becomes 'accumulated water' with radioactive nuclides. We have focused the leaching of fission products into the accumulated water under lower temperature (323 K). FUGEN spent oxide fuel segments were immersed to determine the leaching factor of fission product and actinide elements. Since PCV made from carbon steel is one of the most important boundaries to prevent from fission products release, corrosion behavior has been paid attention to evaluate their integrity. Carbon steel specimens were immersion- and electrochemical-tested in diluted seawater with simulants of the accumulated water at 323 K in order to evaluate the effect of fission products in particular cesium and radiation. (author)

  2. International nuclear fuel cycle fact book. [Contains glossary

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I.W.; Lakey, L.T.; Schneider, K.J.; Silviera, D.J.

    1987-01-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained has been obtained from nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NEA activities reports; proceedings of conferences and workshops; and so forth. Sources do not agree completely with each other, and the data listed herein does not reflect any one single source but frequently is a consolidation/combination of information. Lack of space as well as the intent and purpose of the Fact Book limit the given information to that pertaining to the Nuclear Fuel Cycle and to data considered of primary interest or most helpful to the majority of users.

  3. Nanoporous Glasses for Nuclear Waste Containment

    Directory of Open Access Journals (Sweden)

    Thierry Woignier

    2016-01-01

    Full Text Available Research is in progress to incorporate nuclear waste in new matrices with high structural stability, resistance to thermal shock, and high chemical durability. Interactions with water are important for materials used as a containment matrix for the radio nuclides. It is indispensable to improve their chemical durability to limit the possible release of radioactive chemical species, if the glass structure is attacked by corrosion. By associating high structural stability and high chemical durability, silica glass optimizes the properties of a suitable host matrix. According to an easy sintering stage, nanoporous glasses such as xerogels, aerogels, and composite gels are alternative ways to synthesize silica glass at relatively low temperatures (≈1,000–1,200°C. Nuclear wastes exist as aqueous salt solutions and we propose using the open pore structure of the nanoporous glass to enable migration of the solution throughout the solid volume. The loaded material is then sintered, thereby trapping the radioactive chemical species. The structure of the sintered materials (glass ceramics is that of nanocomposites: actinide phases (~100 nm embedded in a vitreous silica matrix. Our results showed a large improvement in the chemical durability of glass ceramic over conventional nuclear glass.

  4. Containment

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    The primary mission of the Containment Group is to ensure that underground nuclear tests are satisfactorily contained. The main goal is the development of sound technical bases for containment-related methodology. Major areas of activity include siting, geologic description, emplacement hole stemming, and phenomenological predictions. Performance results of sanded gypsum concrete plugs on the Jefferson, Panamint, Cornucopia, Labquark, and Bodie events are given. Activities are also described in the following areas: computational capabilities site description, predictive modeling, and cavity-pressure measurement. Containment publications are listed. 8 references

  5. Catalytic hydrogen recombination for nuclear containments

    International Nuclear Information System (INIS)

    Koroll, G.W.; Lau, D.W.P.; Dewit, W.A.; Graham, W.R.C.

    1994-01-01

    Catalytic recombiners appear to be a credible option for hydrogen mitigation in nuclear containments. The passive operation, versatility and ease of back fitting are appealing for existing stations and new designs. Recently, a generation of wet-proofed catalyst materials have been developed at AECL which are highly specific to H 2 -O 2 , are active at ambient temperatures and are being evaluated for containment applications. Two types of catalytic recombiners were evaluated for hydrogen removal in containments based on the AECL catalyst. The first is a catalytic combustor for application in existing air streams such as provided by fans or ventilation systems. The second is an autocatalytic recombiner which uses the enthalpy of reaction to produce natural convective flow over the catalyst elements. Intermediate-scale results obtained in 6 m 3 and 10 m 3 spherical and cylindrical vessels are given to demonstrate self-starting limits, operating limits, removal capacity, scaling parameters, flow resistance, mixing behaviour in the vicinity of an operating recombiner and sensitivity to poisoning, fouling and radiation. (author). 13 refs., 10 figs

  6. Venting device for nuclear reactor container

    International Nuclear Information System (INIS)

    Yamashita, Masahiro; Ogata, Ken-ichi.

    1994-01-01

    An airtight vessel of a venting device of a nuclear reactor container is connected with a reactor container by way of a communication pipeline. A feed water tank is disposed at a position higher than the liquid surface of scrubbing water in the airtight vessel for supplying scrubbing water to the airtight vessel. In addition, a scrubbing water storage tank is disposed at a position hither than the feed water tank for supplying scrubbing water to the feed water tank. Storage water in the feed water tank is introduced into the airtight vessel by the predetermined opening operation of a valve by the pressure exerted on the liquid surface and the own weight of the storage water. Further, the storage water in the scrubbing water storage tank is led into the feed water tank by the water head pressure. The scrubbing water for keeping the performance of the venting device of the reactor container can be supplied by a highly reliable method without using AC power source or the like as a driving source. (I.N.)

  7. Hydrogen management in nuclear reactor containment

    International Nuclear Information System (INIS)

    Iyer, Kannan

    2014-01-01

    The talk will present the systematic methodology evolved to assess the hydrogen management in nuclear reactor containment during a severe accident. The focus is on the methodology evolved as the full problem is yet to be solved completely. First, the method to quantify mixing of hydrogen is presented. It is demonstrated that buoyancy modified model is adequate to quantify the process satisfactorily. On noting that the hydrogen levels are higher than the safe limits, effort was directed towards mitigating the concentration. Passive Auto-catalytic Recombiners (PAR) were identified as the potential devices for mitigation. Efforts were then directed to model these and a satisfactory one-step reaction derived from a 12 reaction model was evolved. This model was satisfactory when compared with experimental results with hydrogen concentration below 4%. However, the same when extended to hydrogen concentration of 20%, predicts very high concentration thereby indicating the need for experiments at high hydrogen concentration. (author)

  8. The impact of BWR MK I primary containment failure dynamics on secondary containment integrity

    International Nuclear Information System (INIS)

    Greene, S.R.

    1987-01-01

    During the past four years, the ORNL BWRSAT Program has developed a series of increasingly sophisticated BWR secondary containment models. These models have been applied in a variety of studies to evaluate the severe accident mitigation capability of BWR secondary containments. This paper describes the results of a recent ORNL study of the impact of BWR MK I primary containment failure dynamics on secondary containment integrity. A 26-cell MELCOR Browns Ferry secondary containment model is described and the predicted thermodynamic response of the secondary containment to a variety of postulated primary containment failure modes is presented. The effects of primary containment failure location, timing, and ultimate hole size on secondary containment response is investigated, and the potential impact of hydrogen deflagrations on secondary containment integrity is explored

  9. Nuclear primary energy carriers. Short version

    Energy Technology Data Exchange (ETDEWEB)

    Jaeck, W

    1978-04-01

    Basing on our present knowledge the following energy sources for energy supply must be taken into consideration in the long term: regenerative energy sources, fission energy gained by breeder reactors, nuclear fusion. While regenerative energy sources were treated at full length in the study 'Energy Sources for Tomorrow' the present study specifies the other two energy options. The availability and the reliability of nuclear primary energy carrier supply is described in detail and the conversion systems available or still being developed are investigated with regard to their specific consumption of primary energy. Topical questions concerning the proliferation stability of the fuel cycles and techniques are subject to the INFCE program. With reference to the nuclear energy documentation activities of the Federal Government this study is supposed to supply further fundamental material on nuclear primary energy carriers, consumption and readiness for application. Thus it will contribute to the question: 'Is nuclear energy an option which guarantees energy supply in the long term for the Federal Republic of Germany'. (orig.) 891 UA 892 ARA.

  10. Nuclear primary energy carriers. Pt. 1

    International Nuclear Information System (INIS)

    1978-04-01

    Basing on our present knowledge the following energy sources for energy supply must be taken into consideration in the long term: regenerative energy sources, fission energy gained by breeder reactors, nuclear fusion. While regenerative energy sources were treated at full length in the study 'Energy Sources for tomorrow' the present study specifies the other two energy options. The availability and the reliability of nuclear primary energy carrier supply is described in detail and the conversion systems available or still being developed are investigated with regard to their specific consumption of primary energy. Topical questions concerning the proliferation stability of the fuel cycles and techniques are subject to the INFCE programme. With reference to the nuclear energy documentation activities of the Federal Govenment this study is supposed to supply further fundamental material on nuclear primary energy carriers, consumption and readiness for application. Thus it will contribute to the question: 'Is nuclear energy an option which guarantees energy supply in the long term for the Federal Republic of Germany'. (orig.) [de

  11. Inspection of Nuclear Power Plant Containment Structures

    Energy Technology Data Exchange (ETDEWEB)

    Graves, H.L.; Naus, D.J.; Norris, W.E.

    1998-12-01

    Safety-related nuclear power plant (NPP) structures are designed to withstand loadings from a number of low-probability external and interval events, such as earthquakes, tornadoes, and loss-of-coolant accidents. Loadings incurred during normal plant operation therefore generally are not significant enough to cause appreciable degradation. However, these structures are susceptible to aging by various processes depending on the operating environment and service conditions. The effects of these processes may accumulate within these structures over time to cause failure under design conditions, or lead to costly repair. In the late 1980s and early 1990s several occurrences of degradation of NPP structures were discovered at various facilities (e.g., corrosion of pressure boundary components, freeze- thaw damage of concrete, and larger than anticipated loss of prestressing force). Despite these degradation occurrences and a trend for an increasing rate of occurrence, in-service inspection of the safety-related structures continued to be performed in a somewhat cursory manner. Starting in 1991, the U.S. Nuclear Regulatory Commission (USNRC) published the first of several new requirements to help ensure that adequate in-service inspection of these structures is performed. Current regulatory in-service inspection requirements are reviewed and a summary of degradation experience presented. Nondestructive examination techniques commonly used to inspect the NPP steel and concrete structures to identify and quantify the amount of damage present are reviewed. Finally, areas where nondestructive evaluation techniques require development (i.e., inaccessible portions of the containment pressure boundary, and thick heavily reinforced concrete sections are discussed.

  12. Cavity pressure history of contained nuclear explosions

    Energy Technology Data Exchange (ETDEWEB)

    Chapin, C E [Lawrence Radiation Laboratory, University of California, Livermore, CA (United States)

    1970-05-01

    Knowledge of pressure in cavities created by contained nuclear explosions is useful for estimating the possibility of venting radioactive debris to the atmosphere. Measurements of cavity pressure, or temperature, would be helpful in evaluating the correctness of present code predictions of underground explosions. In instrumenting and interpreting such measurements it is necessary to have good theoretical estimates of cavity pressures. In this paper cavity pressure is estimated at the time when cavity growth is complete. Its subsequent decrease due to heat loss from the cavity to the surrounding media is also predicted. The starting pressure (the pressure at the end of cavity growth) is obtained by adiabatic expansion to the final cavity size of the vaporized rock gas sphere created by the explosion. Estimates of cavity size can be obtained by stress propagation computer codes, such as SOC and TENSOR. However, such estimates require considerable time and effort. In this paper, cavity size is estimated using a scheme involving simple hand calculations. The prediction is complicated by uncertainties in the knowledge of silica water system chemistry and a lack of information concerning possible blowoff of wall material during cavity growth. If wall material blows off, it can significantly change the water content in the cavity, compared to the water content in the ambient media. After cavity growth is complete, the pressure will change because of heat loss to the surrounding media. Heat transfer by convection, radiation and conduction is considered, and its effect on the pressure is calculated. Analysis of cavity heat transfer is made difficult by the complex nature of processes which occur at the wall where melting, vaporization and condensation of the gaseous rock can all occur. Furthermore, the melted wall material could be removed by flowing or dripping to the cavity floor. It could also be removed by expansion of the steam contained in the melt (blowoff) and by

  13. Periodic inspection of CANDU nuclear power plant containment components

    International Nuclear Information System (INIS)

    1989-09-01

    This Standard is one in a series intended to provide uniform requirements for CANDU nuclear power plants. It provides requirements for the periodic inspection of containment components including the containment pressure suppression systems

  14. Corrosion process studies in a nuclear waste container

    International Nuclear Information System (INIS)

    Guasp, Ruben A.; Lanzani, Liliana A.; Coronel, Pascual; Bruzzoni, Pablo; Semino, Carlos J.

    1999-01-01

    Latest results on corrosion behavior studies on high activity nuclear waste container are reported. Corrosion evaluation on lead base alloys and modeling to predict carbon steel external container cover generalized corrosion, are the main issues of these studies. (author)

  15. Nuclear waste storage container with metal matrix

    International Nuclear Information System (INIS)

    Sump, K.R.

    1978-01-01

    The invention relates to a storage container for high-level waste having a metal matrix for the high-level waste, thereby providing greater impact strength for the waste container and increasing heat transfer properties

  16. Seal welded cast iron nuclear waste container

    International Nuclear Information System (INIS)

    Filippi, A.M.; Sprecace, R.P.

    1987-01-01

    An article of manufacture is described comprising a cast iron container having an opening at one end and a cast iron plug; a first nickel-carbon alloy fusion weldable insert surrounding the opening and metallurgically bonded to the cast iron container at the one end of the container; a second nickel-carbon alloy insert metallurgically bonded to the cast iron plug located within the opening and surrounded by the first insert the inserts being jointed by a fusion bond in the opening without heating the cast iron container to an austenite formation temperature thereby sealing the interior of the container from the exterior ambient outside the opening; the nickel-carbon alloy containing about 2 to 5 w% carbon; and both the nickel-carbon alloy insert and the cast iron container have a microstructure containing a graphite phase

  17. Nuclear waste storage container with metal matrix

    Science.gov (United States)

    Sump, Kenneth R.

    1978-01-01

    The invention relates to a storage container for high-level waste having a metal matrix for the high-level waste, thereby providing greater impact strength for the waste container and increasing heat transfer properties.

  18. Hydrogen combustion study in the containment of Atucha-I nuclear power plant

    International Nuclear Information System (INIS)

    Baron, J.H.; Gonzalez Videla, E.

    1997-01-01

    In this paper the combustion of hydrogen was modeled and studied in the containment vessel of the Atucha I nuclear power station using the CONTAIN package. The hydrogen comes from the oxidation of metallic materials during the severe accidents proposed. The CONTAIN package is an integrated tool that analyzes the physical, chemical and radiation conditions that affect the containment structure of the radioactive materials unloaded from the primary system during a severe accident in the reactor. (author) [es

  19. Statistical view on nuclear multifragmentation: Primary decays

    International Nuclear Information System (INIS)

    Raduta, A.H.; Raduta, A.R.

    1997-01-01

    An overall view on the universe of primary decays appearing in the process of nuclear multifragmentation via a microcanonical Monte Carlo Metropolis type simulation is given. General characteristics like mass and charge distributions, relative probabilities of evaporation, fission, fragmentation and vaporization, average number of fragments and distributions of a number of intermediate mass fragments offer valuable information about the intimacy of the process. The capability of the model to describe unitary very different breakup regimes is pointed out. Predictions for charge distributions, isotopic yields, and fission mass distributions are compared with experimental data. copyright 1997 The American Physical Society

  20. Nuclear Resonance Fluorescence and Isotopic Mapping of Containers

    Science.gov (United States)

    Johnson, Micah S.; McNabb, Dennis P.

    2009-03-01

    National security programs have expressed interest in developing systems to isotopically map shipping containers, fuel assemblies, and waste barrels for various materials including special nuclear material (SNM). Current radiographic systems offer little more than an ambiguous density silhouette of a container's contents. In this paper we will present a system being developed at LLNL to isotopically map containers using the nuclear resonance fluorescence (NRF) method. Recent experimental measurements on NRF strengths in SNM are discussed.

  1. The WINCON programme - validation of fast reactor primary containment codes

    International Nuclear Information System (INIS)

    Sidoli, J.E.A.; Kendall, K.C.

    1988-01-01

    In the United Kingdom safety studies for the Commercial Demonstration Fast Reactor (CDFR) include an assessment of the capability of the primary containment in providing an adequate containment for defence against the hazards resulting from a hypothetical Whole Core Accident (WCA). The assessment is based on calculational estimates using computer codes supported by measured evidence from small-scale experiments. The hydrodynamic containment code SEURBNUK-EURDYN is capable of representing a prescribed energy release, the sodium coolant and cover gas, and the main containment and safety related internal structures. Containment loadings estimated using SEURBNUK-EURDYN are used in the structural dynamic code EURDYN-03 for the prediction of the containment response. The experiments serve two purposes, they demonstrate the response of the CDFR containment to accident loadings and provide data for the validation of the codes. This paper summarises the recently completed WINfrith CONtainment (WINCON) experiments that studied the response of specific features of current CDFR design options to WCA loadings. The codes have been applied to some of the experiments and a satisfactory prediction of the global response of the model containment is obtained. This provides confidence in the use of the codes in reactor assessments. (author)

  2. Vent control device for nuclear reactor container

    International Nuclear Information System (INIS)

    Kubota, Ryuji.

    1989-01-01

    The present invention concerns automatic prevention of abnormal over-pressure and hydrogen gas flashing in a BWR type reactor container. That is, (1) if the pressure in the container is abnormally increased, the gas in the pressure suppression chamber is released to reduce the pressure thereby preventing over-pressure damage to the container. (2) Then, if exhaust gases are burnt to cause flashing explosion danger for the gases in the reactor container, the gas release is interrupted. The foregoing two functioins are automatically conducted in this device. Specifically, when the pressure in the reactor container reaches a predetermined allowable limit, a remote control operation valve is opened by automatic control means to release the gas in the vessel. Since the gas flow rate at the start of the release exceeds flame propagation velocity, there is no worry for flashing explosion. Further, if the pipeway flow velocity near the atmospheric release is reduced to less than the flame propagation velocity of the hydrogen gas, the opened valve is automatically closed. Accordingly, propagation of hydrogen gas flame into the container thus causing explosion can surely be prevented. (K.M.)

  3. Radiolysis effects on fuel corrosion within a failed nuclear waste container

    International Nuclear Information System (INIS)

    Sunder, S.; Shoeshmith, D.W.; Christensen, H.C.

    2003-01-01

    The concept of geological disposal of used nuclear fuel in corrosion resistant containers is being investigated in several countries. In the Canadian Nuclear Fuel Waste Management Program (CNFWMP), it is assumed that the used fuel will be disposed of in copper containers. Since the predicted lifetimes of these containers are very long (>106 years), only those containers emplaced with an undetected defect will fail within the period for which radionuclide release from the fuel must be considered. Early failure could lead to the entry of water into the container and subsequent release of radionuclides. The release rate of radionuclides from the used fuel will depend upon its dissolution rate. The primary mechanism for release will be the corrosion of the fuel driven by radiolytically-produced oxidants. The studies carried out to determine the effects of water radiolysis on fuel corrosion are reviewed, and some of the procedures used to predict corrosion rates of used fuel in failed nuclear waste containers described. (author)

  4. Seal for an object containing nuclear fuel

    International Nuclear Information System (INIS)

    Scheuerpflug, W.; Nentwich, D.

    1977-01-01

    This seal which cannot be counterfeited, specially for sealing nuclear objects, e.g. fuel rods, not only makes any damage which has taken place obvious, but makes identification according to a key possible. For this purpose a minimum number of 'particles' or small bodies, which are identical but of different permeability, are fixed inside a short tube during 'loading' of the seal in a certain or an accidental sequence. The sequence of the spheres, which represents a key, can only be determined by special electromagnetic measuring equipment. On first opening the seal, this key sequence is irrevocably destroyed. (HP) [de

  5. Nuclear containment systems and in-service inspection status of Korea nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Jihong, Park; Jaekeun, Hong; Banuk, Park [Korea Institute of Machinery and Materials, Dept. of Authorized Test and Evaluation, Kyungnam (Korea, Republic of)

    2007-07-01

    20 unit nuclear power plants in Korea have been operated and maintained since the first unit started in commercial service in 1978. Most recently 4 units were under construction and several units were planned to be constructed. by industries. 4 types of nuclear containment systems have been constructed until now: first, metal containments, then pre-stressed concrete containments with grouted tendon systems, followed by pre-stressed concrete containments with un-grouted tendon systems, and Korea standard nuclear containments. All the nuclear containments should be inspected periodically. Therefore for periodic in-service inspection, several appropriate technical requirements should be applied differently depending on the specific nuclear containment types. With the changes of times, nuclear containment systems have undergone a remarkable change, and finally nuclear containment system of Korea standard nuclear power plant was settled down, and as a matter of course it dominates the trend of present and future nuclear containment systems. Overall in-service inspection results of most Korea nuclear containments have not showed any serious evidence of degradation.

  6. 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.)

  7. Halon 1301 protection system for nuclear containments

    International Nuclear Information System (INIS)

    McHale, E.T.

    1981-01-01

    Halon 1301 can provide protection against any combustion hazard that hydrogen gas might present in an LWR containment following a loss-of-coolant accident. A development program was conducted, comprising analytical study, laboratory experiments and large-scale testing, to define the requirements for a Halon 1301 system and to examine certain operational problems that were hypothesized. Some results of the study are presented in this paper

  8. Severe accidents and nuclear containment integrity (SANCY). SANCY summary report

    Energy Technology Data Exchange (ETDEWEB)

    Lindholm, I. [VTT Processes, Espoo (Finland)

    2004-07-01

    SANCY project investigates physical phenomena related to severe nuclear accidents with importance to Finnish nuclear power plants. Currently the major topics are the ex-vessel coolability issues, long-term severe accident management and containment leak tightness and adoption and development of new calculation tools considering also the needs of the future Olkiluoto 3 plant. SANCY employs both experimental and analytical methods. (orig.)

  9. Leakage detection system in nuclear reactor container

    International Nuclear Information System (INIS)

    Kurosawa, Masahiko.

    1993-01-01

    The present invention comprises an injection means for adding radioactive materials to coolants in a container cooler, a gamma ray amplitude analyzer connected to coolant pipelines and a means for recording/transferring the data of the result of the measurement, a gamma ray amplitude analyzer connected to a drain water sump and a means for recording/transferring the data of the result of the measurement, a gamma ray amplitude analyzer connected to various kinds of pipelines and a means for recording/transferring the data of the result of the measurement, and a data processing means for comparing and analyzing the measured data of each of the gamma ray amplitude analyzers inputted from each of date recording/transferring means. The gamma ray amplitude analysis for each of the pipelines and drain water sump are conducted at an appropriate frequency, and the measured data are compared and analyzed, to improve the detection accuracy for a trace amount of leakage from each of the pressure pipelines and the container cooler coolant pipelines, thereby enabling to specify the pipeline having leakage. Maintenance efficiency is improved, and severe rupture accident in each of pressure pipelines is prevented previously. (N.H.)

  10. Passive filtration of air egressing from nuclear containment

    Energy Technology Data Exchange (ETDEWEB)

    Malloy, III, John D

    2017-09-26

    A nuclear reactor includes a reactor core comprising fissile material disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor. A containment compartment contains the radiological containment. A heat sink includes a chimney configured to develop an upward-flowing draft in response to heated fluid flowing into a lower portion of the chimney. A fluid conduit is arranged to receive fluid from the containment compartment and to discharge into the chimney. A filter may be provided, with the fluid conduit including a first fluid conduit arranged to receive fluid from the containment compartment and to discharge into an inlet of the filter, and a second fluid conduit arranged to receive fluid from an outlet of the filter and to discharge into the chimney. As the draft is developed passively, there is no need for a blower or pump configured to move fluid through the fluid conduit.

  11. The function of single containment and double containment of PWR nuclear power plant

    International Nuclear Information System (INIS)

    Chen Weijing.

    1985-01-01

    The function and structures of single containment and double containment of PWR nuclear power plant were described briefiy. The dissimilarites of diffent type of containments, which effects the impact of environment are discused. The impact of environment, effected by 'source term', containment gas leak rate and diffusion pattern of the released gas, under different operating condition is analysed. Especially, the impact of environment under LOCA accident is fully analysed

  12. Aging management of containment structures in nuclear power plants

    International Nuclear Information System (INIS)

    Naus, D.J.; Oland, C.B.; Ellingwood, B.R.; Graves, H.L. III; Norris, W.E.

    1994-01-01

    Research is being conducted by ORNL under US Nuclear Regulatory Commission (USNRC) sponsorship to address aging management of nuclear power plant containment and other safety-related structures. Documentation is being prepared to provide the USNRC with potential structural safety issues and acceptance criteria for use in continued service evaluations of nuclear power plants. Accomplishments include development of a Structural Materials Information Center containing data and information on the time variation of 144 material properties under the influence of pertinent environmental stressors or aging factors, evaluation of models for potential concrete containment degradation factors, development of a procedure to identify critical structures and degradation factors important to aging management, evaluations of nondestructive evaluation techniques. assessments of European and North American repair practices for concrete, review of parameters affecting corrosion of metals embedded in concrete, and development of methodologies for making current condition assessments and service life predictions of new or existing reinforced concrete structures in nuclear power plants

  13. Crash testing of nuclear fuel shipping containers

    International Nuclear Information System (INIS)

    Jefferson, R.M.; Yoshimura, H.R.

    1977-08-01

    In an attempt to understand the dynamics of extra severe transportation accidents and to evaluate state-of-the-art computational techniques for predicting the dynamic response of shipping casks involved in vehicular system crashes, the Environmental Control Technology Division of ERDA undertook a program with Sandia to investigate these areas. The program encompasses the following distinct major efforts. The first of these utilizes computational methods for predicting the effects of the accident environment and, subsequently, to calculate the damage incurred by a container as the result of such an accident. The second phase involves the testing of 1 / 8 -scale models of transportation systems. Through the use of instrumentation and high-speed motion photography the accident environments and physical damage mechanisms are studied in detail. After correlating the results of these first two phases, a full scale event involving representative hardware is conducted. To date two of the three selected test scenarios have been completed. Results of the program to this point indicate that both computational techniques and scale modeling are viable engineering approaches to studying accident environments and physical damage to shipping casks

  14. Crash testing of nuclear fuel shipping containers

    International Nuclear Information System (INIS)

    Jefferson, R.M.; Yoshimura, H.R.

    1977-12-01

    In an attempt to understand the dynamics of extra severe transportation accidents and to evaluate state-of-the-art computational techniques for predicting the dynamic response of shipping casks involved in vehicular system crashes, the Environmental Control Technology Division of ERDA undertook a program with Sandia to investigate these areas. This program, which began in 1975, encompasses the following distinct major efforts. The first of these utilizes computational methods for predicting the effects of the accident environment and, subsequently, to calculate the damage incurred by a container as the result of such an accident. The second phase involves the testing of 1 / 8 -scale models of transportation systems. Through the use of instrumentation and high-speed motion photography, the accident environments and physical damage mechanisms are studied in detail. After correlating the results of these first two phases, a full scale event involving representative hardware is conducted. To date two of the three selected test scenarios have been completed. Results of the program to this point indicate that both computational techniques and scale modeling are viable engineering approaches to studying accident environments and physical damage to shipping casks

  15. Vulnerability analysis in a pwr nuclear power plant containment building

    OpenAIRE

    Musolas Otaño, Antoni Maria

    2013-01-01

    When supervising a nuclear power plant, the containment building is crucial. Its functions are guaranteeing structural integrity and avoiding leaks in case of accident. Both events are considered of high risk. Once a given overpressure is registered inside the containment building, three possible outputs are considered: serviceability, breakdown, and collapse. The aim is the study of vulnerability. The vulnerability of the containment building under an overpressure is described by the conditi...

  16. Nuclear reactor installation with outer shell enclosing a primary pressure vessel

    International Nuclear Information System (INIS)

    1975-01-01

    The high temperature nuclear reactor installation described includes a fluid cooled nuclear heat source, a primary pressure vessel and outer shell around the primary pressure vessel and acting as a protection for it against outside projectiles. A floor is provided internally dividing the outside shell into two upper and lower sections and an inside wall dividing the lower section into one part containing the primary pressure vessel and a second part, both made pressure tight with respect to each other and with the outside shell and forming with the latter a secondary means of containment [fr

  17. Pressure release in containments of nuclear power stations

    International Nuclear Information System (INIS)

    Pauli, W.; Pellaud, B.; Saitoh, A.

    1992-01-01

    In France, Germany, Sweden and Switzerland, the licensing authorities have decided to equip nuclear reactor containments with a filter venting system to ensure survival of the containment after postulated severe nuclear accidents. This is a curious paradox. For years, the established wisdom was unambiguously 'Keep the containment tight. It's the ultimate barrier.' Three Mile Island seemed to prove the point. Yet, an old mechanical engineer's rule is 'Every pressure vessel must have a safety valve.' Filtered containment venting attempts to reconcile these two conflicting objectives by allowing a filtered pressure relief after an accident, in order to prevent containment failure due to overpressure, while keeping the release within acceptable limits. Achieving this dual objective is a matter of proper timing, i.e. pressure relief, not too early, not too late. (author)

  18. Device for separating, purifying and recovering nuclear fuel material, impurities and materials from impurity-containing nuclear fuel materials or nuclear fuel containing material

    International Nuclear Information System (INIS)

    Sato, Ryuichi; Kamei, Yoshinobu; Watanabe, Tsuneo; Tanaka, Shigeru.

    1988-01-01

    Purpose: To separate, purify and recover nuclear fuel materials, impurities and materials with no formation of liquid wastes. Constitution: Oxidizing atmosphere gases are introduced from both ends of a heating furnace. Vessels containing impurity-containing nuclear fuel substances or nuclear fuel substance-containing material are continuously disposed movably from one end to the other of the heating furnace. Then, impurity oxides or material oxides selectively evaporated from the impurity-containing nuclear fuel substances or nuclear fuel substance-containing materials are entrained in the oxidizing atmosphere gas and the gases are led out externally from a discharge port opened at the intermediate portion of the heating furnace, filters are disposed to the exit to solidify and capture the nuclear fuel substances and traps are disposed behind the filters to solidify and capture the oxides by spontaneous air cooling or water cooling. (Sekiya, K.)

  19. 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

  20. 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.

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

    International Nuclear Information System (INIS)

    Butler, T.A.

    1982-01-01

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

  2. 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.)

  3. 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

  4. Damage assessment of nuclear containment against aircraft crash

    Energy Technology Data Exchange (ETDEWEB)

    Iqbal, Mohd Ashraf, E-mail: iqbal_ashraf@rediffmail.com; Sadique, Md. Rehan, E-mail: rehan.sadique@gmail.com; Bhargava, Pradeep, E-mail: bhpdpfce@iitr.ac.in; Bhandari, N.M., E-mail: nmbcefce@iitr.ac.in

    2014-10-15

    Highlights: • Damage assessment of nuclear containment is studied against aircraft crash. • Four impact locations have been identified at the outer containment shell. • The mid of the total height has been found to be most vulnerable location. • The crown of dome has been found to be the strongest location. • Phantom F4 caused more localized and severe damage compared to other aircrafts. - Abstract: The behavior of nuclear containment structure has been studied against aircraft crash with an emphasis on the influence of strike location. The impact locations identified on the BWR Mark III type nuclear containment structure are mid-height, junction of dome and cylinder, crown of dome and arc of dome. The containment at each of the above locations has been impacted normally by Phantom F-4, Boeing 707-320 and Airbus A320 aircrafts. The loading of the aircraft has been assigned through the corresponding reaction-time response curve. ABAQUS/Explicit finite element code has been used to carry out the three-dimensional numerical simulations. The concrete damaged plasticity model was used to simulate the behavior of concrete while the behavior of steel reinforcement was incorporated using the Johnson–Cook elasto-viscoplastic material model. The mid-height of containment has been found to experience most severe deformation against each aircraft. Phantom F4 has been found to be most disastrous at each location. The results have been compared with those of the available studies with respect to the containment deformation.

  5. 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.)

  6. A lecture on nuclear physics in primary school

    International Nuclear Information System (INIS)

    Arh, S.

    2004-01-01

    I am going to propose the contents of a lecture on nuclear physics and radioactivity in primary school. Contemporary technology, medicine and science exploit intensively the discovered knowledge about processes in atoms and in a nucleus. Mankind has gained huge profit from peaceful applications of nuclear reactions and ionizing radiation. We use the products of nuclear industry every day. But about half of the school population never hears a professional explanation about what is going on in nuclear power plants. Only on some secondary schools students learn about nuclear physics. The lack of knowledge about nuclear processes is the main reason why people show great fear when hearing the words: radiation, radioactivity, nuclear, etc. At last it is now time to give some fundamental lessons on nuclear physics and radioactivity also to pupils in primary school. From my four-year teaching experience in primary school I am suggesting a programme of lectures on nuclear physics and radioactivity. At the end of the lessons we would visit the Krsko Nuclear Power Plant or the Nuclear Training Centre Milan Copic. This could be included in the so called natural science day. Pupils come from the eight class (14 years old) of primary school and have no problems following the explanation. (author)

  7. Adventitious Carbon on Primary Sample Containment Metal Surfaces

    Science.gov (United States)

    Calaway, M. J.; Fries, M. D.

    2015-01-01

    Future missions that return astromaterials with trace carbonaceous signatures will require strict protocols for reducing and controlling terrestrial carbon contamination. Adventitious carbon (AC) on primary sample containers and related hardware is an important source of that contamination. AC is a thin film layer or heterogeneously dispersed carbonaceous material that naturally accrues from the environment on the surface of atmospheric exposed metal parts. To test basic cleaning techniques for AC control, metal surfaces commonly used for flight hardware and curating astromaterials at JSC were cleaned using a basic cleaning protocol and characterized for AC residue. Two electropolished stainless steel 316L (SS- 316L) and two Al 6061 (Al-6061) test coupons (2.5 cm diameter by 0.3 cm thick) were subjected to precision cleaning in the JSC Genesis ISO class 4 cleanroom Precision Cleaning Laboratory. Afterwards, the samples were analyzed by X-ray photoelectron spectroscopy (XPS) and Raman spectroscopy.

  8. 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)

  9. The role of nuclear reactor containment in severe accidents

    International Nuclear Information System (INIS)

    1989-04-01

    The containment is a structural envelope which completely surrounds the nuclear reactor system and is designed to confine the radioactive releases in case of an accident. This report summarises the work of an NEA Senior Group of Experts who have studied the potential role of containment in accidents exceeding design specifications (so-called severe accidents). Some possibilities for enhancing the ability of plants to reduce the risk of significant off-site consequences by appropriate management of the acident have been examined

  10. Permanent cavity seal ring for a nuclear reactor containment arrangement

    International Nuclear Information System (INIS)

    Swidwa, K.J.; Salton, R.B.; Marshall, J.R.

    1990-01-01

    This patent describes a nuclear reactor containment arrangement. It comprises: a reactor pressure vessel which thermally expands and contracts during cyclic operation of the reactor, the vessel having a peripheral wall and a horizontally outwardly extending flange thereon; a containment wall having a shelf, the wall spaced from and surrounding the peripheral wall of the reactor pressure vessel defining an annular expansion gap therebetween, and an annular ring seal extending across the annular expansion gap to provide a water-tight seal therebetween

  11. Processing of nuclear power plant waste streams containing boric acid

    International Nuclear Information System (INIS)

    1996-10-01

    Boric acid is used in PWR type reactor's primary coolant circuit to control the neutron flux. However, boric acid complicates the control of water chemistry of primary coolant and the liquid radioactive waste produced from NPP. The purpose of this report is to provide member states with up-to-date information and guidelines for the treatment and conditioning of boric acid containing wastes. It contains chapters on: (a) characteristics of waste streams; (b) options for management of boric acid containing waste; (c) treatment/decontamination of boric acid containing waste; (d) concentration and immobilization of boric acid containing waste; (e) recovery and re-use of boric acid; (f) selected industrial processes in various countries; and (g) the influence of economic factors on process selection. 72 refs, 23 figs, 5 tabs

  12. Modeling of hydrogen behaviour in a PWR nuclear power plant containment with the CONTAIN code

    International Nuclear Information System (INIS)

    Bobovnik, G.; Kljenak, I.

    2001-01-01

    Hydrogen behavior in the containment during a severe accident in a two-loop Westinghouse-type PWR nuclear power plant was simulated with the CONTAIN code. The accident was initiated with a cold-leg break of the reactor coolant system in a steam generator compartment. In the input model, the containment is represented with 34 cells. Beside hydrogen concentration, the containment atmosphere temperature and pressure and the carbon monoxide concentration were observed as well. Simulations were carried out for two different scenarios: with and without successful actuation of the containment spray system. The highest hydrogen concentration occurs in the containment dome and near the hydrogen release location in the early stages of the accident. Containment sprays do not have a significant effect on hydrogen stratification.(author)

  13. Natural circulating passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  14. Passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  15. Analysis CFD for the hydrogen transport in the primary containment of a BWR

    International Nuclear Information System (INIS)

    Jimenez P, D. A.; Del Valle G, E.; Gomez T, A. M.

    2014-10-01

    This study presents a qualitative and quantitative comparison among the CFD GASFLOW and OpenFOAM codes which are related with the phenomenon of hydrogen transport and other gases in the primary containment of a Boiling Water Reactor (BWR). GASFLOW is a commercial license code that is well validated and that was developed in Germany for the analysis of the gases transport in containments of nuclear reactors. On the other hand, OpenFOAM is an open source code that offers several evaluation solvers for different types of phenomena; in this case, the solver reacting-Foam is used to analyze the hydrogen transport inside the primary containment of the BWR. The results that offer the solver reacting-Foam of OpenFOAM are evaluated in the hydrogen transport calculation and the results are compared with those of the program of commercial license GASFLOW to see if is viable the use of the open source code in the case of the hydrogen transport in the primary containment of a BWR. Of the obtained results so much quantitative as qualitative some differences were identified between both codes, the differences (with a percentage of maximum error of 4%) in the quantitative results are small and they are considered acceptable for this analysis type, also, these differences are attributed mainly to the used transport models, considering that OpenFOAM uses a homogeneous model and GASFLOW uses a heterogeneous model. (Author)

  16. Aircraft crash upon outer containment of nuclear power plant

    International Nuclear Information System (INIS)

    Abbas, H.; Paul, D.K.; Godbole, P.N.; Nayak, G.C.

    1996-01-01

    In this paper, analysis of an aircraft crash upon an outer containment of a nuclear power plant is presented. The effect of target yielding is considered simultaneously by calculating the reaction time in a time marching scheme. The concrete model employed is capable of predicting the cracking and yielding. The response for different cracking strains and different locations of aircraft strike for different aircraft has been studied. Critical location of aircraft strike for the containment has been investigated. The analytical procedure and the material model used are found to be capable of representing the aircraft impact response of the containment structure. (orig.)

  17. 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.)

  18. Nuclear safety review for qualification of class 1E motor inside containment for nuclear power stations

    International Nuclear Information System (INIS)

    Li Shixin; Wu Qi; Zhang Yunbo; Wu Caixia

    2013-01-01

    In nuclear power plants with pressurized water reactors, the review for class 1E motor inside containment qualification process and documents is an important aspect of nuclear safety equipment review, and the reviewers should make evaluations for the qualification test results in terms of the compliance with standard and regulation, and the consistency with inside containment environment. Firstly, this paper introduces the qualification test of class 1E motor inside containment for nuclear power generating stations, such as aging test and design-basis-event test. Second, there is a discussion about typical problems of review. At last, comparison of IEEE334 with RCC-E is conducted and explored. (authors)

  19. A regulatory view of nuclear containment on UK licensed sites

    International Nuclear Information System (INIS)

    Bradford, P.M.; McNair, I.J.

    1997-01-01

    Members of the UK regulatory body, HM Nuclear Installations Inspectorate (NII) have previously presented conference papers and official reports which have dealt separately with either reactor applications or chemical plant applications. The objective of this paper is to draw together a brief overview of the role of containment in protecting against potential radiological and related hazards, and to describe the factors which influence the NII's assessment of containment safety cases. It draws upon the NII's experience of regulating many types of nuclear facility, from those designed in the late 1940s through to the modern plants, such as Sizewell 'B' and THORP. The paper reviews the legislative and regulatory background within which the facilities exist and are operated. Finally, the paper reviews recent, ongoing and planned research in the field of containment, which has been designed to behave under challenge. (author)

  20. Collecting and recirculating condensate in a nuclear reactor containment

    International Nuclear Information System (INIS)

    Schultz, T.L.

    1993-01-01

    An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank. 3 figures

  1. Collecting and recirculating condensate in a nuclear reactor containment

    Science.gov (United States)

    Schultz, Terry L.

    1993-01-01

    An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank.

  2. Aging management of containment structures in nuclear power plants

    International Nuclear Information System (INIS)

    Naus, D.J.; Oland, C.B.; Ellingwood, B.R.; Graves, H.L. III; Norris, W.E.

    1996-01-01

    Research is being conducted by Oak Ridge National Laboratory under US nuclear regulatory commission (USNRC) sponsorship to address aging management of nuclear power plant containment and other safety-related structures. Documentation is being prepared to provide the USNRC with potential structural safety issues and acceptance criteria for use in continued service evaluations of nuclear power plants. Accomplishments include development of a structural materials information center containing data and information on the time variation of 144 material properties under the influence of pertinent environmental stressors or aging factors, evaluation of models for potential concrete containment degradation factors, development of a procedure to identify critical structures and degradation factors important to aging management, evaluations of non-destructive evaluation techniques, assessments of European and North American repair practices for concrete, review of parameters affecting corrosion of metals embedded in concrete, and development of methodologies for making current condition assessments and service life predictions of new or existing reinforced concrete structures in nuclear power plants. (orig.)

  3. Aging management of containment structures in nuclear power plants

    International Nuclear Information System (INIS)

    Naus, D.J.; Oland, C.B.; Ellingwood, B.R.

    1994-01-01

    Research is being conducted by Oak Ridge National Laboratory under U.S. Nuclear Regulatory Commission sponsorship to address aging management of nuclear power plant containment and other safety-related structures. Documentation is being prepared to provide the US-NRC with potential structural safety issues and acceptance criteria for use in continued service evaluations of nuclear power plants. Accomplishments include development of a Structural Materials Information Center containing data and information on the time variation of 144 material properties under the influence of pertinent environmental stressors or aging factors, evaluation of models for potential concrete containment degradation factors, development of a procedure to identify critical structures and degradation factors important to aging management, evaluations of nondestructive evaluation techniques, assessments of European and North American repair practices for concrete, review of parameters affecting corrosion of metals embedded in concrete, and development of methodologies for making current condition assessments and service life predictions of new or existing reinforced concrete structures in nuclear power plants. (author). 29 refs., 2 figs

  4. Sulphate in Liquid Nuclear Waste: from Production to Containment

    Energy Technology Data Exchange (ETDEWEB)

    Lenoir, M.; Grandjean, A.; Ledieu, A.; Dussossoy, J.L.; Cau Dit Coumes, C.; Barre, Y.; Tronche, E. [CEA Marcoule, DEN/DTCD/SECM/LDMC, Batiment 208 BP17171, Bagnols sur Ceze, 30207 (France)

    2009-06-15

    Nuclear industry produces a wide range of low and intermediate level liquid radioactive wastes which can include different radionuclides such as {sup 90}Sr. In La Hague reprocessing plant and in the nuclear research centers of CEA (Commissariat a l'Energie Atomique), the coprecipitation of strontium with barium sulphate is the technique used to treat selectively these contaminated streams with the best efficiency. After the decontamination process, low and intermediate level activity wastes incorporating significant quantities of sulphate are obtained. The challenge is to find a matrix easy to form and with a good chemical durability which is able to confine this kind of nuclear waste. The current process used to contain sulphate-rich nuclear wastes is bituminization. However, in order to improve properties of containment matrices and simplify the process, CEA has chosen to supervise researches on other materials such as cements or glasses. Indeed, cements are widely used for the immobilization of a variety of wastes (low and intermediate level wastes) and they may be an alternative matrix to bitumen. Even if Portland cement, which is extensively used in the nuclear industry, presents some disadvantages for the containment of sulphate-rich nuclear wastes (risk of swelling and cracking due to delayed ettringite formation), other cement systems, such as calcium sulfo-aluminate binders, may be valuable candidates. Another matrix to confine sulphate-rich waste could be the glass. One of the advantages of this material is that it could also immobilize sulphate containing high level nuclear waste which is present in some countries. This waste comes from the use of ferrous sulfamate as a reducing agent for the conversion of Pu{sup 4+} to Pu{sup 3+} in the partitioning stage of the actinides during reprocessing. Sulphate solubility in borosilicate glasses has already been studied in CEA at laboratory and pilot scales. At a pilot scale, low level liquid waste has been

  5. Simulation of hydrogen distribution in an Indian Nuclear Reactor Containment

    Energy Technology Data Exchange (ETDEWEB)

    Prabhudharwadkar, Deoras M. [Department of Mechanical Engineering, Indian Institute of Technology, Mumbai (India); Iyer, Kannan N., E-mail: kiyer@iitb.ac.i [Department of Mechanical Engineering, Indian Institute of Technology, Mumbai (India); Mohan, Nalini; Bajaj, Satinder S. [Nuclear Power Corporation of India Ltd., Mumbai (India); Markandeya, Suhas G. [Bhabha Atomic Research Centre, Trombay, Mumbai (India)

    2011-03-15

    Research highlights: This work addresses hydrogen dispersion in commercial nuclear reactor containment. The numerical tool used for simulation is first benchmarked with experimental data. Parametric results are then carried out for different release configurations. Results lead to the conclusion that the dispersal is buoyancy dominated. Also, the hydrogen concentration is high enough to demand mitigation devices. - Abstract: The management of hydrogen in a Nuclear Reactor Containment after LOCA (Loss Of Coolant Accident) is of practical importance to preserve the structural integrity of the containment. This paper presents the results of systematic work carried out using the commercial Computational Fluid Dynamics (CFD) software FLUENT to assess the concentration distribution of hydrogen in a typical Indian Nuclear Reactor Containment. In order to obtain an accurate estimate of hydrogen concentration distribution, a suitable model for turbulence closure is required to be selected. Using guidelines from the previous studies reported in the literature and a comparative simulation study using simple benchmark problems, the most suitable turbulence model for hydrogen mixing prediction was identified. Subsequently, unstructured meshes were generated to represent the containment of a typical Indian Nuclear Reactor. Analyses were carried out to quantify the hydrogen distribution for three cases. These were (1) Uniform injection of hydrogen for a given period of time at room temperature, (2) Time varying injection as has been computed from an accident analysis code, (3) Time varying injection (as used in case (2)) at a high temperature. A parametric exercise was also carried out in case (1) where the effect of various inlet orientations and locations on hydrogen distribution was studied. The results indicate that the process of hydrogen dispersal is buoyancy dominated. Further for typical injection rates encountered following LOCA, the dispersal is quite poor and most

  6. Simulation of hydrogen distribution in an Indian Nuclear Reactor Containment

    International Nuclear Information System (INIS)

    Prabhudharwadkar, Deoras M.; Iyer, Kannan N.; Mohan, Nalini; Bajaj, Satinder S.; Markandeya, Suhas G.

    2011-01-01

    Research highlights: → This work addresses hydrogen dispersion in commercial nuclear reactor containment. → The numerical tool used for simulation is first benchmarked with experimental data. → Parametric results are then carried out for different release configurations. → Results lead to the conclusion that the dispersal is buoyancy dominated. → Also, the hydrogen concentration is high enough to demand mitigation devices. - Abstract: The management of hydrogen in a Nuclear Reactor Containment after LOCA (Loss Of Coolant Accident) is of practical importance to preserve the structural integrity of the containment. This paper presents the results of systematic work carried out using the commercial Computational Fluid Dynamics (CFD) software FLUENT to assess the concentration distribution of hydrogen in a typical Indian Nuclear Reactor Containment. In order to obtain an accurate estimate of hydrogen concentration distribution, a suitable model for turbulence closure is required to be selected. Using guidelines from the previous studies reported in the literature and a comparative simulation study using simple benchmark problems, the most suitable turbulence model for hydrogen mixing prediction was identified. Subsequently, unstructured meshes were generated to represent the containment of a typical Indian Nuclear Reactor. Analyses were carried out to quantify the hydrogen distribution for three cases. These were (1) Uniform injection of hydrogen for a given period of time at room temperature, (2) Time varying injection as has been computed from an accident analysis code, (3) Time varying injection (as used in case (2)) at a high temperature. A parametric exercise was also carried out in case (1) where the effect of various inlet orientations and locations on hydrogen distribution was studied. The results indicate that the process of hydrogen dispersal is buoyancy dominated. Further for typical injection rates encountered following LOCA, the dispersal is

  7. 75 FR 16645 - Increase in the Primary Nuclear Liability Insurance Premium

    Science.gov (United States)

    2010-04-02

    ... Primary Nuclear Liability Insurance Premium AGENCY: Nuclear Regulatory Commission. ACTION: Final rule... impractical. The NRC is amending its regulations to increase the primary premium for liability insurance... protection requirements and indemnity agreements to increase the primary nuclear liability insurance layer...

  8. Uncanistered Spent Nuclear fuel Disposal Container System Description Document

    International Nuclear Information System (INIS)

    Pettit, N. E.

    2001-01-01

    The Uncanistered Spent Nuclear Fuel (SNF) Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident

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

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  10. Pressure test behaviour of embalse nuclear power plant containment structure

    International Nuclear Information System (INIS)

    Bruschi, S.; Marinelli, C.

    1984-01-01

    It's described the structural behaviour of the containment structure during the pressure test of the Embalse plant (CANDU type, 600MW), made of prestressed concrete with an epoxi liner. Displacement, strain, temperature, and pressure measurements of the containment structure of the Embalse Nuclear Power Plant are presented. The instrumentation set up and measurement specifications are described for all variables of interest before, during and after the pressure test. The analytical models to simulate the heat transfer due to sun heating and air convenction and to predict the associated thermal strains and displacements are presented. (E.G.) [pt

  11. Evaluation of a hydrogen sensor for nuclear reactor containment monitoring

    International Nuclear Information System (INIS)

    Hoffheins, B.S.; McKnight, T.E.; Lauf, R.J.; Smith, R.R.; James, R.E.

    1997-01-01

    Measurement of hydrogen concentration in containment atmospheres in nuclear plants is a key safety capability. Current technologies require extensive sampling systems and subsequent maintenance and calibration costs can be very expensive. A new hydrogen sensor has been developed that is small and potentially inexpensive to install and maintain. Its size and low power requirement make it suitable in distributed systems for pinpointing hydrogen buildup. This paper will address the first phase of a testing program conducted to evaluate this sensor for operation in reactor containments

  12. Study on alternatives of inertisation of nuclear power plant containment

    International Nuclear Information System (INIS)

    Baron, J.H.; Zarate, S.M.

    1998-01-01

    In the course of a severe accident in a nuclear power plant, the hydrogen generation and other flammable gases, during the core degradation phase and the interaction corium-concrete, could produce the failure of the containment by overpressure of by combustion. According to the analysis of the potential effects of hydrogen evolution, following accidents inside the containment trough a Defense-in depth principle, which attempts to assure that the containment must not fail catastrophically, two techniques have been evaluated: a: Inertisation pre-accident and b: Inertisation post-accident. The technique of inertisation pre-accident consists in replacing the air of the containment with inert-gas like nitrogen (N 2 ) or carbon dioxide (CO 2 ) during the normal operation. The inertisation post-accident in combination with early venting system consists in replacing the air of the containment with inert-gas like nitrogen (N 2 ) or carbon dioxide (CO 2 ), immediately after the beginning of the accident, while the radioactivity is still negligible inside the containment. A system of inertisation pre-accident with nitrogen is used on BWR Mark I and Mark II. Investigations on the inertisation post-accident of the containment atmosphere during severe accidents have been carried out with different objectives from principles of the decade of 1980. Studies concerning hydrogen problem for the nuclear power plants Atucha I and CAREM-25 have permitted to know that the hydrogen generation during an accidental sequence with core degradation, would result important, being able to arrive to form explosive mixtures. In the present work, the applicability of the techniques of inertisation is analyzed for the containment of the Atucha I and CAREM-25, considering the particular design characteristics of these plants. (author) [es

  13. Feasibility study of a contained pulsed nuclear propulsion engine

    International Nuclear Information System (INIS)

    Parlos, A.G.; Metzger, J.D.

    1994-01-01

    The result of a feasibility analysis of a contained pulsed nuclear propulsion (CPNP) engine concept utilizing the enormously dense energy generated by small nuclear detonations is presented in this article. This concept was initially proposed and studied in the 1950s and 1960s under the program name HELIOS. The current feasibility of the concept is based upon materials technology that has advanced to a state that allows the design of pressure vessels required to contain the blast associated with small nuclear detonations. The impulsive nature of the energy source provides the means for circumventing the materials thermal barriers that are inherent in steady-state nuclear propulsion concepts. The rapid energy transfer to the propellant results in high thrust levels for times less than 1 s following the detonation. The preliminary feasibility analysis using off-the-shelf materials technology appears to indicate that the CPNP concept can have thrust-to-weight ratios on the order of 1 or greater. Though the specific impulse is not a good indicator for impulsive engines, an operating-cycle averaged specific impulse of approximately 1000 or greater seconds was calculated. 16 refs

  14. Differential lead retention in zircons: implications for nuclear waste containment.

    Science.gov (United States)

    Gentry, R V; Sworski, T J; McKown, H S; Smith, D H; Eby, R E; Christie, W H

    1982-04-16

    An innovative ultrasensitive technique was used for lead isotopic analysis of individual zircons extracted from granite core samples at depths of 960, 2170, 2900, 3930, and 4310 meters. The results show that lead, a relatively mobile element compared to the nuclear waste-related actinides uranium and thorium, has been highly retained at elevated temperatures (105 degrees to 313 degrees C) under conditions relevant to the burial of synthetic rock waste containers in deep granite holes.

  15. Nuclear resonance vibrational spectroscopic studies of iron-containing biomolecules

    International Nuclear Information System (INIS)

    Ohta, Takehiro; Seto, Makoto

    2014-01-01

    In this review, we report recent nuclear resonance vibrational spectroscopic (NRVS) studies of iron-containing biomolecules and their model complexes. The NRVS is synchrotron-based element-specific vibrational spectroscopic methods. Unlike Raman and infrared spectroscopy, the NRVS can investigate all iron motions without selection rules, which provide atomic level insights into the structure/reactivity correlation of biologically relevant iron complexes. (author)

  16. Core catcher for nuclear reactor core meltdown containment

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Bowman, F.L.

    1978-01-01

    A bed of graphite particles is placed beneath a nuclear reactor core outside the pressure vessel but within the containment building to catch the core debris in the event of failure of the emergency core cooling system. Spray cooling of the debris and graphite particles together with draining and flooding of coolant fluid of the graphite bed is provided to prevent debris slump-through to the bottom of the bed

  17. Nuclear criticality safety calculational analysis for small-diameter containers

    International Nuclear Information System (INIS)

    LeTellier, M.S.; Smallwood, D.J.; Henkel, J.A.

    1995-11-01

    This report documents calculations performed to establish a technical basis for the nuclear criticality safety of favorable geometry containers, sometimes referred to as 5-inch containers, in use at the Portsmouth Gaseous Diffusion Plant. A list of containers currently used in the plant is shown in Table 1.0-1. These containers are currently used throughout the plant with no mass limits. The use of containers with geometries or material types other than those addressed in this evaluation must be bounded by this analysis or have an additional analysis performed. The following five basic container geometries were modeled and bound all container geometries in Table 1.0-1: (1) 4.32-inch-diameter by 50-inch-high polyethylene bottle; (2) 5.0-inch-diameter by 24-inch-high polyethylene bottle; (3) 5.25-inch-diameter by 24-inch-high steel can (open-quotes F-canclose quotes); (4) 5.25-inch-diameter by 15-inch-high steel can (open-quotes Z-canclose quotes); and (5) 5.0-inch-diameter by 9-inch-high polybottle (open-quotes CO-4close quotes). Each container type is evaluated using five basic reflection and interaction models that include single containers and multiple containers in normal and in credible abnormal conditions. The uranium materials evaluated are UO 2 F 2 +H 2 O and UF 4 +oil materials at 100% and 10% enrichments and U 3 O 8 , and H 2 O at 100% enrichment. The design basis safe criticality limit for the Portsmouth facility is k eff + 2σ < 0.95. The KENO study results may be used as the basis for evaluating general use of these containers in the plant

  18. Nuclear containment structure subjected to commercial and fighter aircraft crash

    International Nuclear Information System (INIS)

    Sadique, M.R.; Iqbal, M.A.; Bhargava, P.

    2013-01-01

    Highlights: • Nuclear containment response has been studied against aircraft crash. • Concrete damaged plasticity and Johnson–Cook elasto-viscoplastic models were employed. • Boeing 747-400 and Boeing 767-400 aircrafts caused global failure of containment. • Airbus A320 and Boeing 707-320 aircrafts caused local damage. • Tension damage of concrete was found more prominent compared to compression damage. -- Abstract: The response of a boiling water reactor (BWR) nuclear containment vessel has been studied against commercial and fighter aircraft crash using a nonlinear finite element code ABAQUS. The aircrafts employed were Boeing 747-400, Boeing 767-400, Airbus A-320, Boeing 707-320 and Phantom F4. The containment was modeled as a three-dimensional deformable reinforced concrete structure while the loading of aircraft was assigned using the respective reaction–time curve. The location of strike was considered near the junction of dome and cylinder, and the angle of incidence, normal to the containment surface. The material behavior of the concrete was incorporated using the damaged plasticity model while that of the reinforcement, the Johnson–Cook elasto-viscoplastic model. The containment could not sustain the impact of Boeing 747-400 and Boeing 767-400 aircrafts and suffered rupture of concrete around the impact region leading to global failure. On the other hand, the maximum local deformation at the point of impact was found to be 0.998 m, 0.099 m, 0.092 m, 0.089 m, and 0.074 m against Boeing 747-400, Phantom F4, Boeing 767, Boeing 707-320 and Airbus A-320 aircrafts respectively. The results of the present study were compared with those of the previous analytical and numerical investigations with respect to the maximum deformation and overall behavior of the containment

  19. Nuclear containment structure subjected to commercial and fighter aircraft crash

    Energy Technology Data Exchange (ETDEWEB)

    Sadique, M.R., E-mail: rehan.sadique@gmail.com; Iqbal, M.A., E-mail: iqbalfce@iitr.ernet.in; Bhargava, P., E-mail: bhpdpfce@iitr.ernet.in

    2013-07-15

    Highlights: • Nuclear containment response has been studied against aircraft crash. • Concrete damaged plasticity and Johnson–Cook elasto-viscoplastic models were employed. • Boeing 747-400 and Boeing 767-400 aircrafts caused global failure of containment. • Airbus A320 and Boeing 707-320 aircrafts caused local damage. • Tension damage of concrete was found more prominent compared to compression damage. -- Abstract: The response of a boiling water reactor (BWR) nuclear containment vessel has been studied against commercial and fighter aircraft crash using a nonlinear finite element code ABAQUS. The aircrafts employed were Boeing 747-400, Boeing 767-400, Airbus A-320, Boeing 707-320 and Phantom F4. The containment was modeled as a three-dimensional deformable reinforced concrete structure while the loading of aircraft was assigned using the respective reaction–time curve. The location of strike was considered near the junction of dome and cylinder, and the angle of incidence, normal to the containment surface. The material behavior of the concrete was incorporated using the damaged plasticity model while that of the reinforcement, the Johnson–Cook elasto-viscoplastic model. The containment could not sustain the impact of Boeing 747-400 and Boeing 767-400 aircrafts and suffered rupture of concrete around the impact region leading to global failure. On the other hand, the maximum local deformation at the point of impact was found to be 0.998 m, 0.099 m, 0.092 m, 0.089 m, and 0.074 m against Boeing 747-400, Phantom F4, Boeing 767, Boeing 707-320 and Airbus A-320 aircrafts respectively. The results of the present study were compared with those of the previous analytical and numerical investigations with respect to the maximum deformation and overall behavior of the containment.

  20. Containment analysis for the simultaneous detonation of two nuclear explosives

    International Nuclear Information System (INIS)

    Terhune, R.W.; Glenn, H.D.; Burton, D.E.; Rambo, J.T.

    1977-01-01

    The explosive phenomenology associated with the simultaneous detonation of two 2.2-kt nuclear explosives is examined. A comprehensive spatial-time pictorial of the resultant shock-wave phenomenology is given. The explosives were buried at depths of 200 m and 280 m, corresponding to a separation of approximately 4 final cavity radii. Constitutive relations for the surrounding medium were derived from the geophysical logs and core samples taken from an actual emplacement configuration at the Nevada Test Site. Past calculational studies indicate that successful containment may depend upon the development of a strong tangential-stress field (or ''containment cage'') surrounding the cavity at late times. A series of conditions that must be met to insure formation of this cage are presented. Calculational results, based on one- and two-dimensional finite-difference codes of continuum mechanics, describe how each condition has been fulfilled and illustrate the dynamic sequence of events important to the formation of the containment cage. They also indicate, at least for the geological site chosen, that two nuclear explosives do not combine to threaten containment

  1. Seismic analysis of liquid storage container in nuclear reactors

    International Nuclear Information System (INIS)

    Zhang Zhengming; He Shuyan; Xu Ming

    2007-01-01

    Seismic analysis of liquid storage containers is always difficult in the seismic design of nuclear reactor equipment. The main reason is that the liquid will generate significant seismic loads under earthquake. These dynamic liquid loads usually form the main source of the stresses in the container. For this kind of structure-fluid coupling problem, some simplified theoretical methods were usually used previously. But this cannot satisfy the requirements of engineering design. The Finite Element Method, which is now full developed and very useful for the structural analysis, is still not mature for the structure-fluid coupling problem. This paper introduces a method suitable for engineering mechanical analysis. Combining theoretical analysis of the dynamic liquid loads and finite element analysis of the structure together, this method can give practical solutions in the seismic design of liquid storage containers

  2. Photonuclear-based Detection of Nuclear Smuggling in Cargo Containers

    Science.gov (United States)

    Jones, J. L.; Haskell, K. J.; Hoggan, J. M.; Norman, D. R.; Yoon, W. Y.

    2003-08-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) and the Los Alamos National Laboratory (LANL) have performed experiments in La Honda, California and at the Idaho Accelerator Center in Pocatello, Idaho to assess and develop a photonuclear-based detection system for shielded nuclear materials in cargo containers. The detection system, measuring photonuclear-related neutron emissions, is planned for integration with the ARACOR Eagle Cargo Container Inspection System (Sunnyvale, CA). The Eagle Inspection system uses a nominal 6-MeV electron accelerator and operates with safe radiation exposure limits to both container stowaways and to its operators. The INEEL has fabricated custom-built, helium-3-based, neutron detectors for this inspection application and is performing an experimental application assessment. Because the Eagle Inspection system could not be moved to LANL where special nuclear material was available, the response of the Eagle had to be determined indirectly so as to support the development and testing of the detection system. Experiments in California have successfully matched the delayed neutron emission performance of the ARACOR Eagle with that of the transportable INEEL electron accelerator (i.e., the Varitron) and are reported here. A demonstration test is planned at LANL using the Varitron and shielded special nuclear materials within a cargo container. Detector results are providing very useful information regarding the challenges of delayed neutron counting near the photofission threshold energy of 5.5 - 6.0 MeV, are identifying the possible utilization of prompt neutron emissions to allow enhanced signal-to-noise measurements, and are showing the overall benefits of using higher electron beam energies.

  3. Non destructive evaluation of containment nuclear plants structures

    Energy Technology Data Exchange (ETDEWEB)

    Garnier, V. [Aix Marseille Univ., Aix en Provence (France). LMA, CNRS UPR 7051, IUT; Verdier, J. [Toulouse Univ. (France). UPS, INSA, LMDC; Sbartai, Z.M. [Bordeaux Univ., Talence (France). I2M; and others

    2015-07-01

    French Projects of Investment for the Future, called ''Research for Nuclear Safety and Radiation Protection'' have been initiated to further research on the causes, the management, the impact of the observed nuclear accidents and to propose and validate solutions to limit the risk and the consequences. In this context the ''Non Destructive Evaluation of nuclear plants containment'' project (ENDE) with eight partners (six research institutes and two industrials) supported by the ''National Agency of Research'', proposes a methodology for the Non Destructive Evaluation of the containment capacity to fulfil its two major functions: strength and leak tightness. The NDE measurements will be performed under conditions representing the specific solicitations of a decennial inspection, and after or during a reference accident. The project aims to develop NDEs, to combine them by data fusion and to select the most efficient combinations with quantitative criteria. The work is based on: - Structuring the knowledge and developing an experimental plan. - Evaluating the material in representative conditions of accidental solicitations (water saturation, porosity, strength, elastic modulus, stress) and the diffuse thermal damage (micro cracks) - Monitoring the transition from diffuse to continuous damage (cracks) and monitoring a crack under stress (opening and width). - Implementing ND Techniques on-site. The ND techniques retained after selection will be implemented on a containment mock-up on the 1/3 scale. This mock-up developed by EDF (Electricite de France) will be available in 2016. It will be comparable to those of real size containment regarding pressure and temperature conditions. The measures deduced from the NDEs will be introduced in another project (Macena) that aims to simulate the water and heat transfers as well as creep occurring in a reference accident. We will present the methodology and the results

  4. Assessment of materials for nuclear fuel immobilization containers

    Energy Technology Data Exchange (ETDEWEB)

    Nuttall, K; Urbanic, V F

    1981-09-01

    A wide range of engineering metals and alloys has been assessed for their suitability as container materials for irradiated nuclear fuel intended for permanent disposal in a deep, underground hard-rock vault. The expected range of service conditions in the disposal vault are discussed, as well as the material properties required for this application. An important requirement is that the container last at least 500 years without being breached. The assessment is treated in two parts. Part I concentrates on the physical and mechanical metallurgy, with special reference to strength, weldability, potential embrittlement mechanisms and some economic aspects. Part II discusses possible mechanisms of metallic corrosion for the various engineering alloys and the expected range of environmental conditions in the vault. Localized corrosion and delayed fracture processes are identified as being most likely to limit container lifetime. Hence an essential requirement is that such processes either be absent or proceed at an insignificant rate. Three groups of alloys are recommended for further consideration as possible container materials: AISI 300 series austenitic stainless steels, high nickel-base alloys and very dilute titanium-base alloys. Specific alloys from each group are indicated as having the optimum combination of required properties, including cost. For container designs where the outer container shell does not independently support the service loads, copper should also be considered. The final material selection will depend primarily on the enviromental conditions in the vault. 42 figures, 31 tables.

  5. Lead corrosion evaluation in high activity nuclear waste container (Argentina)

    International Nuclear Information System (INIS)

    Guasp, R.; Lanzani, L.; Bruzzoni, P.; Cufre, W.; Semino, C.J.

    2000-01-01

    This report describes a study of high activity nuclear waste canister corrosion in a deep geological disposal. In this canister design, the vitrified nuclear waste stainless steel container is shielded by a 100 mm thick lead wall. For mechanical resistance, the canister will also have a thin carbon steel external liner. Experimental and mathematical modeling studies are aimed to asses the corrosion kinetics of the carbon steel liner in first instance and then, once this liner has been corroded away, the corrosion kinetics of the main lead barrier. Being that oxygen reduction is the main cathodic reaction that supports the anodic oxidation of iron, a model is described predicting the rate of oxygen consumption in a sealed deep nuclear waste disposal vault as a result of the canister corrosion. Oxidation processes other than container corrosion, and that can account also for oxygen depletion, are not taken into consideration. Corrosion experimental studies on lead and its alloys in groundwater are also reported. These experiments are aimed to improve the corrosion resistance of commercial lead in groundwater. (author)

  6. Effects of Containment on Radionuclide Releases from Underground Nuclear Explosions

    Science.gov (United States)

    Carrigan, C. R.; Sun, Y.

    2016-12-01

    Confirming the occurrence of an underground nuclear explosion can require capturing short-lived noble gas radioisotopes produced by the explosion, sometimes referred to as the "smoking gun" for nuclear explosion detection. It is well known that the radioisotopic distribution resulting from the detonation evolves with time in the explosion cavity. In effect, the explosion cavity or chimney behaves as a chemical reactor. As long as the parent and daughter radionuclides remain in a closed and well-mixed cavity, parameters, such as radioxenon isotopic ratios, can be calculated analytically from a decay-chain network model. When gases from the cavity migrate into the containment regime, consideration of a "leaky reactor" model is more appropriate. We consider several implications of such a leaky reactor model relevant to interpretations of gas samples from the subsurface during an on-site inspection that could potentially be carried out under the Comprehensive Nuclear Test Ban Treaty. Additionally, we have attempted to validate our leaky reactor model against atmospheric observations of radioactive xenon isotopes detected by radionuclide monitoring stations in Japan and Russia following the February 2013 DPRK underground nuclear explosion (Carrigan et al., 2016). While both model uncertainty and observational error are significant, our model of isotopic evolution appears to be in broad agreement with radionuclide observations, and for the first time links atmospheric measurements of radioxenon isotopic ratios to estimates of seismic yield. Carrigan et al., Scientific Reports 6, Article number: 23032 (2016) doi:10.1038/srep23032

  7. Alcohol-free alkoxide process for containing nuclear waste

    Science.gov (United States)

    Pope, James M.; Lahoda, Edward J.

    1984-01-01

    Disclosed is a method of containing nuclear waste. A composition is first prepared of about 25 to about 80%, calculated as SiO.sub.2, of a partially hydrolyzed silicon compound, up to about 30%, calculated as metal oxide, of a partially hydrolyzed aluminum or calcium compound, about 5 to about 20%, calculated as metal oxide, of a partially hydrolyzed boron or calcium compound, about 3 to about 25%, calculated as metal oxide, of a partially hydrolyzed sodium, potassium or lithium compound, an alcohol in a weight ratio to hydrolyzed alkoxide of about 1.5 to about 3% and sufficient water to remove at least 99% of the alcohol as an azeotrope. The azeotrope is boiled off and up to about 40%, based on solids in the product, of the nuclear waste, is mixed into the composition. The mixture is evaporated to about 25 to about 45% solids and is melted and cooled.

  8. Risk of containment sump plugging in EDF nuclear power plants

    International Nuclear Information System (INIS)

    2003-10-01

    The fuel of PWR type reactors in usual functioning is cooled by water circulating in the primary circuit, kept at a 155 bars pressure. In the case of break in the circuit, the reactor is stopped automatically. But it is necessary to evacuate the power that continues to escape from the fuel, because of the radioactivity of products created during the reactor operating. In this aim a system called system of security injection allows to send water in the reactor core when the vapor that is releases by the primary circuit opening is condensed in the containment by the spraying system in the containment that sprays water under the containment dome. The two systems of security and spraying are supplied by a tank but when it is empty they are supplied by water recovered in the bottom of containment in sumps. The two systems operate in closed circuit and allow to evacuate the residual power. The purpose of this work is to study the risk of filters clogging that are in the sumps. (N.C.)

  9. Membrane systems and their use in nuclear power plants. Treatment of primary coolant

    Energy Technology Data Exchange (ETDEWEB)

    Kus, Pavel; Bartova, Sarka; Skala, Martin; Vonkova, Katerina [Research Centre Rez, Husinec-Rez (Czech Republic). Technological Circuits Innovation Dept.; Zach, Vaclav; Kopa, Roman [CEZ a.s., Temelin (Czech Republic). Nuclear Power Plant Temelin

    2016-03-15

    In nuclear power plants, drained primary coolant containing boric acid is currently treated in the system of evaporators and by ion exchangers. Replacement of the system of evaporators by membrane system (MS) will result in lower operating cost mainly due to lower operation temperature. In membrane systems the feed primary coolant is separated into two output streams: retentate and permeate. Retentate stream consists of the concentrated boric acid solution together with other components, while permeate stream consists of purified water. Results are presented achieved by testing a pilot-plant unit of reverse osmosis in nuclear power plant (NPP) Temelin.

  10. Detecting nuclear materials smuggling: performance evaluation of container inspection policies.

    Science.gov (United States)

    Gaukler, Gary M; Li, Chenhua; Ding, Yu; Chirayath, Sunil S

    2012-03-01

    In recent years, the United States, along with many other countries, has significantly increased its detection and defense mechanisms against terrorist attacks. A potential attack with a nuclear weapon, using nuclear materials smuggled into the country, has been identified as a particularly grave threat. The system for detecting illicit nuclear materials that is currently in place at U.S. ports of entry relies heavily on passive radiation detectors and a risk-scoring approach using the automated targeting system (ATS). In this article we analyze this existing inspection system and demonstrate its performance for several smuggling scenarios. We provide evidence that the current inspection system is inherently incapable of reliably detecting sophisticated smuggling attempts that use small quantities of well-shielded nuclear material. To counter the weaknesses of the current ATS-based inspection system, we propose two new inspection systems: the hardness control system (HCS) and the hybrid inspection system (HYB). The HCS uses radiography information to classify incoming containers based on their cargo content into "hard" or "soft" containers, which then go through different inspection treatment. The HYB combines the radiography information with the intelligence information from the ATS. We compare and contrast the relative performance of these two new inspection systems with the existing ATS-based system. Our studies indicate that the HCS and HYB policies outperform the ATS-based policy for a wide range of realistic smuggling scenarios. We also examine the impact of changes in adversary behavior on the new inspection systems and find that they effectively preclude strategic gaming behavior of the adversary. © 2011 Society for Risk Analysis.

  11. An assessment of materials for nuclear fuel immobilization containers

    International Nuclear Information System (INIS)

    Nuttall, K.; Urbanic, V.F.

    1981-09-01

    A wide range of engineering metals and alloys was assessed for their suitability as container materials for irradiated nuclear fuel intended for permanent disposal in a deep, underground hard-rock vault. The container must last at least 500 years without being breached. Materials were assessed for their physical and mechanical metallurgy, weldability, potential embrittlement mechanisms, and economics. A study of the possible mechanisms of metallic corrosion for the various engineering alloys and the expected range of environmental conditons in the vault showed that localized corrosion and delayed fracture processes are the most likely to limit container lifetime. Thus such processes either must be absent or proceed at an insignificant rate. Three groups of alloys are recommended for further study: AISI 300 series austenitic stainless steels, high nickel-base alloys and very dilute titanium-base alloys. Specific alloys from each group are indicated as having the optimum combination of required properties, including cost. For container designs where the outer container shell does not independently support the service loads, copper should also be considered. The final material selection will depend primarily on the environmental conditions in the vault

  12. Characterization of nuclear reactor containment penetrations. Preliminary report

    International Nuclear Information System (INIS)

    Bump, T.R.; Seidensticker, R.W.; Shackelford, M.A.; Gambhir, V.K.; McLennan, G.L.

    1984-06-01

    This report summarizes the survey work conducted by Argonne National Laboratory on the design and details of major penetrations in 22 nuclear power plants. The survey includes all containment types and materials in current use. It also includes details of all types of penetrations (except for electrical penetration assemblies and valves) and the seals and gaskets used in them. The report provides a test matrix for testing major penetrations and for testing seals and gaskets in order to evaluate their leakage potential under severe accident conditions

  13. Detection of nuclear material by photon activation inside cargo containers

    Science.gov (United States)

    Gmar, Mehdi; Berthoumieux, Eric; Boyer, Sébastien; Carrel, Frédérick; Doré, Diane; Giacri, Marie-Laure; Lainé, Frédéric; Poumarède, Bénédicte; Ridikas, Danas; Van Lauwe, Aymeric

    2006-05-01

    Photons with energies above 6 MeV can be used to detect small amounts of nuclear material inside large cargo containers. The method consists in using an intense beam of high-energy photons (bremsstrahlung radiation) in order to induce reactions of photofission on actinides. The measurement of delayed neutrons and delayed gammas emitted by fission products brings specific information on localization and quantification of the nuclear material. A simultaneous measurement of both of these delayed signals can overcome some important limitations due to matrix effects like heavy shielding and/or the presence of light elements as hydrogen. We have a long experience in the field of nuclear waste package characterization by photon interrogation and we have demonstrated that presently the detection limit can be less than one gram of actinide per ton of package. Recently we tried to extend our knowledge to assess the performance of this method for the detection of special nuclear materials in sea and air freights. This paper presents our first results based on experimental measurements carried out in the SAPHIR facility, which houses a linear electron accelerator with the energy range from 15 MeV to 30 MeV. Our experiments were also modeled using the full scale Monte Carlo techniques. In addition, and in a more general frame, due to the lack of consistent data on photonuclear reactions, we have been working on the development of a new photonuclear activation file (PAF), which includes cross sections for more than 600 isotopes including photofission fragment distributions and delayed neutron tables for actinides. Therefore, this work includes also some experimental results obtained at the ELSA electron accelerator, which is more adapted for precise basic nuclear data measurements.

  14. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    Energy Technology Data Exchange (ETDEWEB)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E. [Sandia National Labs., Albuquerque, NM (United States); Tills, J. [J. Tills and Associates, Inc., Sandia Park, NM (United States)

    1997-12-01

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.

  15. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    International Nuclear Information System (INIS)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E.; Tills, J.

    1997-12-01

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions

  16. The stress corrosion cracking of copper nuclear waste containers

    International Nuclear Information System (INIS)

    King, F.; Litke, C.D.; Ikeda, B.M.

    1999-01-01

    The extent of stress corrosion cracking (SCC) of copper nuclear waste containers is being predicted on the basis of a 'limited propagation' argument. In this argument, it is accepted that crack initiation may occur, but it is argued that the environmental conditions and material properties required for a through-wall crack to propagate will not be present. In this paper, the effect of one environmental parameter, the supply of oxidant (J ox ), on the crack growth rate is examined. Experiments have been conducted on two grades of Cu in NANO 2 environments using two loading techniques. The supply of oxidant has been varied either electrochemically in bulk solution using different applied current densities or by embedding the loaded test specimens in compacted buffer material containing O 2 as the oxidant. Measured and theoretical crack growth rates as a function of J ox are compared with the predicted oxidant flux to the containers in a disposal vault and an estimate of the maximum crack depth on a container obtained. (author)

  17. The stress corrosion cracking of copper nuclear waste containers

    International Nuclear Information System (INIS)

    King, F.; Litke, C.D.; Ikeda, B.M.

    1999-01-01

    The extent of stress corrosion cracking (SCC) of copper nuclear waste containers is being predicted on the basis of a limited propagation argument. In this argument, it is accepted that crack initiation may occur, but it is argued that the environmental conditions and material properties required for a through-wall crack to propagate will not be present. In this paper, the effect of one environmental parameter, the supply of oxidant (J OX ), on the crack growth rate is examined. Experiments have been conducted on two grades of Cu in NaNO 2 environments using two loading techniques. The supply of oxidant has been varied either electrochemically in bulk solution using different applied current densities or by embedding the loaded test specimens in compacted buffer material containing O 2 as the oxidant. Measured and theoretical crack growth rates as a function of J OX are compared with the predicted oxidant flux to the containers in a disposal vault and an estimate of the maximum crack depth on a container obtained

  18. 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

  19. Physical modeling of spent-nuclear-fuel container

    Directory of Open Access Journals (Sweden)

    Wang Liping

    2012-11-01

    Full Text Available A new physical simulation model was developed to simulate the casting process of the ductile iron heavy section spent-nuclear-fuel container. In this physical simulation model, a heating unit with DR24 Fe-Cr-Al heating wires was used to compensate the heat loss across the non-natural surfaces of the sample, and a precise and reliable casting temperature controlling/monitoring system was employed to ensure the thermal behavior of the simulated casting to be similar to the actual casting. Also, a mould system was designed, in which changeable mould materials can be used for both the outside and inside moulds for different applications. The casting test was carried out with the designed mould and the cooling curves of central and edge points at different isothermal planes of the casting were obtained. Results show that for most isothermal planes, the temperature control system can keep the temperature differences within 6 ℃ between the edge points and the corresponding center points, indicating that this new physical simulation model has high simulation accuracy, and the mould developed can be used for optimization of casting parameters of spent-nuclear-fuel container, such as composition of ductile iron, the pouring temperature, the selection of mould material and design of cooling system. In addition, to maintain the spheroidalization of the ductile iron, the force-chilling should be used for the current physical simulation to ensure the solidification of casting in less than 2 h.

  20. Iodine removal in containment filtered venting system during nuclear accident

    International Nuclear Information System (INIS)

    Bera, Subrata; Deo, Anuj Kumar; Nagrale, D.B.; Paul, U.K.; Prasad, M.; Gaikwad, A.J.

    2015-01-01

    Post Fukushima nuclear accident, containment filtered venting system is being introduced in Indian nuclear power plant to strengthen the defense in depth safety barrier by depressurizing the containment building along with minimization of radioactivity release to environment during a severe accident. Radioactive iodine is one of the major contributors to radiation dose during early release phase of a severe accident. Physical and Chemical form of iodine and iodine bearing compounds includes particulates, elemental and organic. In the most efficient design of CFVS, wet scrubbing mechanism has been employed through use of venture scrubber. The Iodine removal process in wet scrubber involves two processes: chemical reaction in highly alkaline aqueous solution and impingement of particulates with water droplets produced in the venturi nozzle. In this paper, venturi has been modeled using the Calvert model. The variation of efficiency has been estimated for the different particle sizes. The impact of the shape parameter of log-normal distribution on the amount of scrubbed iodine has also been assessed. Release phase wise the scrubbed amount of iodine in the venturi based CFVS system has been estimated for a typical BWR. (author)

  1. A feasibility study for a contained pulsed nuclear propulsion concept

    International Nuclear Information System (INIS)

    Parlos, A.G.; Metzger, J.D.

    1993-01-01

    A preliminary analysis of a pulsed propulsion concept is performed utilizing the enormously dense energy generated by small nuclear detonations. The concept feasibility is based on the premise that current materials technology has undergone significant breakthroughs, allowing design of pressure vessels capable of containing the blast associated with such detonations. Furthermore, the rapid energy transfer to the propellant, allows generation of high thrust levels for up to 10 ms following the detonation. Preliminary reevaluation of the concept using off-the-shelf materials technology appears to indicate that the contained pulsed nuclear propulsion concept has no major flaws, and it can provide thrust levels resulting in average thrust-to-weight ratios on the order of 2--2.5 over an engine operating cycle. Furthermore, even though the specific impulse is not a good performance indicator for impulsive engines, operating-cycle-averaged specific impulse of approximately 1800 s has been calculated. The engine mass associated with this performance is on the order of 50 Mg. The concept appears attractive for a number of missions planned for the Space Exploration Initiative, however, there are still a number of issues that must be addressed

  2. Primary processes initiated by nuclear transformations in solids

    International Nuclear Information System (INIS)

    Sano, Hirotoshi

    1975-01-01

    Primary processes of hot atom production initiated by nuclear transformation were discussed from past studies using Moessbauer spectroscopy. Many insulators (dielectric substances) showed various effect, such as abnormaly oxdized condition, following nuclear disintegration within the time duration of the life of Moessbauer nuclear excited state. Supposing these hot atom processes belonged to radiochemical processes, radiochemical characteristics of a certain chemical substance could be clarified by placing Moessbauer nuclide in the neighbourhood of the chemical substance to be studied. Chemical effects of disintegrated atom in the first and second composition, chemical substances produced in the surroundings of disintegrated atom, and environmental disturbance of disintegrated atom were studied and discussed. (Tsukamoto, Y.)

  3. Leakage evaluation in the PCV (Primary Containment Vessel) using chemical and radiochemical data

    International Nuclear Information System (INIS)

    Maeda, Katsuji; Nagasawa, Katsumi

    1998-01-01

    Keeping the reliability of nuclear power plant operation, the primary coolant leakage in the PCV is strictly restricted by the Technical Specifications. It is very important to detect an indication of leakage and estimate the source of leakage to provide countermeasures. Usually the indication of leakage will be detected by increase of drain flow in the PCV sump. There are some possibilities of leakage sources in the PCV, such as reactor water, main steam, condensate, feedwater and closed cooling water. The leakage source contain different chemical and radiochemical species. This means that the leakage source can be presumed and detected by using chemical information from the PCV atmosphere and sump water. To detect the leakage indication and the source quickly and exactly, the PCV Leakage Detection Expert System has been developed. This paper describes how to evaluate the leakage indication and source in the PCV by using chemical and radiochemical data. (author)

  4. Method for the chemical reprocessing of irradiated nuclear fuels, in particular nuclear fuels containing uranium

    International Nuclear Information System (INIS)

    Koch, G.

    1976-01-01

    In the chemical processing of irradiated uranium-containing nuclear fuels which are hydrolyzed with aqueous nitric acid, a suggestion is made to use as quaternary ammonium nitrate trialkyl-methyl ammonium nitrates as extracting agent, in which the sum of C atoms is greater than 16. In the illustrated examples, tricaprylmethylammonium nitrate, trilaurylmethylammonium nitrate and tridecylmethylammonium nitrate are named. (HPH/LH) [de

  5. Evaluation of prestress losses in nuclear reactor containments

    International Nuclear Information System (INIS)

    Lundqvist, Peter; Nilsson, Lars-Olof

    2011-01-01

    Research highlights: → Prestress losses in reactor containments were estimated using prediction models. → The predicted prestress losses were compared to long-term measurements. → The accuracy of the models was improved by considering actual drying conditions. → Predictions by CEB/FIP MC 1999 and ACI 209 were closest to the measured losses. - Abstract: The most critical safety barrier in a nuclear power plant, the concrete containment, is prestressed by hundreds of tendons, both horizontally and vertically. The main purpose of the containment is to prevent radioactive discharge to the environment in the case of a serious internal accident. Due to creep and shrinkage of concrete and relaxation of the prestressing steel, tendon forces decrease with time. These forces are thus measured in Swedish containments with unbonded tendons at regular in-service inspections. In this paper, the prestress losses obtained from these in-service inspections are compared to losses estimated using several prediction models for creep, shrinkage and relaxation. In an attempt to increase the accuracy of these models, existing expressions for the development of shrinkage were modified using previous findings on the humidity and temperature inside two Swedish containments. The models which were used and modified for predicting creep and shrinkage were CEB-FIP Model Codes 1990 and 1999, ACI 209, Model B3 and GL2000. Eurocode 2 was used for the prediction of relaxation. The results show that the most accurate of the models were CEB/FIP MC 99 and ACI 209. Depending on the model, the accuracy of the prediction models was increased by 0.5-1.2 percentage points of prestress losses when using the modified development of shrinkage. Furthermore, it was found that the differences between the different models depend mainly on the prediction of creep. Possible explanations for the deviation between the calculated and measured models can be the influence of reinforcement on creep and shrinkage

  6. Evaluation of prestress losses in nuclear reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Lundqvist, Peter, E-mail: peter.lundqvist@kstr.lth.s [Div. of Structural Engineering, Lund University, Lund (Sweden); Nilsson, Lars-Olof [Div. of Building Materials, Lund University, Lund (Sweden)

    2011-01-15

    Research highlights: Prestress losses in reactor containments were estimated using prediction models. The predicted prestress losses were compared to long-term measurements. The accuracy of the models was improved by considering actual drying conditions. Predictions by CEB/FIP MC 1999 and ACI 209 were closest to the measured losses. - Abstract: The most critical safety barrier in a nuclear power plant, the concrete containment, is prestressed by hundreds of tendons, both horizontally and vertically. The main purpose of the containment is to prevent radioactive discharge to the environment in the case of a serious internal accident. Due to creep and shrinkage of concrete and relaxation of the prestressing steel, tendon forces decrease with time. These forces are thus measured in Swedish containments with unbonded tendons at regular in-service inspections. In this paper, the prestress losses obtained from these in-service inspections are compared to losses estimated using several prediction models for creep, shrinkage and relaxation. In an attempt to increase the accuracy of these models, existing expressions for the development of shrinkage were modified using previous findings on the humidity and temperature inside two Swedish containments. The models which were used and modified for predicting creep and shrinkage were CEB-FIP Model Codes 1990 and 1999, ACI 209, Model B3 and GL2000. Eurocode 2 was used for the prediction of relaxation. The results show that the most accurate of the models were CEB/FIP MC 99 and ACI 209. Depending on the model, the accuracy of the prediction models was increased by 0.5-1.2 percentage points of prestress losses when using the modified development of shrinkage. Furthermore, it was found that the differences between the different models depend mainly on the prediction of creep. Possible explanations for the deviation between the calculated and measured models can be the influence of reinforcement on creep and shrinkage of concrete and

  7. Axisymmetric modeling of prestressing tendons in nuclear containment dome

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Se-Jin [DAEWOO E and C, Institute of Construction Technology, 60 Songjook-dong, Jangan-gu, Suwon, Kyonggi 440-210 (Korea, Republic of)]. E-mail: jsj@dwconst.co.kr; Chung, Chul-Hun [Department of Civil and Environmental Engineering, Dankook University, San 8, Hannam-dong, Youngsan-gu, Seoul 140-714 (Korea, Republic of)

    2005-12-15

    Simple axisymmetric modeling of a nuclear containment building has been often employed in practice to estimate structural behavior for the axisymmetric loadings such as internal pressure. In this case, the prestressing tendons placed in the containment dome should be axisymmetrically approximated, since most dome tendons are not arranged in an axisymmetric manner. Some procedures are proposed that can realistically implement the actual three-dimensional tendon stiffness and prestressing effect into the axisymmetric model. Prestressing tendons, which are arranged in two or three ways depending on a containment type, are converted into the equivalent layer to consider the stiffness contribution in meridional and hoop directions. In order to reflect the prestressing effect, the equivalent load method and the initial stress method are devised, respectively, and the corresponding loads or stresses are derived in terms of the axisymmetric model. The proposed schemes are verified through some numerical examples comparing the results of the axisymmetric models to those of the actual three-dimensional model. The examples show that the proper level of the prestressing in the hoop direction of the axisymmetric dome plays an important role in tracing the actual behavior induced by the prestressing. Finally, some correction factors are discussed that can further improve the analysis results.

  8. Containment hydrogen removal system for a nuclear power plant

    International Nuclear Information System (INIS)

    Callaghan, V.M.; Flynn, E.P.; Pokora, B.M.

    1984-01-01

    A hydrogen removal system (10) separates hydrogen from the containment atmosphere of a nuclear power plant using a hydrogen permeable membrane separator (30). Water vapor is removed by condenser (14) from a gas stream withdrawn from the containment atmosphere. The gas stream is then compressed by compressor (24) and cooled (28,34) to the operating temperature of the hydrogen permeable membrane separator (30). The separator (30) separates the gas stream into a first stream, rich in hydrogen permeate, and a second stream that is hydrogen depleted. The separated hydrogen is passed through a charcoal adsorber (48) to adsorb radioactive particles that have passed through the hydrogen permeable membrane (44). The hydrogen is then flared in gas burner (52) with atmospheric air and the combustion products vented to the plant vent. The hydrogen depleted stream is returned to containment through a regenerative heat exchanger (28) and expander (60). Energy is extracted from the expander (60) to drive the compressor (24) thereby reducing the energy input necessary to drive the compressor (24) and thus reducing the hydrogen removal system (10) power requirements

  9. Axisymmetric modeling of prestressing tendons in nuclear containment dome

    International Nuclear Information System (INIS)

    Jeon, Se-Jin; Chung, Chul-Hun

    2005-01-01

    Simple axisymmetric modeling of a nuclear containment building has been often employed in practice to estimate structural behavior for the axisymmetric loadings such as internal pressure. In this case, the prestressing tendons placed in the containment dome should be axisymmetrically approximated, since most dome tendons are not arranged in an axisymmetric manner. Some procedures are proposed that can realistically implement the actual three-dimensional tendon stiffness and prestressing effect into the axisymmetric model. Prestressing tendons, which are arranged in two or three ways depending on a containment type, are converted into the equivalent layer to consider the stiffness contribution in meridional and hoop directions. In order to reflect the prestressing effect, the equivalent load method and the initial stress method are devised, respectively, and the corresponding loads or stresses are derived in terms of the axisymmetric model. The proposed schemes are verified through some numerical examples comparing the results of the axisymmetric models to those of the actual three-dimensional model. The examples show that the proper level of the prestressing in the hoop direction of the axisymmetric dome plays an important role in tracing the actual behavior induced by the prestressing. Finally, some correction factors are discussed that can further improve the analysis results

  10. A program to assess microbial impacts on nuclear waste containment

    International Nuclear Information System (INIS)

    Horn, J.; Meike, A.

    1996-01-01

    In this paper we discuss aspects of a comprehensive program to identify and bound potential effects of microorganisms on long-term nuclear waste containment, using as examples, studies conducted within the Yucca Mountain Project. A comprehensive program has been formulated which cuts across standard disciplinary lines to address the specific concerns of microbial activity in a radioactive waste repository. Collectively, this program provides bounding parameters of microbial activities that modify the ambient geochemistry and hydrology, modify corrosion rates, and transport and transform radionuclides under conditions expected to be encountered after geological waste emplacement. This program is intended to provide microbial reaction rates and bounding conditions in a form that can be integrated into existing chemical and hydrological models. The inclusion of microbial effects will allow those models to more accurately assess long term repository integrity

  11. Gamma radiation induced changes in nuclear waste glass containing Eu

    Science.gov (United States)

    Mohapatra, M.; Kadam, R. M.; Mishra, R. K.; Kaushik, C. P.; Tomar, B. S.; Godbole, S. V.

    2011-10-01

    Gamma radiation induced changes were investigated in sodium-barium borosilicate glasses containing Eu. The glass composition was similar to that of nuclear waste glasses used for vitrifying Trombay research reactor nuclear waste at Bhabha Atomic Research Centre, India. Photoluminescence (PL) and electron paramagnetic resonance (EPR) techniques were used to study the speciation of the rare earth (RE) ion in the matrix before and after gamma irradiation. Judd-Ofelt ( J- O) analyses of the emission spectra were done before and after irradiation. The spin counting technique was employed to quantify the number of defect centres formed in the glass at the highest gamma dose studied. PL data suggested the stabilisation of the trivalent RE ion in the borosilicate glass matrix both before and after irradiation. It was also observed that, the RE ion distributes itself in two different environments in the irradiated glass. From the EPR data it was observed that, boron oxygen hole centre based radicals are the predominant defect centres produced in the glass after irradiation along with small amount of E’ centres. From the spin counting studies the concentration of defect centres in the glass was calculated to be 350 ppm at 900 kGy. This indicated the fact that bulk of the glass remained unaffected after gamma irradiation up to 900 kGy.

  12. Evolution of elevated containment temperatures at Calvert Cliffs Nuclear Power Plant

    International Nuclear Information System (INIS)

    Branch, R.D. Jr.

    1991-01-01

    In this paper the author describes the events which caused Calvert Cliffs Nuclear Power Plant engineers to recognize a need for monitoring of ambient temperatures within containment. The early attempts at temperature monitoring programs are discussed and critiqued. Primary failings of these early programs included a failure to collect temperature data under a variety of external conditions and a lack of quality assurance to make the data useful for design change. From these early attempts Calvert Cliffs developed a new, extensive temperature monitoring program designed to collect data over a two-year period. The author outlines the planned temperature monitoring program and discusses its expected results

  13. Constitutive model for evaluation of nuclear containment structures

    Energy Technology Data Exchange (ETDEWEB)

    Gocevski, Vladimir [Hydro-Quebec, 75 Rene-Levesque Boulevard, West Montreal, QC H2Z 1A4 (Canada)

    2006-09-15

    The paper presents the new constitutive relations for a homogenized reinforced concrete material. Two-stage homogenization procedure is described, i.e. prior to cracking (Phase I) and after cracking (Phase II) of the concrete matrix. Hence, the localization phenomenon and the 'size effect' are properly described. The constitutive law incorporated in the main algorithm of the commercially available finite element code COSMOS/M is further discussed. The model is applied to simulate some relevant aging mechanisms. Therefore, in the proposed paper the assessment of the prestressed concrete aging of the containment structure of Gentilly-2 nuclear power plant using an advanced numerical procedure will be presented. Aging mechanisms considered possible are discussed, the present conditions are assessed and the mechanisms that are likely to impair proper future functioning of the structure are identified. The results of the numerical analysis of the reinforced concrete structure subjected to loads such as thermal and seismic loads are presented and discussed. Attention is given to the analysis of the effects of concrete swelling due to alkali-aggregate reaction. The paper also includes an evaluation of a potential damage in the context of a high velocity impact of a commercial aircraft into the containment structure. (author)

  14. Analysis of long-term behaviour of nuclear reactor containment

    Energy Technology Data Exchange (ETDEWEB)

    Hora, Z. [Czech Technical University in Prague, Faculty of Civil Engineering, Department of Mechanics, Thakurova 7, 166 29 Prague 6 (Czech Republic)]. E-mail: Zbynek.Hora@fsv.cvut.cz; Patzak, B. [Czech Technical University in Prague, Faculty of Civil Engineering, Department of Mechanics, Thakurova 7, 166 29 Prague 6 (Czech Republic)

    2007-02-15

    For assessment of safety and durability of a nuclear power plant (NPP), knowledge of the containment behaviour under various service and extreme conditions is crucial. To perform reliable analysis of such a large-scale structure, a sufficiently realistic but still feasible numerical model must be used, in which the relevant physical phenomena are reflected. Therefore, a constitutive model for concrete including effects of moisture and heat transfer, cement hydration, creep, shrinkage and optionally microcracking of concrete should be chosen. The present paper focuses on the simulation of the service life of NPP containment, aiming to determine the material and model parameters to enable reliable prediction of structural behaviour under various conditions. The purpose of the work is to provide a numerical model calibrated using existing measurements to predict the long-term behaviour reliably. Extensive in situ measurements are used to calibrate the model and to check the validity of the model hypotheses. Moreover, the material model parameters are systematically re-calibrated based on the continuous monitoring of the structure. The structural integrity test is reanalysed numerically to show the model capability of predicting behaviour of the structure under given loading and climate conditions.

  15. Modeling of thermophoretic deposition of aerosols in nuclear reactor containments

    International Nuclear Information System (INIS)

    Fernandes, A.; Loyalka, S.K.

    1996-01-01

    Aerosol released in postulated or real nuclear reactor accidents can deposit on containment surfaces via motion induced by temperature gradients in addition to the motion due to diffusion and gravity. The deposition due to temperature gradients is known as thermophoretic deposition, and it is currently modeled in codes such as CONTAIN in direct analogy with heat transfer, but there have been questions about such analogies. This paper focuses on a numerical solution of the particle continuity equation in laminar flow condition characteristics of natural convection. First, the thermophoretic deposition rate is calculated as a function of the Prandtl and Schmidt numbers, the thermophoretic coefficient K, and the temperature difference between the atmosphere and the wall. Then, the cases of diffusion alone and a boundary-layer approximation (due to Batchelor and Shen) to the full continuity equation are considered. It is noted that an analogy with heat transfer does not hold, but for the circumstances considered in this paper, the deposition rates from the diffusion solution and the boundary-layer approximation can be added to provide reasonably good agreement (maximum deviation 30%) with the full solution of the particle continuity equation. Finally, correlations useful for implementation in the reactor source term codes are provided

  16. 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

  17. Container materials in environments of corroded spent nuclear fuel

    Science.gov (United States)

    Huang, F. H.

    1996-07-01

    Efforts to remove corroded uranium metal fuel from the K Basins wet storage to long-term dry storage are underway. The multi-canister overpack (MCO) is used to load spent nuclear fuel for vacuum drying, staging, and hot conditioning; it will be used for interim dry storage until final disposition options are developed. Drying and conditioning of the corroded fuel will minimize the possibility of gas pressurization and runaway oxidation. During all phases of operations the MCO is subjected to radiation, temperature and pressure excursions, hydrogen, potential pyrophoric hazard, and corrosive environments. Material selection for the MCO applications is clearly vital for safe and efficient long-term interim storage. Austenitic stainless steels (SS) such as 304L SS or 316L SS appear to be suitable for the MCO. Of the two, Type 304L SS is recommended because it possesses good resistance to chemical corrosion, hydrogen embrittlement, and radiation-induced corrosive species. In addition, the material has adequate strength and ductility to withstand pressure and impact loading so that the containment boundary of the container is maintained under accident conditions without releasing radioactive materials.

  18. Lift-based up-ender and methods using same to manipulate a shipping container containing unirradiated nuclear fuel

    Science.gov (United States)

    Nilles, Michael J.

    2017-08-01

    A shipping container containing an unirradiated nuclear fuel assembly is lifted off the ground by operating a crane to raise a lifting tool comprising a winch. The lifting tool is connected with the shipping container by a rigging line connecting with the shipping container at a lifting point located on the shipping container between the top and bottom of the shipping container, and by winch cabling connecting with the shipping container at the top of the shipping container. The shipping container is reoriented by operating the winch to adjust the length of the winch cabling so as to rotate the shipping container about the lifting point. Shortening the winch cabling rotates the shipping container about the lifting point from a horizontal orientation to a vertical orientation, while lengthening the winch cabling rotates the shipping container about the lifting point from the vertical orientation to the horizontal orientation.

  19. Direct containment heating integral effects tests in geometries of European nuclear power plants

    International Nuclear Information System (INIS)

    Meyer, Leonhard; Albrecht, Giancarlo; Caroli, Cataldo; Ivanov, Ivan

    2009-01-01

    The DISCO test facility at Forschungszentrum Karlsruhe (FZK) has been used to perform experiments to investigate direct containment heating (DCH) effects during a severe accident in European nuclear power plants, comprising the EPR, the French 1300 MWe plant P'4, the VVER-1000 and the German Konvoi plant. A high-temperature iron-alumina melt is ejected by steam into scaled models of the respective reactor cavities and the containment vessel. Both heat transfer from dispersed melt and combustion of hydrogen lead to containment pressurization. The main experimental findings are presented and critical parameters are identified. The consequences of DCH are limited in reactors with no direct pathway between the cavity and the containment dome (closed pit). The situation is more severe for reactors which do have a direct pathway between the cavity and the containment (open pit). The experiments showed that substantial fractions of corium may be dispersed into the containment in such cases, if the pressure in the reactor coolant system is elevated at the time of RPV failure. Primary system pressures of 1 or 2 MPa are sufficient to lead to full scale DCH effects. Combustion of the hydrogen produced by oxidation as well as the hydrogen initially present appears to be the crucial phenomenon for containment pressurization.

  20. Direct containment heating integral effects tests in geometries of European nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Leonhard [Forschungszentrum Karlsruhe (FZK), Postfach 3640, 76021 Karlsruhe (Germany)], E-mail: meyer@iket.fzk.de; Albrecht, Giancarlo [Forschungszentrum Karlsruhe (FZK), Postfach 3640, 76021 Karlsruhe (Germany); Caroli, Cataldo [Institut de Radioprotection et de Surete Nucleaire, BP 17, 92262 Fontenay-aux-Roses Cedex (France); Ivanov, Ivan [Technical University of Sofia, BG-1797 Sofia (Bulgaria)

    2009-10-15

    The DISCO test facility at Forschungszentrum Karlsruhe (FZK) has been used to perform experiments to investigate direct containment heating (DCH) effects during a severe accident in European nuclear power plants, comprising the EPR, the French 1300 MWe plant P'4, the VVER-1000 and the German Konvoi plant. A high-temperature iron-alumina melt is ejected by steam into scaled models of the respective reactor cavities and the containment vessel. Both heat transfer from dispersed melt and combustion of hydrogen lead to containment pressurization. The main experimental findings are presented and critical parameters are identified. The consequences of DCH are limited in reactors with no direct pathway between the cavity and the containment dome (closed pit). The situation is more severe for reactors which do have a direct pathway between the cavity and the containment (open pit). The experiments showed that substantial fractions of corium may be dispersed into the containment in such cases, if the pressure in the reactor coolant system is elevated at the time of RPV failure. Primary system pressures of 1 or 2 MPa are sufficient to lead to full scale DCH effects. Combustion of the hydrogen produced by oxidation as well as the hydrogen initially present appears to be the crucial phenomenon for containment pressurization.

  1. Artifacts Generated During Azoalkane Peroxy Radical Oxidative Stress Testing of Pharmaceuticals Containing Primary and Secondary Amines.

    Science.gov (United States)

    Nefliu, Marcela; Zelesky, Todd; Jansen, Patrick; Sluggett, Gregory W; Foti, Christopher; Baertschi, Steven W; Harmon, Paul A

    2015-12-01

    We report artifactual degradation of pharmaceutical compounds containing primary and secondary amines during peroxy radical-mediated oxidative stress carried out using azoalkane initiators. Two degradation products were detected when model drug compounds dissolved in methanol/water were heated to 40°C with radical initiators such as 2,2'-azobis(2-methylpropionitrile) (AIBN). The primary artifact was identified as an α-aminonitrile generated from the reaction of the amine group of the model drug with formaldehyde and hydrogen cyanide, generated as byproducts of the stress reaction. A minor artifact was generated from the reaction between the amine group and isocyanic acid, also a byproduct of the stress reaction. We report the effects of pH, initiator/drug molar ratio, and type of azoalkane initiator on the formation of these artifacts. Mass spectrometry and nuclear magnetic resonance were used for structure elucidation, whereas mechanistic studies, including stable isotope labeling experiments, cyanide analysis, and experiments exploring the effects of butylated hydroxyanisole addition, were employed to support the degradation pathways. © 2015 Wiley Periodicals, Inc. and the American Pharmacists Association.

  2. Corrosion issues in high-level nuclear waste containers

    Science.gov (United States)

    Asl, Samin Sharifi

    In this dissertation different aspects of corrosion and electrochemistry of copper, candidate canister material in Scandinavian high-level nuclear waste disposal program, including the thermodynamics and kinetics of the reactions that are predicted to occur in the practical system have been studied. A comprehensive thermodynamic study of copper in contact with granitic groundwater of the type and composition that is expected in the Forsmark repository in Sweden has been performed. Our primary objective was to ascertain whether copper would exist in the thermodynamically immune state in the repository, in which case corrosion could not occur and the issue of corrosion in the assessment of the storage technology would be moot. In spite of the fact that metallic copper has been found to exist for geological times in granitic geological formations, copper is well-known to be activated from the immune state to corrode by specific species that may exist in the environment. The principal activator of copper is known to be sulfur in its various forms, including sulfide (H2S, HS-, S2-), polysulfide (H2Sx, HSx -, Sx 2-), poly sulfur thiosulfate ( SxO3 2-), and polythionates (SxO6 2-). A comprehensive study of this aspect of copper chemistry has never been reported, and yet an understanding of this issue is vital for assessing whether copper is a suitable material for fabricating canisters for the disposal of HLNW. Our study identifies and explores those species that activate copper; these species include sulfur-containing entities as well as other, non-sulfur species that may be present in the repository. The effects of temperature, solution pH, and hydrogen pressure on the kinetics of the hydrogen electrode reaction (HER) on copper in borate buffer solution have been studied by means of steady-state polarization measurements, including electrochemical impedance spectroscopy (EIS). In order to obtain electrokinetic parameters, such as the exchange current density and the

  3. Nuclear reactor installation with outer shell enclosing a primary pressure vessel

    International Nuclear Information System (INIS)

    1975-01-01

    The high temperature nuclear reactor installation described includes a fluid cooled nuclear heat source, a primary pressure vessel containing the heat source, an outer shell enclosing the primary pressure vessel and acting as a secondary means of containment for this vessel against outside projectiles. Multiple auxiliary equipment points are arranged outside the outer shell which comprises a part of a lower wall around the primary pressure vessel, an annular part integrated in the lower wall and extending outwards as from this wall and an upper part integrated in the annular part and extending above this annular part and above the primary pressure vessel. The annular part and the primary pressure vessel are formed with vertical penetrations which can be closed communicating respectively with the auxiliary equipment points and with inside the pressure vessel whilst handling gear is provided in the upper part for vertically raising reactor components through these penetrations and for transporting them over the annular part and over the primary pressure vessel [fr

  4. Chemical mining of primary copper ores by use of nuclear technology

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, A E [Lawrence Radiation Laboratory, University of California, Livermore, CA (United States)

    1970-05-15

    Chemical mining of primary copper ores, with nuclear explosives to break the ore and in-situ hydrostatic pressure to accelerate dissolution of primary ore minerals, may be feasible. A contained nuclear explosion well below the water table would be used to provide a mass of broken ore in a flooded 'chimney'. The hydrostatic pressure in the chimney should increase the solubility of oxygen in a water-sulfuric acid system enough to allow primary copper minerals such as chalcopyrite and bornite to be dissolved in an acceptably short time. Circulation and collection would be accomplished through drill holes. This method should be especially applicable to the deep portions of porphyry copper deposits that are not economical to mine by present techniques. (author)

  5. System for chemical decontamination of nuclear reactor primary systems

    International Nuclear Information System (INIS)

    Schlonski, J.S.; McGiure, M.F.; Corpora, G.J.

    1992-01-01

    This patent describes a method of chemically decontaminating a nuclear reactor primary system, having a residual heat removal system with one or more residual heat removal heat exchangers, each having an upstream and a downstream side, at or above ambient pressure. It comprises: injecting decontamination chemicals using an injection means; circulating the injected decontamination chemicals throughout the primary system; directing the circulated decontamination chemicals and process fluids to a means for removing suspended solids and dissolved materials after the circulated chemicals and process fluids have passed through the residual heat removal heat exchanger; decontaminating the process fluids; and feeding the decontaminated process fluids to the injection means. This patent also describes a chemical decontamination system for use at, or above, ambient pressure in a nuclear reactor primary system having a residual heat removal system. It comprises: means for injecting decontamination chemicals into the primary system; means for removing dissolved and suspended materials and decontamination chemicals from the primary system; one or more residual heat removal pumps; means located downstream of one of the residual heat removal heat exchangers; and a return line connecting the means

  6. Primary loop simulation of the SP-100 space nuclear reactor

    International Nuclear Information System (INIS)

    Borges, Eduardo M.; Braz Filho, Francisco A.; Guimaraes, Lamartine N.F.

    2011-01-01

    Between 1983 and 1992 the SP-100 space nuclear reactor development project for electric power generation in a range of 100 to 1000 kWh was conducted in the USA. Several configurations were studied to satisfy different mission objectives and power systems. In this reactor the heat is generated in a compact core and refrigerated by liquid lithium, the primary loops flow are controlled by thermoelectric electromagnetic pumps (EMTE), and thermoelectric converters produce direct current energy. To define the system operation point for an operating nominal power, it is necessary the simulation of the thermal-hydraulic components of the space nuclear reactor. In this paper the BEMTE-3 computer code is used to EMTE pump design performance evaluation to a thermalhydraulic primary loop configuration, and comparison of the system operation points of SP-100 reactor to two thermal powers, with satisfactory results. (author)

  7. Numerical simulation of aircraft crash on nuclear containment structure

    Energy Technology Data Exchange (ETDEWEB)

    Iqbal, M.A., E-mail: iqbalfce@iitr.ernet.in [Department of Civil Engineering, Indian Institute of Technology Roorkee, Roorkee 247667 (India); Rai, S.; Sadique, M.R.; Bhargava, P. [Department of Civil Engineering, Indian Institute of Technology Roorkee, Roorkee 247667 (India)

    2012-02-15

    Highlights: Black-Right-Pointing-Pointer The deformation was more localised at the center of cylindrical portion. Black-Right-Pointing-Pointer The peak deflection at the junction of dome and cylinder was found to be 67 mm. Black-Right-Pointing-Pointer The peak deflection at midpoint of the cylindrical portion was found to be 88.9 mm. Black-Right-Pointing-Pointer The strain rate was found to be an important parameter to effect the deformation. Black-Right-Pointing-Pointer The model without strain rate and 290 s{sup -1} strain rate predicted very high deformations. - Abstract: Numerical simulations were carried with ABAQUS/Explicit finite element code in order to predict the response of BWR Mark III type nuclear containment against Boeing 707-320 aircraft crash. The load of the aircraft was applied using and force history curve. The damaged plasticity model was used to predict the behavior of concrete while the Johnson-Cook elasto-viscoplastic material model was used to incorporate the behavior of steel reinforcement. The crash was considered to occur at two different locations i.e., the midpoint of the cylindrical portion and the junction of dome and cylinder. The midpoint of the cylindrical portion experienced more deformation. The strain rate in the material model was varied and found to have a significant effect on the response of containment. The results of the present investigation were compared with those of the studies available in literature and a close agreement with the previous results was found in terms of maximum target deformation.

  8. Oxygen Containment System Options for Nuclear Thermal Propulsion Testing

    Data.gov (United States)

    National Aeronautics and Space Administration — All nuclear thermal propulsion (NTP) ground testing conducted in the 1950s and 1960s during the ROVER/(Nuclear Engine Rocket Vehicle Application (NERVA) program...

  9. 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

  10. 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.

  11. Radionuclide transport through perforations in nuclear waste containers

    International Nuclear Information System (INIS)

    Aidun, C.K.; Bloom, S.G.; Raines, G.E.

    1987-11-01

    Previous analytical models for the steady-state radionuclide release rate through perforations in nuclear waste containers into the surrounding medium are based on a zero wall thickness assumption. In this paper we investigate the effect of the wall thickness on the mass transfer rate through isolated cylindrical holes. We solve the steady-state diffusion equation for the concentration field and derive a model based on the analytical solution. By direct comparison, we show that the zero wall thickness model overpredicts the mass transfer rate by about 1300 percent for a circular hole with 1-cm radius and a wall thickness of 10 cm. As expected, the zero-thickness model becomes even less accurate as the hole radius decreases; it predicts a greater release rate from a large number of small holes than the mass transfer rate from an uncontained waste form cylinder. In contrast, the results predicted by our model remain bounded for isolated holes and never exceed the mass transfer from an uncontained waste form. 6 refs., 9 figs., 3 tabs

  12. Characterization of nuclear reactor containment penetrations. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Shackelford, M.H.; Bump, T.R.; Seidensticker, R.W.

    1985-02-01

    This report concludes a preliminary report prepared by ANL for Sandia, published as NUREG/CR-3855, in June 1984. The preliminary report, NUREG/CR-3855, presented the results of a survey of nuclear reactor containment penetrations, covering the number of plants surveyed at that time (22 total). Since that time, an additional 26 plants have been included in the survey. This final report serves two purposes: (1) to add the summary data sheets and penetration details for the additional plants now included in the survey; and (2) to confirm, revise, or add to analyses and discussions presented in the first report which, of course, were based solely on the earlier sample of 22 plants. This final report follows the outline and format of the preliminary survey report. In general, changes and additions to the preliminary report are implied, rather than stated as such to avoid repeated reference to that report. If no changes have been made in a section the title of the section of the previous report is simply repeated followed by ''No Changes''. Some repetition is used for continuity and clarity.

  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. The role of inertial containment fusion in replacing nuclear tests

    Energy Technology Data Exchange (ETDEWEB)

    Schaper, Annette [Hessische Stiftung Friedens- und Konfliktforschung, Frankfurt am Main (Germany)

    2008-07-01

    Nuclear weapon physicists need to understand the process of a nuclear explosion, and their major experimental tools had been nuclear tests. Since a couple of years, the established nuclear weapon states observe a testing moratorium. Nevertheless, they still want to keep their nuclear arsenals, and consequently to ensure the reliability, safety, and security of their nuclear warheads. For this purpose, they use experimental tools that replace nuclear tests, among them ICF. ICF plays a central role in the so-called ''stockpile stewardship program'' that the U.S. has implemented when it participated in the negotiations on a Comprehensive Test Ban Treaty. Several questions arise and are discussed in the presentation: Does ICF allow to simulate the extreme conditions of a nuclear explosion? Which are the functions of nuclear testing that ICF can replace and which are beyond its capabilities? Would ICF be a useful tool for the design of new nuclear warheads? Why are so huge sums spent on ICF in a military context although the usefulness for nuclear weapons seems rather limited?.

  15. A containment analysis for SBLOCA in the refurbished Wolsong-1 Nuclear Power Plant

    International Nuclear Information System (INIS)

    Kim, Tech-Mo; Park, Jong-Ho

    2011-01-01

    Highlights: → The CANDU safety analysis has been accomplished for the refurbished Wolsong-1 NPP. → GOTHIC and SMART-IST codes and new methodology are used for the containment analysis. → The parametric studies for Iodine Chemistry (IMOD-2) model are performed. → And, IMOD-2 model is very sensitive to paint thickness and dousing water pH. → The radioactive doses to the public in SBLOCA event are far below the acceptable limits. - Abstract: A small break leading to a loss of coolant accident (SBLOCA), being one of the topic accidents in the nuclear plant diagnosis in recent years, has been analysed and evaluated for the refurbished Wolsong-1 Nuclear Power Plant (NPP). The Industry Standard Toolset (IST) codes developed by CANDU (Canadian Deuterium Uranium reactor) Owners Group and updated models including design change parameters are applied newly to the event analyses. The GOTHIC code has been used for the containment thermal-hydraulic analysis of Wolsong-1. Also, the SMART-IST code fitted in the Iodine Chemistry (IMOD-2) model has been used to predict nuclide behavior within the containment considering various aspects. The IMOD-2 was incorporated into SMART-IST as a module dealing with chemical transformations and mass transfer of iodine species in containment. IMOD-2 model is very sensitive to paint and chemicals. The parametric studies for the IMOD-2 model are performed to decide the analysis value set. The iodine release amount increases as the paint thickness increases. But, the iodine release amount increases as the water pH (dousing and primary heat transport (PHT)) decreases. The developed containment analysis methodology and the results of SBLOCA without Emergency Coolant Injection (ECI) are presented herein. Under the most heat-up conditions, the radionuclide release from the failed fuel into the containment and subsequently to the environment is such that the radioactive doses to the public are below the acceptable limits.

  16. Ultimate capacity and influenced factors analysis of nuclear RC containment subjected to internal pressure

    International Nuclear Information System (INIS)

    Song Chenning; Hou Gangling; Zhou Guoliang

    2014-01-01

    Ultimate compressive bearing capacity, influenced factors and its rules of nuclear RC containment are key problems of safety assessment, accident treatment and structure design, etc. Ultimate compressive bearing capacity of nuclear RC containment is shown by concrete damaged plasticity model and steel double liner model of ABAQUS. The study shows that the concrete of nuclear RC containment cylinder wall becomes plastic when the internal pressure is up to 0.87 MPa, the maximum tensile strain of steel liner exceeds 3000 × 10 6 and nuclear RC containment reaches ultimate status when the internal pressure is up to 1.02 MPa. The result shows that nuclear RC containment is in elastic condition under the design internal pressure and the bearing capacity meets requirement. Prestress and steel liner play key parts in the ultimate internal pressure and failure mode of nuclear RC containment. The study results have value for the analysis of ultimate compressive bearing capacity, structure design and safety assessment. (authors)

  17. CONTAIN LMR/1B-Mod.1, A computer code for containment analysis of accidents in liquid-metal-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Murata, K.K.; Carroll, D.E.; Bergeron, K.D.; Valdez, G.D.

    1993-01-01

    The CONTAIN computer code is a best-estimate, integrated analysis tool for predicting the physical, chemical, and radiological conditions inside a nuclear reactor containment building following the release of core material from the primary system. CONTAIN is supported primarily by the U. S. Nuclear Regulatory Commission (USNRC), and the official code versions produced with this support are intended primarily for the analysis of light water reactors (LWR). The present manual describes CONTAIN LMR/1B-Mod. 1, a code version designed for the analysis of reactors with liquid metal coolant. It is a variant of the official CONTAIN 1.11 LWR code version. Some of the features of CONTAIN-LMR for treating the behavior of liquid metal coolant are in fact present in the LWR code versions but are discussed here rather than in the User's Manual for the LWR versions. These features include models for sodium pool and spray fires. In addition to these models, new or substantially improved models have been installed in CONTAIN-LMR. The latter include models for treating two condensables (sodium and water) simultaneously, sodium atmosphere and pool chemistry, sodium condensation on aerosols, heat transfer from core-debris beds and to sodium pools, and sodium-concrete interactions. A detailed description of each of the above models is given, along with the code input requirements

  18. Stress corrosion cracking of candidate materials for nuclear waste containers

    International Nuclear Information System (INIS)

    Maiya, P.S.; Shack, W.J.; Kassner, T.F.

    1989-09-01

    Types 304L and 316L stainless steel (SS), Incoloy 825, Cu, Cu-30%Ni, and Cu-7%Al have been selected as candidate materials for the containment of high-level nuclear waste at the proposed Yucca Mountain Site in Nevada. The susceptibility of these materials to stress corrosion cracking has been investigated by slow-strain-rate tests (SSRTs) in water which simulates that from well J-13 (J-13 water) and is representative of the groundwater present at the Yucca Mountain site. The SSRTs were performed on specimens exposed to simulated J-13 water at 93 degree C and at a strain rate 10 -7 s -1 under crevice conditions and at a strain rate of 10 -8 s -1 under both crevice and noncrevice conditions. All the tests were interrupted after nominal elongation strains of 1--4%. Examination by scanning electron microscopy showed some crack initiation in virtually all specimens. Optical microscopy of metallographically prepared transverse sections of Type 304L SS suggests that the crack depths are small (<10 μm). Preliminary results suggest that a lower strain rate increases the severity of cracking of Types 304L and 316L SS, Incoloy 825, and Cu but has virtually no effect on Cu-30%Ni and Cu-7%Al. Differences in susceptibility to cracking were evaluated in terms of a stress ratio, which is defined as the ratio of the increase in stress after local yielding in the environment to the corresponding stress increase in an identical test in air, both computed at the same strain. On the basis of this stress ratio, the ranking of materials in order of increasing resistance to cracking is: Types 304L SS < 316L SS < Incoloy 825 congruent Cu-30%Ni < Cu congruent Cu-7%Al. 9 refs., 12 figs., 7 tabs

  19. Requirements for the coatings of a nuclear power plant containment building

    International Nuclear Information System (INIS)

    Orantie, K.; Kuosa, H.; Haekkae-Roennholm, E.

    2001-06-01

    The report presents the criteria for the inside coatings of nuclear power plant containment buildings including: radiation resistance, decontamination, chemical resistance in accident situations and fire resistance

  20. Nuclear fuel assembly incorporating primary and secondary structural support members

    International Nuclear Information System (INIS)

    Carlson, W.R.; Gjertsen, R.K.; Miller, J.V.

    1987-01-01

    A nuclear fuel assembly, comprising: (a) an upper end structure; (b) a lower end structure; (c) elongated primary structural members extending longitudinally between and rigidly interconnecting the upper and lower end structures, the upper and lower end structures and primary structural members together forming a rigid structural skeleton of the fuel assembly; (d) transverse grids supported on the primary structural members at axially spaced locations therealong between the upper and lower end structures; (e) fuel rods extending through and supported by the grids between the upper and lower end structures so as to extend in generally side-by-side spaced relation to one another and to the primary structural members; and (f) elongated secondary structural members extending longitudinally between but unconnected with the upper and lower end structures, the secondary structural members extending through and rigidly interconnected with the grids to extend in generally side-by-side spaced relation to one another, to the fuel rods and to the primary structural members so as to bolster the stiffness of the structural skeleton of the fuel assembly

  1. Effect of decontamination on nuclear power plant primary circuit materials

    International Nuclear Information System (INIS)

    Brezina, M.; Kupca, L.

    1991-01-01

    The effect of repeated decontamination on the properties of structural materials of the WWER-440 primary coolant circuit was examined. Three kinds of specimens of 08Kh18Ni10T steel were used for radioactivity-free laboratory experiments; they included material obtained from assembly additions to the V-2 nuclear power plant primary piping, and a sheet of the CSN 17247 steel. Various chemical, electrochemical and semi-dry electrochemical decontamination procedures were tested. Chemical decontamination was based on the conventional AP(20/5)-CITROX(20/20) procedure and its variants; NP-CITROX type procedures with various compositions were also employed. Solutions based on oxalic acid were tested for the electrochemical and semi-dry electrochemical decontamination. The results of measurements of mass losses of the surfaces, of changes in the corrosion resistance and in the mechanical properties of the materials due to repeated decontamination are summarized. (Z.S.). 12 figs., 1 tab., 8 refs

  2. New CSA guideline for the exemption or clearance from regulatory control of materials that contain, or potentially contain, nuclear substances

    International Nuclear Information System (INIS)

    Rhodes, M.; Kwong, A.

    2011-01-01

    The Canadian Standards Association (CSA) guideline N292.5, Guideline for the exemption or clearance from regulatory control of materials that contain, or potentially contain, nuclear substances, was recently developed to address a need for guidance on approaches for clearance of materials from facilities licensed by the Canadian Nuclear Safety Commission (CNSC) consistent with Canadian and international recommendations. This guideline is also applicable to determining if an activity associated with materials that contain nuclear substances is exempt from requiring a CNSC licence. The guideline summarizes the regulatory requirements associated with the exemption and clearance of materials and provides a graded approach to designing a survey based on the risk of residual contamination being present. (author)

  3. Emergency cooling process and device for nuclear reactor containment

    International Nuclear Information System (INIS)

    Costes, D.

    1985-01-01

    The emergency cooling system of a PWR containment, according to the principal patent, comprises a turbine fed by the humid air of the containment, a condenser in which the air flowing out of the turbine is dryed and cooled by an external coolant and a compressor actuated by the turbine and returning the dryed air in the containment [fr

  4. Nuclear Containment Inspection Using AN Array of Guided Wave Sensors for Damage Localization

    Science.gov (United States)

    Cobb, A. C.; Fisher, J. L.

    2010-02-01

    Nuclear power plant containments are typically both the last line of defense against the release of radioactivity to the environment and the first line of defense to protect against intrusion from external objects. As such, it is important to be able to locate any damage that would limit the integrity of the containment itself. Typically, a portion of the containment consists of a metallic pressure boundary that encloses the reactor primary circuit. It is made of thick steel plates welded together, lined with concrete and partially buried, limiting areas that can be visually inspected for corrosion damage. This study presents a strategy using low frequency (<50 kHz) guided waves to find corrosion-like damage several meters from the probe in a mock-up of the containment vessel. A magnetostrictive sensor (MsS) is scanned across the width of the vessel, acquiring waveforms at a fixed interval. A beam forming strategy is used to localize the defects. Experimental results are presented for a variety of damage configurations, demonstrating the efficacy of this technique for detecting damage smaller than the ultrasonic wavelength.

  5. North Korea’s 2009 Nuclear Test: Containment, Monitoring, Implications

    Science.gov (United States)

    2010-04-02

    inspections as prima facie evidence of a violation. One generally-accepted means of evading detection of nuclear tests, especially low-yield tests...In an attempt to extend these bans to cover all nuclear tests, negotiations on the CTBT were completed in 1996. The treaty’s basic obligation is to...Verification refers to determining whether a nation is in compliance with its treaty obligations , which in this case means determining whether a suspicious

  6. Process for testing noise emission from containers or pipelines made of steel, particularly for nuclear reactor plants

    International Nuclear Information System (INIS)

    Votava, E.; Stipsits, G.; Sommer, R.

    1982-01-01

    In a process for noise emission testing of steel containers or pipelines, particularly for testing primary circuit components of nuclear reactor plants, measuring sensors and/or associated electronic amplifiers are used, which are tuned for receiving the frequency band of the sound emission spectrum above a limiting frequency f G , but are limited or non-resonant for frequency bands less than f G . (orig./HP) [de

  7. Performance of the primary containment of a BWR during a severe accident whit the code RELAP/SCDAPSIM

    International Nuclear Information System (INIS)

    Castillo G, F.

    2015-01-01

    In this thesis work, it was developed a model of the vacuum breaker valves and down comers for a BWR Mark II primary containment for the code RELAP/SCDAPSIM Mod. 3.4. This code was used to simulate a Station Blackout (Sbo) that evolves to a severe accident scenario. To accomplish this task, the vacuum breaker valves and down comers were included in a simplified model of the primary containment that includes both wet well and dry well, which was coupled with a model of the Nuclear Steam Supply System (NSSS), in order to study the behavior of the primary containment during the evolution of the accident scenario. In the analysis of the results of the simulation, the behavior of the wet well and dry well during the event was particularly monitored, by analyzing the evolution of temperature and pressure profiles in such volumes, this to determine the impact of the inclusion of the breaker vacuum valves and down comers. The results show that the effect of this extension of the model is that more conservative results are obtained, i.e., higher pressures are reached in both wet well and dry well than when it is used a containment model that does not include neither the vacuum valves nor the down comers. The most relevant results obtained show that the Rcic alone is able to keep the core fully covered, but even in such a case, it evaporates about 15% of the initial inventory of liquid water in the Pressure Suppression Pool (Psp). When the Rcic operation is lost, 20% more of the liquid water inventory in the Psp is further reduced within four to twelve hours (approximately), time at which the simulation crashed. Besides, there is a significant increase of pressure in the containment. As the accident evolves, the pressure in the containment continues increasing, but there is still considerable margin to reach the design pressure of the containment. At the end of the simulation, the results show a gauge pressure value of 224,550 Pa in the Psp and 187,482 Pa in the wet well

  8. Aging of steel containments and liners in nuclear power plants

    International Nuclear Information System (INIS)

    Naus, D.J.; Oland, C.B.; Ellingwood, B.; Norris, W.E.

    1998-02-01

    Aging of the containment pressure boundary in light water reactor plants is being addressed to understand the significant factors relating occurrence of corrosion efficacy of inspection and structural capacity reduction of steel containments and liners of concrete containments. and to make recommendations on use of risk models in regulatory decisions. Current regulatory in-service inspection requirements are reviewed and a summary of containment related degradation experience is presented. Current and emerging nondestructive examination techniques and a degradation assessment methodology for characterizing and quantifying the amount of damage present are described. Quantitative tools for condition assessment of aging structures using time dependent structural reliability analysis methods are summarized. Such methods provide a framework for addressing the uncertainties attendant to aging in the decision process. Results of this research provide a means for establishing current and estimating future structural capacity margins of containments, and to address the significance of incidences of reported containment degradation

  9. Primary and secondary structure of U8 small nuclear RNA

    International Nuclear Information System (INIS)

    Reddy, R.; Henning, D.; Busch, H.

    1985-01-01

    U8 small nuclear RNA is a new, capped, 140 nucleotides long RNA species found in Novikoff hepatoma cells. Its sequence is: m3GpppAmUmCGUCAGGA GGUUAAUCCU UACCUGUCCC UCCUUUCGGA GGGCAGAUAG AAAAUGAUGA UUGGAGCUUG CAUGAUCUGC UGAUUAUAGC AUUUCCGUGU AAUCAGGACC UGACAACAUC CUGAUUGCUU CUAUCUGAUUOH. This RNA is present in approximately 25,000 copies/cell, and it is enriched in nucleolar preparations. Like U1, U2, U4/U6, and U5 RNAs, U8 RNA was also present as a ribonucleoprotein associated with the Sm antigen. The rat U8 RNA was highly homologous (greater than 90%) to a recently characterized 5.4 S RNA from mouse cells infected with spleen focus-forming virus. In addition to the U8 RNA, three other U small nuclear RNAs were found in anti-Sm antibody immunoprecipitates from labeled rat and HeLa cells. Each of these contained a m3GpppAm cap structure; their apparent chain lengths were 60, 130, and 65 nucleotides. These U small nuclear RNAs are designated U7, U9, and U10 RNAs, respectively

  10. Gas flow in and out of a nuclear waste container

    International Nuclear Information System (INIS)

    Zwahlen, E.D.; Pigford, T.H.; Chambre, P.L.; Lee, W.W.L.

    1989-05-01

    We analyze the flow of gases out of and into a high-level-waste container in the unsaturated tuff of Yucca Mountain. Containers are expected to fail eventually by localized cracks and penetrations. Even though the penetrations may be small, argon gas initially in the hot container can leak out. As the waste package cools, the pressure inside the container can become less than atmospheric, and air can leak in. 14 C released from the hot fuel-cladding surface can leak out of penetrations, and air inleakage can mobilize additional 14 C and other volatile radioactive species as it oxidizes the fuel cladding and the spent fuel. In an earlier paper we studied the gas flow through container penetrations occurring at the time of emplacement. Here we analyze the flow of gas for various penetration sizes occurring at 300 years. 3 refs., 2 figs

  11. The ''nuclear car wash'': a scanner to detect illicit special nuclear material in cargo containers

    International Nuclear Information System (INIS)

    Slaughter, D. R.; Accatino, M. R.; Bernstein, A.; Dougan, A. D.; Hall, J. M.; Loshak, A.; Manatt, D. R.; Pohl, B. A.; Prussin, S. G.; Walling, R. S.; Weirup, D. L.

    2004-01-01

    There is an urgent need to improve the reliability of screening cargo containers for illicit nuclear material that may be hidden there for terrorist purposes. A screening system is described for detection of fissionable material hidden in maritime cargo containers. The system makes use of a low intensity neutron beam for producing fission; and the detection of the abundant high-energy γ rays emitted in the β-decay of short-lived fission products and β-delayed neutrons. The abundance of the delayed γ rays is almost an order of magnitude larger than that of the delayed neutrons normally used to detect fission and they are emitted on about the same time scale as the delayed neutrons, i.e., ∼1 min. The energy and temporal distributions of the delayed γ rays provide a unique signature of fission. Because of their high energy, these delayed γ rays penetrate loW--Z cargoes much more readily than the delayed neutrons. Coupled with their higher abundance, the signal from the delayed γ rays escaping from the container is predicted to be as much as six decades more intense than the delayed neutron signal, depending upon the type and thickness of the intervening cargo. The γ rays are detected in a large array of scintillators located along the sides of the container as it is moved through them. Measurements have confirmed the signal strength in somewhat idealized experiments and have also identified one interference when 14.5 MeV neutrons from the D, T reaction are used for the interrogation. The interference can be removed easily by the appropriate choice of the neutron source

  12. Tendency of nuclear pumps for PWR primary system

    International Nuclear Information System (INIS)

    Shibata, Takeshi

    1976-01-01

    At present, large PWR power stations of more than 1,000 MW are successively constructed, and the pumps used there have become large. The progress and tendency of the technical development of main pumps in primary system are described. The increase of the capacity of power stations is accomplished by increasing the circulating coolant quantity per loop or the number of loops. Same standard primary coolant pumps are employed in the plants from 500 to 1,100 MW. The type of primary coolant pumps changed from canned type to shaft seal type, and the advantages of the shaft seal type are cheap production cost, high efficiency, and the easy utilization of inertia force. The bearings and shaft seals are thermally insulated from primary coolant. As for auxiliary pumps, reciprocating filling-up pumps and centrifugal high pressure injection pumps are used for 500 MW plants, but only centrifugal pumps are used for both purposes in 800 MW plants, and in 1,100 MW plants, the pumps of both types for separate purposes and centrifugal pumps for combined purposes are installed. Horizontal or vertical pumps of same type are used as containment vessel-spraying pumps and excess heat-eliminating pumps. The type of boric acid pumps changed from canned type to mechanical seal type. (Kako, I.)

  13. Requirements for containment system components in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1988-02-01

    This Standard specifies the requirements and establishes the rules for design, fabrication, and installation of pressure-retaining containment system components. In this Standard the term 'components' includes non registered items

  14. Multi-barrier containment for nuclear power plants

    International Nuclear Information System (INIS)

    Koehler, A.C.; Leppke, D.M.; Corcoran, R.W.

    1968-01-01

    The multi-barrier containment concept and the basic structural design citeria for its adaptation to PWR systems are described. Some of the significant factors that led to the acceptance of the concept are also discussed

  15. Requirements for containment system components in CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-02-01

    This Standard specifies the requirements and establishes the rules for design, fabrication, and installation of pressure-retaining containment system components. In this Standard the term `components` includes non registered items.

  16. Introduction to the controlled nuclear fusion (magnetic containment systems)

    International Nuclear Information System (INIS)

    Cabrera, J.A.; Guasp, J.; Martin, R.

    1975-01-01

    The magnetic containment systems, their more important features, and their potentiality to became thermonuclear reactors is described. The work is based upon the first part of a set of lectures dedicated to Plasma and Fusion Physics. (author)

  17. Locks at reactor safety containments of nuclear power plants

    International Nuclear Information System (INIS)

    1978-12-01

    The rule is to be applied for material access locks. A material access lock as defined by this rule is a pressure resistant and technically gas-tight hollow body that is connected with the containment and has got two doors, the inner door of which connects the lock chamber with the interior of the containment and the outer door connects the lock chamber with the outside. It serves exclusively for transferring material or objects. (orig./HP) [de

  18. Airlocks in the containment of nuclear power plants

    International Nuclear Information System (INIS)

    1978-01-01

    An equipment hatch according to this regulator is a pressure-resisting and technically gas-tight hollow body which is connected with the containment and has got two doors, the inner door connecting the inner space of the hatch with the interior of the containment and the outer door connecting the inner space of the hatch with the exterior. It is exclusively destinied for the transport of material or objects and not of persons. (orig./HP) [de

  19. Steam explosions-induced containment failure studies for Swiss nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Zuchuat, O.; Schmocker, U. [Swiss Federal Nuclear Safety Inspectorate, Villigen (Switzerland); Esmaili, H.; Khatib-Rahbar, M.

    1998-01-01

    The assessment of the consequences of both in-vessel and ex-vessel energetic fuel-coolant interaction for Beznau (a Westinghouse pressurized water reactor with a large, dry containment), Goesgen (a Siemens/KWU pressurized water reactor with a large, dry containment) and Leibstadt (a General Electric boiling water reactor-6 with a free standing steel, MARK-III containment) nuclear power plants is presented in this paper. The Conditional Containment Failure Probability of the steel containment of these Swiss nuclear power plants is determined based on different probabilistic approaches. (author)

  20. 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)

  1. Nuclear relaxation and critical fluctuations in membranes containing cholesterol

    Science.gov (United States)

    McConnell, Harden

    2009-04-01

    Nuclear resonance frequencies in bilayer membranes depend on lipid composition. Our calculations describe the combined effects of composition fluctuations and diffusion on nuclear relaxation near a miscibility critical point. Both tracer and gradient diffusion are included. The calculations involve correlation functions and a correlation length ξ =ξ0T/(T -Tc), where T -Tc is temperature above the critical temperature and ξ0 is a parameter of molecular length. Several correlation functions are examined, each of which is related in some degree to the Ising model correlation function. These correlation functions are used in the calculation of transverse deuterium relaxation rates in magic angle spinning and quadrupole echo experiments. The calculations are compared with experiments that report maxima in deuterium and proton nuclear relaxation rates at the critical temperature [Veatch et al., Proc. Nat. Acad. Sci. U.S.A. 104, 17650 (2007)]. One Ising-model-related correlation function yields a maximum 1/T2 relaxation rate at the critical temperature for both magic angle spinning and quadrupole echo experiments. The calculated rates at the critical temperature are close to the experimental rates. The rate maxima involve relatively rapid tracer diffusion in a static composition gradient over distances of up to 10-100 nm.

  2. Direct containment heating experiments in Zion Nuclear Power Plant geometry using prototypic materials

    International Nuclear Information System (INIS)

    Binder, J.L.; McUmber, L.M.; Spencer, B.W.

    1993-01-01

    Direct Containment Heating (DCH) experiments have been completed which utilize prototypic core materials. The experiments reported on here are a continuation of the Integral Effects Testing (IET) DCH program. The experiments incorporated a 1/40 scale model of the Zion Nuclear Power Plant containment structures. The model included representations of the primary system volume, RPV lower head, cavity and instrument tunnel, and the lower containment structures. The experiments were steam driven. Iron-alumina thermite with chromium was used as a core melt stimulant in the earlier IET experiments. These earlier IET experiments at Argonne National Laboratory (ANL) and Sandia National Laboratories (SNL) provided useful data on the effect of scale on DCH phenomena; however, a significant question concerns the potential experiment distortions introduced by the use of non-prototypic iron/alumina thermite. Therefore, further testing with prototypic materials has been carried out at ANL. Three tests have been completed, DCH-U1A, U1B and U2. DCH-U1A and U1B employed an inerted containment atmosphere and are counterpart to the IET-1RR test with iron/alumina thermite. DCH-U2 employed nominally the same atmosphere composition of its counterpart iron/alumina test, IET-6. All tests, with prototypic material, have produced lower peak containment pressure rises; 45, 111 and 185 kPa in U1A, U1B and U2, compared to 150 and 250 kPa IET-1RR and 6. Hydrogen production, due to metal-steam reactions, was 33% larger in U1B and U2 compared to IET-1RR and IET-6. The pressurization efficiency was consistently lower for the corium tests compared to the IET tests

  3. Thermal-hydraulic calculation and analysis for QNPP (Qinshan Nuclear Power Plant) containment

    International Nuclear Information System (INIS)

    Xie Hui; Zhou Jie; He Yingchao

    1993-01-01

    Three containment thermal-hydraulic codes CONTEMPT-LT/028, CONTEMPT-4/MOD3 and COMPARE are used to compute and analyse the Qinshan Nuclear Power Plant (QNPP) containment response under LOCA or MSLB conditions. An evaluation of the capability of containment of QNPP is given

  4. 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)

  5. Risk ranking of LANL nuclear material storage containers for repackaging prioritization.

    Science.gov (United States)

    Smith, Paul H; Jordan, Hans; Hoffman, Jenifer A; Eller, P Gary; Balkey, Simon

    2007-05-01

    Safe handling and storage of nuclear material at U.S. Department of Energy facilities relies on the use of robust containers to prevent container breaches and subsequent worker contamination and uptake. The U.S. Department of Energy has no uniform requirements for packaging and storage of nuclear materials other than those declared excess and packaged to DOE-STD-3013-2000. This report describes a methodology for prioritizing a large inventory of nuclear material containers so that the highest risk containers are repackaged first. The methodology utilizes expert judgment to assign respirable fractions and reactivity factors to accountable levels of nuclear material at Los Alamos National Laboratory. A relative risk factor is assigned to each nuclear material container based on a calculated dose to a worker due to a failed container barrier and a calculated probability of container failure based on material reactivity and container age. This risk-based methodology is being applied at LANL to repackage the highest risk materials first and, thus, accelerate the reduction of risk to nuclear material handlers.

  6. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

    International Nuclear Information System (INIS)

    Johnson, G.L.

    1988-09-01

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. One such package would store lightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97/degree/C and whether the cladding of the stored spent fuel ever exceeds 350/degree/C. Limiting the borehole to temperatures of 97/degree/C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350/degree/C cladding limit minimizes the possibility of creep-related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97/degree/C for the full 1000-yr analysis period

  7. Nuclear magnetic resonance characterization of apple juice containing enzyme preparations

    International Nuclear Information System (INIS)

    Prestes, Rosilene A.; Almeida, Denise Milleo; Barison, Andersson; Pinheiro, Luis Antonio; Wosiacki, Gilvan

    2012-01-01

    In this work, 1 H nuclear magnetic resonance ( 1 H NMR) was employed to evaluate changes in apple juice in response to the addition of Panzym Yieldmash and Ultrazym AFP-L enzymatic complexes and compare it with premium apple juice. The juice was processed at different temperatures and concentrations of enzymatic complexes. The differences in the results were attributed mainly to the enzyme concentrations, since temperature did not cause any variation. A quantitative analysis indicated that the concentration of fructose increased while the concentrations of sucrose and glucose decreased in response to increasing concentrations of the enzymatic complexes. (author)

  8. 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

  9. Containment pressure monitoring method after severe accident in nuclear power plant

    International Nuclear Information System (INIS)

    Luo Chuanjie; Zhang Shishui

    2011-01-01

    The containment atmosphere monitoring system in nuclear power plant was designed on the basis of design accident. But containment pressure will increase greatly in a severe accident, and pressure instrument in the containment can't satisfy the monitoring requirement. A new method to monitor the pressure change in the containment after a severe accident was considered, through which accident soften methods can be adopted. Under present technical condition, adding a pressure monitoring channel out of containment for post-severe accident is a considerable method. Daya Bay Nuclear Power Plant implemented this modification, by which the containment release time can be delayed during severe accident, and nuclear safety can be increased. After analysis, this method is safe and feasible. (authors)

  10. Nuclear-powered icebreaking container ship via the Northern Sea Route in its economical aspects

    International Nuclear Information System (INIS)

    Kondo, Koichi; Takamasa, Tomoji

    1996-01-01

    In recent years, major maritime nations such as Japan, Russia and Norway have been investigating the use of the Northern Sea Route (NSR) as a sea route between the Far East and Europe, linking the eastern and western parts of the Eurasian continent. In this study, as part of the examination of suitable merchant ships for the NSR, we make an economic comparison between diesel container ships taking the Suez Canal route and NSR nuclear-powered icebreaking container ship carrying the Marine Reactor X (MRX), which is currently being developed at the Japan Atomic Energy Research Institute. Compared to diesel container ships going via the Suez Canal, the first-year transportation cost of NSR nuclear-powered container ship after commissioning is 30-70% higher and the required freight rate (RFR) is 8-40% higher. If the nuclear reactor in nuclear-powered container ship, which is the reason for higher cost, were replaced by the cassette-type MRX, the reusability of the MRX would reduce this cost difference between nuclear-powered and diesel ships. The study also shows that in terms of the total cost including sales opportunity costs, NSR nuclear-powered container ship can compete sufficiently with diesel container ships on the Suez Canal route. (author)

  11. Dynamic analysis and structural design of underground nuclear reactor containments

    International Nuclear Information System (INIS)

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

    1975-01-01

    All concept options are assumed to be similar in design criteria for structural competence to contain radioactivity and fuel heat and meet the functional, servicing, protective and aesthetic requirements. The choice of underground siting should be based on criteria developed from the sequential consideration of load-causing phenomena, concept and site characteristics. From the criteria, loads for a particular concept and site are calculated and the design formulated. (orig./ORU) [de

  12. A Hydrogen Containment Process for Nuclear Thermal Engine Ground testing

    Science.gov (United States)

    Wang, Ten-See; Stewart, Eric; Canabal, Francisco

    2016-01-01

    The objective of this study is to propose a new total hydrogen containment process to enable the testing required for NTP engine development. This H2 removal process comprises of two unit operations: an oxygen-rich burner and a shell-and-tube type of heat exchanger. This new process is demonstrated by simulation of the steady state operation of the engine firing at nominal conditions.

  13. 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.)

  14. Chemical durability of simulated nuclear glasses containing water

    International Nuclear Information System (INIS)

    Li, H.; Tomozawa, M.

    1995-04-01

    The chemical durability of simulated nuclear waste glasses having different water contents was studied. Results from the product consistency test (PCT) showed that glass dissolution increased with water content in the glass. This trend was not observed during MCC-1 testing. This difference was attributed to the differences in reactions between glass and water. In the PCT, the glass network dissolution controlled the elemental releases, and water in the glass accelerated the reaction rate. On the other hand, alkali ion exchange with hydronium played an important role in the MCC-1. For the latter, the amount of water introduced into a leached layer from ion-exchange was found to be much greater than that of initially incorporated water in the glass. Hence, the initial water content has no effect on glass dissolution as measured by the MCC-1 test

  15. Hermetic cable penetrations for containments of nuclear power reactors meet high safety standards

    International Nuclear Information System (INIS)

    Kusserow, J.; Gurr, W.; Pflug, H.

    1985-05-01

    Different types of cable penetrations for containments of nuclear power reactors have been developed and fabricated in the GDR. The technical parameters achieved are in accordance with the radiation protection requirements

  16. 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)

  17. Experiments to investigate direct containment heating phenomena with scaled models of the Surry Nuclear Power Plant

    International Nuclear Information System (INIS)

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

    1994-01-01

    The Containment Technology Test Facility (CTTF) and the Surtsey Test Facility at Sandia National Laboratories (SNL) are used to perform scaled experiments for the Nuclear Regulatory Commission (NRC) that simulate High Pressure Melt Ejection (HPME) accidents in a nuclear power plant (NPP). These experiments are designed to investigate the effects of direct containment heating (DCH) phenomena on the containment load. High-temperature, chemically reactive melt is ejected by high-pressure steam into a scale model of a reactor cavity. Debris is entrained by the steam blowdown into a containment model where specific phenomena, such as the effect of subcompartment structures, prototypic atmospheres, and hydrogen generation and combustion, can be studied

  18. Fabrication development for high-level nuclear waste containers for the tuff repository

    International Nuclear Information System (INIS)

    Domian, H.A.; Holbrook, R.L.; LaCount, D.F.; Babcock and Wilcox Co., Alliance, OH

    1990-09-01

    This final report completes Phase 1 of an engineering study of potential manufacturing processes for the fabrication of containers for the long-term storage of nuclear waste. An extensive literature and industry review was conducted to identify and characterize various processes. A technical specification was prepared using the American Society of Mechanical Engineers Boiler ampersand Pressure Vessel Code (ASME BPVC) to develop the requirements. A complex weighting and evaluation system was devised as a preliminary method to assess the processes. The system takes into account the likelihood and severity of each possible failure mechanism in service and the effects of various processes on the microstructural features. It is concluded that an integral, seamless lower unit of the container made by back extrusion has potential performance advantages but is also very high in cost. A welded construction offers lower cost and may be adequate for the application. Recommendations are made for the processes to be further evaluated in the next phase when mock-up trials will be conducted to address key concerns with various processes and materials before selecting a primary manufacturing process. 43 refs., 26 figs., 34 tabs

  19. Analysis to the criticality the storage and containers to the Juragua Nuclear Power Plant Fuel

    International Nuclear Information System (INIS)

    Guerra Valdes, R.

    1998-01-01

    Presently analysis the criticality the warehouses and containers the nuclear fuels in Juragua nuclear power plant the property multiplicity determined in these system and it is verified that for the geometry and operation conditions defined in the design as well as in accidents situations, the arrangement the fuel stays subcritical with an appropriate margin

  20. Long-term reliability evaluation of nuclear containments with tendon force degradation

    International Nuclear Information System (INIS)

    Kim, Sang-Hyo; Choi, Moon-Seock; Joung, Jung-Yeun; Kim, Kun-Soo

    2013-01-01

    Highlights: • A probabilistic model on long-term degradation of tendon force is developed. • By using the model, we performed reliability evaluation of nuclear containment. • The analysis is also performed for the case with the strict maintenance programme. • We showed how to satisfy the target safety in the containments facing life extension. - Abstract: The long-term reliability of nuclear containment is important for operating nuclear power plants. In particular, long-term reliability should be clarified when the service life of nuclear containment is being extended. This study focuses not only on determining the reliability of nuclear containment but also presenting the reliability improvement by strengthening the containment itself or by running a strict maintenance programme. The degradation characteristics of tendon force are estimated from the data recorded during in-service inspection of containments. A reliability analysis is conducted for a limit state of through-wall cracking, which is conservative, but most crucial limit state. The results of this analysis indicate that reliability is the lowest at 3/4 height of the containment wall. Therefore, this location is the most vulnerable for the specific limit state considered in this analysis. Furthermore, changes in structural reliability owing to an increase in the number of inspecting tendons are analysed for verifying the effect of the maintenance program's intensity on expected containment reliability. In the last part of this study, an example of obtaining target reliability of nuclear containment by strengthening its structural resistance is presented. A case study is conducted for exemplifying the effect of strengthening work on containment reliability, especially during extended service life

  1. 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.)

  2. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

    International Nuclear Information System (INIS)

    Johnson, G.L.

    1991-11-01

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. On such package would store tightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97 degrees C and whether the cladding of the stored spent fuel ever exceeds 350 degrees C. Limiting the borehole to temperatures of 97 degrees C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350 degrees C cladding limit minimizes the possibility of creep- related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97 degrees C for the full 10000-yr analysis period. For the 4.74-kW load, the peak cladding temperature rises to just below the 350 degrees C limit about 4 years after emplacement. If the packages are stored using the spacing specified in the Site Characterization Plan (15 ft x 126 ft), a maximum of 4.1 kW per container may be stored. If the 0.05-m-thick void between the container and the borehole wall is filled with loosely packed bentonite, the peak cladding temperature rises more than 40 degrees C above the allowed cladding limit. In all cases the dominant heat transfer mode between container components is thermal radiation

  3. Heat removing device for nuclear reactor container facility

    Energy Technology Data Exchange (ETDEWEB)

    Tateno, Seiya; Tominaga, Kenji; Iwata, Yasutaka; Kinoshita, Shoichiro; Niino, Tsuyoshi

    1994-09-30

    A pressure suppression chamber incorporating pool water is disposed inside of a reactor container for condensating steams released to a dry well upon occurrence of abnormality. A pool is disposed at the outer circumference of the pressure suppression chamber having a steel wall surface of the reactor container as a partition wall. The outer circumferential pool is in communication with ocean by way of a lower communication pipeline and an upper communication pipeline. During normal plant operation state, partitioning valves disposed respectively to the upper and lower communication pipelines are closed, so that the outer circumferential pool is kept empty. After occurrence loss of coolant accident, steams generated by after-heat of the reactor core are condensated by pool water of the pressure suppression chamber, and the temperature of water in the pressure suppression chamber is gradually elevated. During the process, the partition valves of the upper and lower communication pipelines are opened to introduce cold seawater to the outer circumferential pool. With such procedures, heat of the outer circumferential pool is released to the sea by natural convection of seawater, thereby enabling to remove residual heat without dynamic equipments. (I.N.).

  4. 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)

  5. Non-destructive evaluation of containment walls in nuclear power plants

    Science.gov (United States)

    Garnier, V.; Payan, C.; Lott, M.; Ranaivomanana, N.; Balayssac, J. P.; Verdier, J.; Larose, E.; Zhang, Y.; Saliba, J.; Boniface, A.; Sbartai, Z. M.; Piwakowski, B.; Ciccarone, C.; Hafid, H.; Henault, J. M.; Buffet, F. Ouvrier

    2017-02-01

    Two functions are regularly tested on containment walls in order to anticipate a possible accident. The first is mechanical to resist a possible internal over-pressure and the second is to prevent leakage. The AAPR reference accident is the rupture of a pipe in the primary circuit of a nuclear plant. In this case, the pressure and temperature can reach 5 bar and 180°C in 20 seconds. The national project `Non-destructive testing of the containment structures of nuclear plants' aims at studying the non-destructive techniques capable to evaluate the concrete properties and its damaging and cracks. This 4-year-project is segmented into two parts. The first consists in developing and selecting the most relevant NDEs in the laboratory to reach these goals. These evaluations are developed in conditions representing the real conditions of the stresses generated during ten-yearly visits of the plants or those related to an accident. The second part consists in applying the selected techniques to two containment structures under pressure. The first structure is proposed by ONERA and the second is a mockup of a containment wall on a 1/3 scale made by EDF within the VeRCoRs project. Communication is focused on the part of the project that concerns the damage and crack process characterization by means of NDT. The tests are done in 3 or 4 points bending in order to study the cracks' generation, their propagation, as well as their opening and closing. The main ultrasonic techniques developed concern linear or non-linear acoustic: acoustic emission [1], Locadiff [2], energy diffusion, surface wave's velocity and attenuation, DAET [3]. The recorded data contribute to providing the mapping of the investigated parameters, either in volume, in surface or globally. Digital image correlation is an important additional asset to validate the coherence of the data. The spatial normalization of the data in the specimen space allows proposing algorithms on the combination of the

  6. Thirty years of measured prestress at Swedish nuclear reactor containments

    International Nuclear Information System (INIS)

    Anderson, Patrick

    2005-01-01

    The main function of the reactor containment, i.e. to ensure tightness at a major internal accident, depends directly on the prestressing system. To secure that the prestress level is sufficient, the tendon force has been measured during the whole time of operation. The general results from these measurements show that the loss of prestress 30 years after tensioning is between 5 and 10%. This is much lower loss than predicted initially at the design stage. More advanced and today commonly used models for predicting prestress loss show better agreement with the results. The main reasons for the relatively low loss are assumed to be: (1) the confirmed slow drying process of the concrete and (2) the high concrete age at the initial tensioning. The results also indicate that the temperature has a major influence on the loss of prestress

  7. Design of reactor containment systems for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

  8. Design of reactor containment systems for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

  9. Mitigating Capability Analysis during LOCA for Korean Standard Nuclear Power Plants in Containment Integrity

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Young; Park, Soo Yong; Kim, D. H.; Song, Y. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-07-01

    The objective of this paper is to establish Containment spray operational technical bases for the typical Korean Standard Nuclear Power plants (Ulchin units 3 and 4) by modeling the plant, and analyzing a loss of coolant accident (LOCA) using the MAAP code. The severe accident phenomena at nuclear power plants have large uncertainties. For the integrity of the reactor vessel and containment safety against severe accidents, it is essential to understand severe accident sequences and to assess the accident progression accurately by computer codes. Furthermore, it is important to attain the capability to analyze a advanced nuclear reactor design for a severe accident prevention and mitigation.

  10. Probabilistic analysis of the efficiency of the damping devices against nuclear fuel container falling

    Science.gov (United States)

    Králik, Juraj

    2017-07-01

    The paper presents the probabilistic and sensitivity analysis of the efficiency of the damping devices cover of nuclear power plant under impact of the container of nuclear fuel of type TK C30 drop. The finite element idealization of nuclear power plant structure is used in space. The steel pipe damper system is proposed for dissipation of the kinetic energy of the container free fall. The experimental results of the shock-damper basic element behavior under impact loads are presented. The Newmark integration method is used for solution of the dynamic equations. The sensitivity and probabilistic analysis of damping devices was realized in the AntHILL and ANSYS software.

  11. Prediction of seismic motion from contained and excavation nuclear detonations

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, R A [Environmental Research Corp., Alexandria, VA (United States)

    1970-05-15

    Capability to predict ground motions from nuclear events is developed on empirical and theoretical bases. Analyses of the experimental data provide basic predictions of peak particle motions and spectra which follow a (yield){sup m} times (distance){sup -n} relationship. The exponents on yield and distance are frequency dependent and derived from experiment and theory. Theory provides a physical understanding of the phenomena which allows extrapolation to off-NTS and atypical events. For example, yield scaling theory predicts significantly higher frequency motions and consequently larger ground accelerations for overburied events such as Gasbuggy, Rulison, Wasp and Wagon Wheel. These conclusions are observed from Gasbuggy (26 kt) which generated ground accelerations comparable to a normal buried event of 200 kt. This result is important in avoiding personal injury and assessing the probability of property damage. Conversely, theory predicts lower ground accelerations and seismic efficiencies for excavation events; these effects are observed from the Cabriolet and Schooner events and consequently predicted for the Sturtevant and Yawl events. With regard to the distance exponent, scattering theory determines a distance exponent which predicts greater attenuation effects on higher frequency motions. This trend is verified experimentally by regression analyses on a large number of data points which determine the distance exponent to range from -1.1 at low frequencies to -1.6 at high frequencies. Results indicate that cube root similarity scaling is not appropriate in the far field except possibly for peak particle displacements at the low frequency end of the spectrum. In addition to the source and transmission factors, current ground motion prediction techniques, on and off-NTS, take into account local site characteristics. Experimental evidence and theoretical models--layered media elastic theory, finite element modeling, and building response modeling

  12. Synthesis of studies on primary containers for MLA-VL wastes

    International Nuclear Information System (INIS)

    Bart, F.; Delassale, F.; Rey, F.; Helie, M.; Levoy, R.; Moitrier, C.; Sicardy, O.; Tiquet, P.

    2004-01-01

    The aim of this study is the presentation of studies realized on primary containers of medium activity long life level. These studies are realized in the framework of the axis 3 of the law of 1991 on the radioactive waste management. The specificity of this document is the presentation of container for ''random'' wastes chemically corrosive in order to complete the range of possible packages. Thus a special program has been developed to demonstrate a conditioning solution which offers to the waste producers a possibility of conditioning these wastes without a preliminary treatment. (A.L.B.)

  13. ORNL experiments to characterize fuel release from the reactor primary containment in severe LMFBR accidents

    International Nuclear Information System (INIS)

    Wright, A.L.; Kress, T.S.; Smith, A.M.

    1980-01-01

    This paper presents results from aerosol source term experiments performed in the ORNL Aerosol Release and Transport (ART) Program sponsored by the US NRC. The tests described were performed to provide information on fuel release from an LMFBR primary containment as a result of a hypothetical core-disruptive accident (HCDA). The release path investigated in these tests assumes that a fuel/sodium bubble is formed after disassembly that transports fuel and fission products through the sodium coolant and cover gas to be relased into the reactor secondary containment. Due to the excellent heat transfer characteristics of the sodium, there is potential for large attenuation of the maximum release

  14. Hydrogen deflagration simulations under typical containment conditions for nuclear safety

    Energy Technology Data Exchange (ETDEWEB)

    Yanez, J., E-mail: jorge.yanez@kit.edu [Institute for Energy and Nuclear Technology, Karlsruhe Institute of Technology, Kaiserstrasse 12, 76131 Karlsruhe (Germany); Kotchourko, A.; Lelyakin, A. [Institute for Energy and Nuclear Technology, Karlsruhe Institute of Technology, Kaiserstrasse 12, 76131 Karlsruhe (Germany)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Lean H{sub 2}-air combustion experiments highly relevant to typical NPP simulated. Black-Right-Pointing-Pointer Analyzed effect of temperature, concentration of H{sub 2}, and steam concentration. Black-Right-Pointing-Pointer Similar conditions and H{sub 2} concentration yielded different combustion regimes. Black-Right-Pointing-Pointer Flame instabilities (FIs) were the effect driving divergences. Black-Right-Pointing-Pointer Model developed for acoustic FI in simulations. Agreement experiments obtained. - Abstract: This paper presents the modeling of low-concentration hydrogen deflagrations performed with the recently developed KYLCOM model specially created to perform calculations in large scale domains. Three experiments carried out in THAI facility (performed in the frames of international OECD THAI experimental program) were selected to be analyzed. The tests allow studying lean mixture hydrogen combustion at normal ambient, elevated temperature and superheated and saturated conditions. The experimental conditions considered together with the facility size and shape grant a high relevance degree to the typical NPP containment conditions. The results of the simulations were thoroughly compared with the experimental data, and the comparison was supplemented by the analysis of the combustion regimes taking place in the considered tests. Results of the analysis demonstrated that despite the comparatively small difference in mixture properties, three different combustion regimes can be definitely identified. The simulations of one of the cases required of the modeling of the acoustic-parametric instability which was carefully undertaken.

  15. A study on the hydrogen distributions in a containment for nuclear plant severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kweon Ha; Kim, Ju Youn; Bae, Kyung Hyo [The Korea Maritime Univ., Busan (Korea, Republic of)

    2012-10-15

    Hydrogen explosion has been considered as one of the major issues since Fukushima nuclear accident. The cause of the explosion has not been discovered, but it is clear that the explosion strongly depends on hydrogen distributions in a containment. In this study hydrogen distributions are calculated and analyzed in the containment of APR 1400(Advanced Power Reactor 1400)

  16. Modelling and simulation of containment on full scope simulator for Qinshan 300 MW Nuclear Power Unit

    International Nuclear Information System (INIS)

    Zou Tingyun

    1996-01-01

    A multi-node containment thermal-hydraulic model has been developed and adapted in Full Scope Simulator for Qinshan 300 MW Nuclear Power Unit with good realtime simulation effects. Containment pressure for LBLOCA calculated by the model is well agreed with those of CONTEMPT-4/MOD3

  17. 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

  18. Define rules for the exporter and importer of minerals or ores containing nuclear elements

    International Nuclear Information System (INIS)

    1969-01-01

    The present resolution establishes regulations for the exporter of minerals or ores containing associated nuclear elements, and for the importer of chemical compounds of technical purity grade, containing a quantity of fissile of fertile materials equal to the existent in the exported material

  19. Use of the radiochemical method for identification of seized nuclear material - container

    International Nuclear Information System (INIS)

    Rosskopfova, O.; Matel, L.; Rajec, P.; Dobias, M.

    2003-01-01

    Analysis of the nuclear material of the seized illicit trafficking container was performed in the laboratory LARCHA. The container was professionally taken apart in the hot cell of HUMA-LAB APEKO. The compact core of container with mass of 36.00 kg was separated and analysed in the LARCHA laboratory. The results of alpha spectrometry proved that the core of container was made of depleted uranium. A suspicious container was seized, that was believed to be used for transportation of radioactive and nuclear materials. Container, according to Slovak Law, was transported to a Civil defence and police started with expertise analysis. On the basis of previous results, it was assumed that the core of the container was made of a nuclear material. The container was cut inside a hot cell to separate parts at HUMA-Lab Kosice and individual parts were prepared for analyses. From shape and type of the detained container it was assumed that the core of the container was made of depleted uranium and suspected cut pieces were sent to the accredited laboratory LARCHA Bratislava for isotope analysis. Samples from different parts of the container were analysed by using HPGe gamma. The results of gamma spectrometry proved the presence of Co-60, Cs-137 and U-235 with values 11.8 Bq, 5.6 Bq a 3.8 Bq respectively. On the basis of measured results it was calculated mass content (%) of uranium isotopes in samples. It was concluded that: - container is composed of different parts, which are made from steel, lead and uranium; - the wipe tests proved the presence of radionuclides of Cs-137 and Co-60; - the core of the container is made of depleted uranium; - chosen methods, which were used proved that are suitable for analysis and identification of nuclear materials. (authors)

  20. Reverse osmosis for the recovery of boric acid from the primary coolant at nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Bártová, Šárka, E-mail: sarka.bartova@cvrez.cz [Research Centre Řež Ltd., Husinec-Řež 130, 250 68 Řež (Czech Republic); Kůs, Pavel [Research Centre Řež Ltd., Husinec-Řež 130, 250 68 Řež (Czech Republic); Skala, Martin [Research Centre Řež Ltd., Husinec-Řež 130, 250 68 Řež (Czech Republic); University of Chemical Technology, Prague, Department of Chemical Engineering, Technická 5, Prague 166 28 (Czech Republic); Vonková, Kateřina [Research Centre Řež Ltd., Husinec-Řež 130, 250 68 Řež (Czech Republic)

    2016-04-15

    Highlights: • RO membranes tested for boric acid recovery from primary coolant of nuclear power plants. • Scanning electron microscopy was used for the characterization of the membranes. • Lab scale experiments performed under various operation conditions. • We proposed configuration of and operation conditions for RO unit in nuclear power plant. - Abstract: At nuclear power plants (NPP), evaporators are used for the treatment of primary coolant and other liquid radioactive waste containing H{sub 3}BO{sub 3}. Because the operation of evaporators is expensive, a number of more cost-effective alternatives has been considered, one of which is reverse osmosis. We tested reverse osmosis modules from several manufactures on a batch laboratory apparatus. SEM images of the tested membranes were taken to distinguish the differences between the membranes. Water permeability through membranes was evaluated from the experiments with pure water. The experiments were performed with feed solutions containing various concentrations of H{sub 3}BO{sub 3} in a range commonly occurring in radioactive waste. The pH of the feed solutions ranged from 5.2 to 11.2. Our results confirmed that the pH of the feed solution plays the most important role in membrane separation efficiency of H{sub 3}BO{sub 3}. Certain modifications to the pH of the feed solution were needed to enable the tested membranes to concentrate the H{sub 3}BO{sub 3} in the retentate stream, separate from the pure water in the permeate stream. On this basis, we propose the configuration of and operational conditions for a reverse osmosis unit at NPP.

  1. The economic potential of a cassette-type-reactor-installed nuclear ice-breaking container ship

    International Nuclear Information System (INIS)

    Kondo, K.; Takamasa, T.

    2000-01-01

    The design concept of the cassette-type-reactor MRX (Marine Reactor X), being under development in Japan for the nuclear ice-breaker container ship is described. The MRX reactor is the monoblock water-cooled and moderated reactor with passive cooling system of natural circulation. It is shown that application of the reactor being under consideration gives an opportunity to decrease greatly the difference in prices for similar nuclear and diesel ships. Economic estimations for applicability of the nuclear ice-breaker container ship with the MRX reactor in Arctics for transportation of standard containers TEU from Europe to Far East as compared with transportation of the same containers by diesel ships via Suets Canal are made [ru

  2. Automated nuclear material recovery and decontamination of large steel dynamic experiment containers

    International Nuclear Information System (INIS)

    Dennison, D.K.; Gallant, D.A.; Nelson, D.C.; Stovall, L.A.; Wedman, D.E.

    1999-01-01

    A key mission of the Los Alamos National Laboratory (LANL) is to reduce the global nuclear danger through stockpile stewardship efforts that ensure the safety and reliability of nuclear weapons. In support of this mission LANL performs dynamic experiments on special nuclear materials (SNM) within large steel containers. Once these experiments are complete, these containers must be processed to recover residual SNM and to decontaminate the containers to below low level waste (LLW) disposal limits which are much less restrictive for disposal purposes than transuranic (TRU) waste limits. The purpose of this paper is to describe automation efforts being developed by LANL for improving the efficiency, increasing worker safety, and reducing worker exposure during the material cleanout and recovery activities performed on these containers

  3. Department of the Navy final environmental impact statement for a container system for the management of naval spent nuclear fuel

    International Nuclear Information System (INIS)

    1996-11-01

    This Final Environmental Impact Statement (EIS) addresses six general alternative systems for the loading, storage, transport, and possible disposal of naval spent nuclear fuel following examination. This EIS describes environmental impacts of (1) producing and implementing the container systems (including those impacts resulting from the addition of the capability to load the containers covered in this EIS in dry fuel handling facilities at Idaho National Engineering Laboratory (INEL)); (2) loading of naval spent nuclear fuel at the Expended Core Facility or at the Idaho Chemical Processing Plant with subsequent storage at INEL; (3) construction of a storage facility (such as a paved area) at alternative locations at INEL; and (4) loading of containers and their shipment to a geologic repository or to a centralized interim storage site outside the State of Idaho once one becomes available. As indicated in the EIS, the systems and facilities might also be used for handling low-level radiological waste categorized as special case waste. The Navy's preferred alternative for a container system for the management of naval spent fuel is a dual-purpose canister system. The primary benefits of a dual-purpose canister system are efficiencies in container manufacturing and fuel reloading operations, and potential reductions in radiation exposure

  4. Containers for short-term storage of nuclear materials at the Los Alamos plutonium facility

    International Nuclear Information System (INIS)

    Hagan, R.; Gladson, J.

    1997-01-01

    The Los Alamos Plutonium Facility for the past 18 yr has stored nuclear samples for archiving and in support of nuclear materials research and processing programs. In the past several years, a small number of storage containers have been found in a deteriorated condition. A failed plutonium container can cause personnel contamination exposure and expensive physical area decontamination. Containers are stored in a physically secure radiation area vault, making close inspection costly in the form of personnel radiation exposure and work time. A moderate number of these containers are used in support of plutonium processing and must withstand daily handling abuse. A 2-yr evaluation of failed containers and those that have shown no deterioration has been conducted. Based on that study, a program was established to formalize our packing methods and materials and standardize the size and shape of containers that are used for short-term use. A standardized set of containers was designed, evaluated, tested, and procured for use in the facility. This paper reviews our vault storage problems, shows some failed containers, and presents our planned solutions to provide safe and secure containment of nuclear materials

  5. Stability of live attenuated rotavirus vaccine with selected preservatives and primary containers.

    Science.gov (United States)

    Lal, Manjari; Jarrahian, Courtney; Zhu, Changcheng; Hosken, Nancy A; McClurkan, Chris L; Koelle, David M; Saxon, Eugene; Roehrig, Andrew; Zehrung, Darin; Chen, Dexiang

    2016-05-11

    Rotavirus infection, which can be prevented by vaccination, is responsible for a high burden of acute gastroenteritis disease in children, especially in low-income countries. An appropriate formulation, packaging, and delivery device for oral rotavirus vaccine has the potential to reduce the manufacturing cost of the vaccine and the logistical impact associated with introduction of a new vaccine, simplify the vaccination procedure, and ensure that the vaccine is safely and accurately delivered to children. Single-dose prefilled presentations can be easy to use; however, they are typically more expensive, can be a bottleneck during production, and occupy a greater volume per dose vis-à-vis supply chain storage and medical waste disposal, which is a challenge in low-resource settings. Multi-dose presentations used thus far have other issues, including increased wastage of vaccine and the need for separate delivery devices. In this study, the goals were to evaluate both the technical feasibility of using preservatives to develop a liquid multi-dose formulation and the primary packaging alternatives for orally delivered, liquid rotavirus vaccines. The feasibility evaluation included evaluation of commonly used preservatives for compatibility with rotavirus vaccines and stability testing of rotavirus vaccine in various primary containers, including Lameplast's plastic tubes, BD's oral dispenser version of Uniject™ (Uniject DP), rommelag's blow-fill-seal containers, and MEDInstill's multi-dose vial and pouch. These presentations were compared to a standard glass vial. The results showed that none of the preservatives tested were compatible with a live attenuated rotavirus vaccine because they had a detrimental effect on the viability of the virus. In the presence of preservatives, vaccine virus titers declined to undetectable levels within 1 month. The vaccine formulation without preservatives maintained a stability profile over 12 months in all primary containers

  6. Nanoparticles for Protein Sensing in Primary Containers: Interaction Analysis and Application.

    Science.gov (United States)

    Pérez Medina Martínez, Víctor; Espinosa-de la Garza, Carlos E; Méndez-Silva, Diego A; Bolívar-Vichido, Mariana; Flores-Ortiz, Luis F; Pérez, Néstor O

    2018-05-01

    Silver nanoparticles (AgNPs) are known to interact with proteins, leading to modifications of the plasmonic absorption that can be used to monitor this interaction, entailing a promising application for sensing adsorption of therapeutic proteins in primary containers. First, transmission electron microscopy in combination with plasmonic absorption and light scattering responses were used to characterize AgNPs and protein-AgNP complexes, including its concentration dependence, using two therapeutic molecules as models: a monoclonal antibody (mAb) and a synthetic copolymer (SC). Upon interaction, a protein corona was formed around AgNPs with the consequent shifting and broadening of their characteristic surface plasmon resonance (SPR) band (400 nm) to 410 nm and longer wavelenghts. Additional studies revealed secondary and three-dimensional structure modifications of model proteins upon interaction with AgNPs by circular dichroism and fluorescence techniques, respectively. Based on the modification of the SPR condition of AgNPs upon interaction with proteins, we developed a novel protein-sensing application of AgNPs in primary containers. This strategy was used to conduct a compatibility assessment of model proteins towards five commercially available prefillable glass syringe (PFS) models. mAb- and SC-exposed PFSs showed that 74 and 94% of cases were positive for protein adsorption, respectively. Interestingly, protein adsorption on 15% of total tested PFSs was negligible (below the nanogram level). Our results highlight the need of a case-by-case compatibility assessment of therapeutic proteins and their primary containers. This strategy has the potential to be easily applied on other containers and implemented during early-stage product development by pharmaceutical companies and for routine use during batch release by packaging manufacturers.

  7. Torsin Mediates Primary Envelopment of Large Ribonucleoprotein Granules at the Nuclear Envelope

    Directory of Open Access Journals (Sweden)

    Vahbiz Jokhi

    2013-04-01

    Full Text Available A previously unrecognized mechanism through which large ribonucleoprotein (megaRNP granules exit the nucleus is by budding through the nuclear envelope (NE. This mechanism is akin to the nuclear egress of herpes-type viruses and is essential for proper synapse development. However, the molecular machinery required to remodel the NE during this process is unknown. Here, we identify Torsin, an AAA-ATPase that in humans is linked to dystonia, as a major mediator of primary megaRNP envelopment during NE budding. In torsin mutants, megaRNPs accumulate within the perinuclear space, and the messenger RNAs contained within fail to reach synaptic sites, preventing normal synaptic protein synthesis and thus proper synaptic bouton development. These studies begin to establish the cellular machinery underlying the exit of megaRNPs via budding, offer an explanation for the “nuclear blebbing” phenotype found in dystonia models, and provide an important link between Torsin and the synaptic phenotypes observed in dystonia.

  8. Containment response analysis for the PSA (Probabilistic Safety Assessment) of the CAREM-25 nuclear power plant

    International Nuclear Information System (INIS)

    Baron, J.H.

    1997-01-01

    This work is part of the probabilistic safety assessment actually under development for the CAREM-25 nuclear power station, and departs from the accident sequences already obtained and quantified by the Event Trees/Fault Trees techniques. At first, the potential containment failure modes for nuclear stations are listed, based on the experience. Then, the CAREM-25 design peculiarities are analyzed, on their possible influence on the containment behavior during severe accidents. Then plan damage states are defined. Furthermore, containment damage states are also defined, and containment event trees are built for each plant damage state. Those sequences considered representative from the annual probability (those which exceed or probability of IE-09 per year, are used to quantify the combinations of plant damage states/containment damage states, based on the estimation of a vulnerability matrix. (author) [es

  9. Risks attached to container- and bunker-storage of nuclear waste

    International Nuclear Information System (INIS)

    Jager, D. de

    1987-12-01

    The results are presented of a literature study into the risks attached to the two dry-storage options selected by the Dutch Central Organization For Radioactive Waste (COVRA): the container- and the bunker-storage for irradiated nuclear-fuel elements and nuclear waste. Since the COVRA does not make it clear how these concepts should have to be realized, the experiences abroad with dry interim-storage are considered. In particular the Castor-container-storage and the bunker storage proposed in the committee MINSK (Possibilities of Interim-storage in the Netherlands of Irradiated nuclear-fuel elements and Nuclear waste) are studied further in depth. The committee MINSK has performed a study into the technical realizability of various interim-storage facilities, among which a storage in bunkers. (author). 75 refs.; 14 figs.; 16 tabs

  10. Development of deterioration models and tests of structural materials for nuclear containment structures(III)

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Byung Hwan [Seoul National University, Seoul (Korea)

    2002-03-01

    The nuclear containment structures are very important infrastructures which require much cost for construction and maintenance. If these structures lose their functions and do not ensure their safety, great losses of human lives and properties will result. Therefore, the nuclear containment structures should secure appropriate safety and functions during these service lives. The nuclear concrete structures start to experience deterioration due to severe environmental condition, even though the concrete structures exhibit generally superior durability. It is, therefore, necessary to take appropriate actions at each stage of planning, design and construction to secure safety and functionability. Thorough examination of deterioration mechanism and comprehensive tests have been conducted to explore the durability characteristics of nuclear concrete structures. 88 refs., 70 figs., 12 tabs. (Author)

  11. Pyrometallurgical separation processes of radionuclides contained in the irradiated nuclear fuel

    International Nuclear Information System (INIS)

    De Cordoba, Guadalupe; Caravaca, Concha; Quinones, Javier; Gonzalez de la Huebra, Angel

    2005-01-01

    Faced with the new options for the high level waste management, the ''Partitioning and Transmutation (P and T)'' of the radio nuclides contained in the irradiated nuclear fuel appear as a promising option from different points of view, such as environmental risk, radiotoxic inventory reduction, economic, etc.. The present work is part of a research project called ''PYROREP'' of the 5th FWP of the EU that studied the feasibility of the actinide separation from the rest of fission products contained in the irradiated nuclear fuel by pyrometallurgical processes with the aim of their transmutation. In order to design these processes it is necessary to determine basic thermodynamic and kinetic data of the radionuclides contained in the nuclear fuel in molten salt media. The electrochemical study of uranium, samarium and molybdenum in the eutectic melt LiCl - KCl has been performed at a tungsten electrode in the temperature range of 450 - 600 deg C in order to obtain these basic properties. (Author)

  12. Calculations of hydrogen detonations in nuclear containments by the random choice method

    International Nuclear Information System (INIS)

    Delichatsios, M.A.; Genadry, M.B.

    1983-01-01

    Computer codes were developed for the prediction of pressure histories at different points of a nuclear containment wall due to postulated internal hydrogen detonations. These pressure histories are required to assess the structural response of a nuclear containment to hydrogen detonations. The compressible flow equations including detonation, which was treated as a sharp fluid discontinuity, were solved by the random choice method which reproduces maximum pressures and discontinuities sharply. The computer codes were validated by calculating pressure profiles and maximum wall pressures for plane and spherical geometries and comparing the results with exact analytic solutions. The two-dimensional axisymmetric program was used to calculate wall pressure histories in an actual nuclear containment. The numerical results for wall pressures are presented in a dimensionless form, which allows their use for different combinations of hydrogen concentration, and initial conditions. (orig.)

  13. Probabilistic fracture failure analysis of nuclear piping containing defects using R6 method

    International Nuclear Information System (INIS)

    Lin, Y.C.; Xie, Y.J.; Wang, X.H.

    2004-01-01

    Failure analysis of in-service nuclear piping containing defects is an important subject in the nuclear power plants. Considering the uncertainties in various internal operating loadings and external forces, including earthquake and wind, flaw sizes, material fracture toughness and flow stress, this paper presents a probabilistic assessment methodology for in-service nuclear piping containing defects, which is especially designed for programming. A general sampling computation method of the stress intensity factor (SIF), in the form of the relationship between the SIF and the axial force, bending moment and torsion, is adopted in the probabilistic assessment methodology. This relationship has been successfully used in developing the software, Safety Assessment System of In-service Pressure Piping Containing Flaws (SAPP-2003), based on a well-known engineering safety assessment procedure R6. A numerical example is given to show the application of the SAPP-2003 software. The failure probabilities of each defect and the whole piping can be obtained by this software

  14. Computational Fluid Dynamics Modeling of Steam Condensation on Nuclear Containment Wall Surfaces Based on Semiempirical Generalized Correlations

    Directory of Open Access Journals (Sweden)

    Pavan K. Sharma

    2012-01-01

    Full Text Available In water-cooled nuclear power reactors, significant quantities of steam and hydrogen could be produced within the primary containment following the postulated design basis accidents (DBA or beyond design basis accidents (BDBA. For accurate calculation of the temperature/pressure rise and hydrogen transport calculation in nuclear reactor containment due to such scenarios, wall condensation heat transfer coefficient (HTC is used. In the present work, the adaptation of a commercial CFD code with the implementation of models for steam condensation on wall surfaces in presence of noncondensable gases is explained. Steam condensation has been modeled using the empirical average HTC, which was originally developed to be used for “lumped-parameter” (volume-averaged modeling of steam condensation in the presence of noncondensable gases. The present paper suggests a generalized HTC based on curve fitting of most of the reported semiempirical condensation models, which are valid for specific wall conditions. The present methodology has been validated against limited reported experimental data from the COPAIN experimental facility. This is the first step towards the CFD-based generalized analysis procedure for condensation modeling applicable for containment wall surfaces that is being evolved further for specific wall surfaces within the multicompartment containment atmosphere.

  15. Lead iron phosphate glass as a containment medium for disposal of high-level nuclear waste

    International Nuclear Information System (INIS)

    Boatner, L.A.; Sales, B.C.

    1989-01-01

    This patent describes lead-iron phosphate glasses containing a high level of Fe 2 O 3 for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90 0 C, with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10 2 to 10 3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe 2 O 3 in forming the lead-iron phosphate glass is critical. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms

  16. Description of the CONTAIN input model for the Dodewaard nuclear power plant

    International Nuclear Information System (INIS)

    Velema, E.J.

    1992-02-01

    This report describes the ECN standard CONTAIN input model for the Dodewaard Nuclear Power Plant (NPP) that has been developed by ECN. This standard input model will serve as a basis for analyses of the phenomena which may occur inside the Dodewaard containment in the event of a postulated severe accident. Boundary conditions for specific containment analyses can easily be implemented in the input model. as a result ECN will be able to respond quickly on requests for analyses from the utilities of the authorities. The report also includes brief descriptions of the Dodewaard NPP and the CONTAIN computer program. (author). 7 refs.; 5 figs.; 3 tabs

  17. Program to develop analytical tools for environmental and safety assessment of nuclear material shipping container systems

    International Nuclear Information System (INIS)

    Butler, T.A.

    1978-11-01

    This paper describes a program for developing analytical techniques to evaluate the response of nuclear material shipping containers to severe accidents. Both lumped-mass and finite element techniques are employed to predict shipping container and shipping container-carrier response to impact. The general impact problem is computationally expensive because of its nonlinear, three-dimensional nature. This expense is minimized by using approximate models to parametrically identify critical cases before more exact analyses are performed. The computer codes developed for solving the problem are being experimentally substantiated with test data from full-scale and scale-model container drop tests. 6 figures, 1 table

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

    OpenAIRE

    Shenton B.; Philipose K.

    2011-01-01

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

  19. Optimal Protection of Reactor Hall Under Nuclear Fuel Container Drop Using Simulation Methods

    Directory of Open Access Journals (Sweden)

    Králik Juraj

    2014-12-01

    Full Text Available This paper presents of the optimal design of the damping devices cover of reactor hall under impact of nuclear fuel container drop of type TK C30. The finite element idealization of nuclear power plant structure is used in software ANSYS. The steel pipe damper system is proposed for dissipation of the kinetic energy of the container free fall in comparison with the experimental results. The probabilistic and sensitivity analysis of the damping devices was considered on the base of the simulation methods in program AntHill using the Monte Carlo method.

  20. Simulation and Experimental Validation of Electromagnetic Signatures for Monitoring of Nuclear Material Storage Containers

    International Nuclear Information System (INIS)

    Aker, Pamela M.; Bunch, Kyle J.; Jones, Anthony M.

    2013-01-01

    Previous research at the Pacific Northwest National Laboratory (PNNL) has demonstrated that the low frequency electromagnetic (EM) response of a sealed metallic container interrogated with an encircling coil is a strong function of its contents and can be used to form a distinct signature which can confirm the presence of specific components without revealing hidden geometry or classified design information. Finite element simulations have recently been performed to further investigate this response for a variety of configurations composed of an encircling coil and a typical nuclear material storage container. Excellent agreement was obtained between simulated and measured impedance signatures of electrically conducting spheres placed inside an AT-400R nuclear container. Simulations were used to determine the effects of excitation frequency and the geometry of the encircling coil, nuclear container, and internal contents. The results show that it is possible to use electromagnetic models to evaluate the application of the EM signature technique to proposed versions of nuclear weapons containers which can accommodate restrictions imposed by international arms control and treaty verification legislation

  1. Effect of controlled potential on SCC of nuclear waste package container materials

    International Nuclear Information System (INIS)

    Lum, B. Y.; Roy, A. K.; Spragge, M. K.

    1999-01-01

    The slow-strain-rate (SSR) test technique was used to evaluate the susceptibility of Titanium (Ti) Gr-7 (UNS R52400) and Ti Gr-12 (UNS R53400) to stress corrosion cracking (SCC). Ti Gr-7 and Ti Gr-12 are two candidate container materials for the multi-barrier package for nuclear waste. The tests were done in a deaerated 90 C acidic brine (pH ∼ 2.7) containing 5 weight percent (wt%) sodium chloride (NaCl) using a strain rate of 3.3 x 10 -6 sec -1 . Before being tested in the acidic brine, specimens of each alloy were pulled inside the test chamber in the dry condition at ambient temperature. Then while in the test solution, specimens were strained under different cathodic (negative) controlled electrochemical potentials. These controlled potentials were selected based on the corrosion potential measured in the test solution before the specimens were strained. Results indicate that the times to failure (TTF) for Ti Gr-12 were much shorter than those for Ti Gr-7. Furthermore, as the applied potential became more cathodic, Ti Gr-12 showed reduced ductility in terms of percent reduction in area (%RA) and true fracture stress (σ f ). In addition, TTF and percent elongation (%El) reached the minimum values when Ti Gr-12 was tested under an impressed potential of -1162 mV. However, for Ti Gr-7, all these ductility parameters were not significantly influenced by the changes in applied potential. In general, the results of hydrogen analysis by secondary ion mass spectrometry (SIMS) showed increased hydrogen concentration at more cathodic controlled potentials. Optical microscopy and scanning electron microscopy (SEM) were used to evaluate the morphology of cracking both at the primary fracture face and the secondary cracks along the gage section of the broken tensile specimen. Transgranular secondary cracks were observed in both alloys possibly resulting from the formation of brittle titanium hydrides due to cathodic charging. The primary fracture face was characterized

  2. Damage evaluation of 500 MWe Indian Pressurized Heavy Water Reactor nuclear containment for aircraft impact

    Energy Technology Data Exchange (ETDEWEB)

    Kukreja, Mukesh [Reactor Safety Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)]. E-mail: mrkukreja@yahoo.com

    2005-08-01

    Safety assessment of Indian nuclear containments has been carried out for aircraft impact. The loading time history for Boeing and Airbus categories of aircrafts is generated based on the principle of momentum transfer of crushable aircrafts. The case studies include the analysis of BWR Mark III containment as a benchmark problem and analyses of Pressurised Heavy Water Reactor containment (inner and outer containment) for impulsive loading due to aircraft impact. Initially, the load is applied on outer containment wall model and subsequently the load is transferred to inner containment after the local perforation of the outer containment wall is noticed in the transient simulation. The analysis methodology evolved in the present work would be useful for studying the behavior of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due to the commercial aircrafts operated in India.

  3. Damage evaluation of 500 MWe Indian Pressurized Heavy Water Reactor nuclear containment for aircraft impact

    International Nuclear Information System (INIS)

    Kukreja, Mukesh

    2005-01-01

    Safety assessment of Indian nuclear containments has been carried out for aircraft impact. The loading time history for Boeing and Airbus categories of aircrafts is generated based on the principle of momentum transfer of crushable aircrafts. The case studies include the analysis of BWR Mark III containment as a benchmark problem and analyses of Pressurised Heavy Water Reactor containment (inner and outer containment) for impulsive loading due to aircraft impact. Initially, the load is applied on outer containment wall model and subsequently the load is transferred to inner containment after the local perforation of the outer containment wall is noticed in the transient simulation. The analysis methodology evolved in the present work would be useful for studying the behavior of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due to the commercial aircrafts operated in India

  4. Concept study of a hydrogen containment process during nuclear thermal engine ground testing

    Directory of Open Access Journals (Sweden)

    Ten-See Wang

    Full Text Available A new hydrogen containment process was proposed for ground testing of a nuclear thermal engine. It utilizes two thermophysical steps to contain the hydrogen exhaust. First, the decomposition of hydrogen through oxygen-rich combustion at higher temperature; second, the recombination of remaining hydrogen with radicals at low temperature. This is achieved with two unit operations: an oxygen-rich burner and a tubular heat exchanger. A computational fluid dynamics methodology was used to analyze the entire process on a three-dimensional domain. The computed flammability at the exit of the heat exchanger was less than the lower flammability limit, confirming the hydrogen containment capability of the proposed process. Keywords: Hydrogen decomposition reactions, Hydrogen recombination reactions, Hydrogen containment process, Nuclear thermal propulsion, Ground testing

  5. The PEACE PIPE: Recycling nuclear weapons into a TRU storage/shipping container

    International Nuclear Information System (INIS)

    Floyd, D.; Edstrom, C.; Biddle, K.; Orlowski, R.; Geinitz, R.; Keenan, K.; Rivera, M.

    1997-01-01

    This paper describes results of a contract undertaken by the National Conversion Pilot Project (NCPP) at the Rocky Flats Environmental Technology Site (RFETS) to fabricate stainless steel ''pipe'' containers for use in certification testing at Sandia National Lab, Albuquerque to qualify the container for both storage of transuranic (TRU) waste at RFETS and other DOE sites and shipping of the waste to the Waste Isolation Pilot Project (WIPP). The paper includes a description of the nearly ten-fold increase in the amount of contained plutonium enabled by the product design, the preparation and use of former nuclear weapons facilities to fabricate the components, and the rigorous quality assurance and test procedures that were employed. It also describes how stainless steel nuclear weapons components can be converted into these pipe containers, a true ''swords into plowshare'' success story

  6. Nuclear power plant containment metallic pressure boundary materials and plans for collecting and presenting their properties

    International Nuclear Information System (INIS)

    Oland, C.B.

    1995-04-01

    A program is being conducted at the Oak Ridge National Laboratory (ORNL to assist the Nuclear Regulatory Commission (NRC)) in their assessment of the effects of degradation (primarily corrosion) on the structural capacity and leaktight integrity of metal containments and steel liners of reinforced concrete structures in nuclear power plants. One of the program objectives is to characterize and quantify manifestations of corrosion on the properties of steels used to construct containment pressure boundary components. This report describes a plan for use in collecting and presenting data and information on ferrous alloys permitted for use in construction of pressure retaining components in concrete and metal containments. Discussions about various degradation mechanisms that could potentially affect the mechanical properties of these materials are also included. Conclusions and recommendations presented in this report will be used to guide the collection of data and information that will be used to prepare a material properties data base for containment steels

  7. A survey of repair practices for nuclear power plant containment metallic pressure boundaries

    Energy Technology Data Exchange (ETDEWEB)

    Oland, C.B.; Naus, D.J. [Oak Ridge National Lab., TN (United States)

    1998-05-01

    The Nuclear Regulatory Commission has initiated a program at the Oak Ridge National Laboratory to provide assistance in their assessment of the effects of potential degradation on the structural integrity and leaktightness of metal containment vessels and steel liners of concrete containments in nuclear power plants. One of the program objectives is to identify repair practices for restoring metallic containment pressure boundary components that have been damaged or degraded in service. This report presents issues associated with inservice condition assessments and continued service evaluations and identifies the rules and requirements for the repair and replacement of nonconforming containment pressure boundary components by welding or metal removal. Discussion topics include base and welding materials, welding procedure and performance qualifications, inspection techniques, testing methods, acceptance criteria, and documentation requirements necessary for making acceptable repairs and replacements so that the plant can be returned to a safe operating condition.

  8. A survey of repair practices for nuclear power plant containment metallic pressure boundaries

    International Nuclear Information System (INIS)

    Oland, C.B.; Naus, D.J.

    1998-05-01

    The Nuclear Regulatory Commission has initiated a program at the Oak Ridge National Laboratory to provide assistance in their assessment of the effects of potential degradation on the structural integrity and leaktightness of metal containment vessels and steel liners of concrete containments in nuclear power plants. One of the program objectives is to identify repair practices for restoring metallic containment pressure boundary components that have been damaged or degraded in service. This report presents issues associated with inservice condition assessments and continued service evaluations and identifies the rules and requirements for the repair and replacement of nonconforming containment pressure boundary components by welding or metal removal. Discussion topics include base and welding materials, welding procedure and performance qualifications, inspection techniques, testing methods, acceptance criteria, and documentation requirements necessary for making acceptable repairs and replacements so that the plant can be returned to a safe operating condition

  9. Process and container system for transferring or transporting fuel elements from a nuclear power station to a store

    International Nuclear Information System (INIS)

    Vox, A.J.

    1984-01-01

    A system of containers with three types of containers (an inside container, a transport container and a storage container) is used. One either sets the inside container open on the lid side into the transport container first in the water pond of the nuclear power station, and one then sets the fuel elements into the inside container, or one places the inside container, loaded with fuel elements away from the transport container, into the transport container. Both containers are then closed and are transported to the store as a unit. The storage container open on the lid side is prepared there, the floor of the transport container is opened and this, together with the inside container, is lifted above the storage container or set above the storage container. The inside container is then lowered onto the storage container, the transport container is removed and the lid of the storage container is closed. (orig./HP) [de

  10. Concept study of a hydrogen containment process during nuclear thermal engine ground testing

    OpenAIRE

    Wang, Ten-See; Stewart, Eric T.; Canabal, Francisco

    2016-01-01

    A new hydrogen containment process was proposed for ground testing of a nuclear thermal engine. It utilizes two thermophysical steps to contain the hydrogen exhaust. First, the decomposition of hydrogen through oxygen-rich combustion at higher temperature; second, the recombination of remaining hydrogen with radicals at low temperature. This is achieved with two unit operations: an oxygen-rich burner and a tubular heat exchanger. A computational fluid dynamics methodology was used to analyze ...

  11. Containment nuclear plant structures evaluation by non destructive testing: strategy and results

    OpenAIRE

    GARNIER, Vincent; HENAULT, Jean-Marie; HAFID, Hamid; VERDIER, Jérôme; CHAIX, Jean François; ABRAHAM, Odile; SBARTAÏ, Zoubir Medhi; BALAYSSAC, Jean Pierre; PIWAKOWSKI, Bogdan; VILLAIN, Géraldine; DEROBERT, Xavier; PAYAN, Cédric; RAKOTONARIVO, Sandrine; LAROSE, Eric; SOGBOSSI, Hognon

    2016-01-01

    Containment nuclear plants structures are an ultimate barrier in the event of an accident. Mechanical resistance and tightness are the two functions that they are expected to provide. To evaluate their capacity to perform them, destructive testing cannot be used to characterize the material. Non-Destructive Tests then represent a relevant solution to test concrete and the struc- ture. The article positions NDT within the context of containment structures supervision and maintenance, and prese...

  12. Specification for self contained emergency luminiare and their qualification for a nuclear power plant

    International Nuclear Information System (INIS)

    Srinivasan, R.; Shanmugam, T.K.

    1999-01-01

    Self contained emergency luminiare (SCEL) for application in a nuclear plant shall meet the illumination level requirement of ANSI/NFPA 101-1988 (Life Safety Code) Section 5.8. The testing shall be done as per IS 9583-1981 requirements. In the selection of self contained emergency luminiare the Sealed Maintenance Free (SMF) battery characteristic and Ampere-Hour ratings are to be carefully evaluated

  13. Fission products distributions in Candu primary heat transport and Candu containment systems during a severe accident

    International Nuclear Information System (INIS)

    Constantin, Marin; Rizoiu, Andrei

    2005-01-01

    The paper is intended to analyse the distribution of the fission products (FPs) in CANDU Primary Heat Transport (PHT) and CANDU Containment Systems by using the ASTEC code (Accident Source Term Evaluation Code). The complexity of the data required by ASTEC and the complexity both of CANDU PHT and Containment System were strong motivations to begin with a simplified geometry in order to avoid the introducing of unmanageable errors at the level of input deck. Thus only 1/4 of the PHT circuit was simulated and a simplified FPs inventory, some simplifications in the feeders geometry and containment were used. The circuit consists of 95 horizontal fuel channels connected to 95 horizontal out-feeders, then through vertical feeders to the outlet-header (a big pipe that collects the water from feeders); the circuit continues from the outlet-header with a riser and then with the steam generator and a pump. After this pump, the circuit was broken; in this point the FPs are transferred to the containment. The containment model consists of 4 rooms connected between by 6 links. The data related to the nodes' definitions, temperatures and pressure conditions were chosen as possible as real data from CANDU NPP loss of coolant accident sequence. Temperature and pressure conditions in the time of the accident were calculated by the CATHENA code and the source term of FPs introduced into the PHT was estimated by the ORIGEN code. The FPs distribution in the nodes of the circuit and the FPs mass transfer per isotope and chemical species are obtained by using SOPHAEROS module of ASTEC code. The distributions into the containment are obtained by the CPA module of ASTEC code (thermalhydraulics calculations in the containment and FPs aerosol transport). The results consist of mass distributions in the nodes of the circuit and the transferred mass to the containment through the break for different species (FPs and chemical species) and mass distributions in the different parts and

  14. Characterization of in-containment cables for nuclear plant life extension

    International Nuclear Information System (INIS)

    DuCharme, A.R.; Bustard, L.D.

    1988-01-01

    Electrical cable is made by a large number of manufacturers and used for a variety of applications in nuclear plants. cables have been identified in the Monticello and Surry Pilot Plant life extension studies and the NRC Nuclear Plant Aging Research Program as components important to the economic and safety aspects of life extension. Currently, fitness for service is largely determined by preoperational testing. The US Department of Energy is supporting work at Sandia National Laboratories to assess the technical basis for the life extension of cables found inside containment at US nuclear plants. The work is being performed in coordination with the Nuclear Management and Resource Council's (NUMARC) NUPLEX Working Group. The initial task of this effort is to characterize the design attributes of in-containment cables. This has been completed via development of a data base depicting the manufacturer, type, material composition, use, qualification, and relative popularity of cables installed in containment. Other ongoing work is focused on a review of cable operational experience and assessment of the issues affecting cable life extension. In the long term, the work aims to identify the technical criteria and life extension strategies needed to support continued cable qualification by nuclear plant owner/operators. 7 refs., 4 tabs

  15. Heat transfer effects in vertically emplaced high level nuclear waste container

    International Nuclear Information System (INIS)

    Moujaes, S.F.; Lei, Y.M.

    1994-01-01

    Modeling free convection heat transfer in an cylindrical annular enclosure is still an active area of research and an important problem to be addressed in the high level nuclear waste repository. For the vertically emplaced waste container, the air gap which is between the container shell and the rock borehole, have an important role of dissipating heat to surrounding rack. These waste containers are vertically emplaced in the borehole 300 meters below ground, and in a horizontal grid of 30 x 8 meters apart. The borehole will be capped after the container emplacement. The expected initial heat generated is between 3--4.74 kW per container depending on the type of waste. The goal of this study is to use a computer simulation model to find the borehole wall, air-gap and the container outer wall temperature distributions

  16. Nuclear Power and Resource Efficiency—A Proposal for a Revised Primary Energy Factor

    Directory of Open Access Journals (Sweden)

    Ola Eriksson

    2017-06-01

    Full Text Available Measuring resource efficiency can be achieved using different methods, of which primary energy demand is commonly used. The primary energy factor (PEF is a figure describing how much energy from primary resources is being used per unit of energy delivered. The PEF for nuclear power is typically 3, which refers to thermal energy released from fission in relation to electricity generated. Fuel losses are not accounted for. However; nuclear waste represents an energy loss, as current plans for nuclear waste management mostly include final disposal. Based on a literature review and mathematical calculations of the power-to-fuel ratio for nuclear power, PEF values for the open nuclear fuel cycle (NFC option of nuclear power and different power mixes are calculated. These calculations indicate that a more correct PEF for nuclear power would be 60 (range 32–88; for electricity in Sweden (41% nuclear power PEF would change from 1.8 to 25.5, and the average PEF for electricity in the European Union (EU would change from 2.5 to 18. The results illustrate the poor resource efficiency of nuclear power, which paves the way for the fourth generation of nuclear power and illustrates the policy implication of using PEFs which are inconsistent with current waste management plans.

  17. PANDA a multi-purpose thermal-hydraulics facility devoted to nuclear reactor containment safety analysis

    International Nuclear Information System (INIS)

    Paladino, Domenico

    2014-01-01

    This paper presents the multi purpose facility PANDA devised for the safety analysis of nuclear reactor containment. The passive safety systems for LWRs have been explained with details about the PAssive Nachzerfallswärmeabfuhr und Druck-Abbau Testanlage (PANDA)

  18. Method for forming nuclear fuel containers of a composite construction and the product thereof

    International Nuclear Information System (INIS)

    Cheng, B.-C.; Rosenbaum, H.S.; Armijo, J.S.

    1981-01-01

    An improved method of producing a composite nuclear fuel container is described which comprises a casing or fuel sheath of zirconium or its alloy with a lining cladding of deposited copper superimposed over the inside surface of the zirconium or alloy and a layer of oxide of the zirconium or alloy formed on the inside surface of the casing or sheath. (U.K.)

  19. Impact of ACI-ASME code on design and construction of nuclear containment structures

    International Nuclear Information System (INIS)

    Reedy, R.F.

    1978-01-01

    The effect of the ACI-ASME code for design and construction of concrete containment structures on the nuclear and concrete industries is examined. Topics covered include purpose of the code, general requirements, responsibilities and duties, design and construction specifications, quality assurance, inspection, the liner, and stamping

  20. Foreign research reactor spent nuclear fuel inventories containing HEU and LEU of US-origin

    International Nuclear Information System (INIS)

    Matos, J.E.

    1995-01-01

    This paper provides estimates of the quantities and types of foreign research reactor spent nuclear fuel containing HEU and LEU of US-origin that are anticipated during the period beginning in January 1996 and extending for 10-15 years

  1. Electric cable insulation pyrolysis and ignition resulting from potential hydrogen burn scenarios for nuclear containment buildings

    International Nuclear Information System (INIS)

    Berlad, A.L.; Jaung, R.; Pratt, W.T.

    1982-01-01

    Electric cable insulation in nuclear containment buildings may participate in pyrolysis and combustion processes engendered by hydrogen burn phenomena. This paper examines these pyrolysis/ignition processes of those polymeric materials present in the electric cable insulation and their possible relation to hydrogen burn scenarios

  2. 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

  3. Nuclear fuel element containing particles of an alloyed Zr, Ti, and Ni getter material

    International Nuclear Information System (INIS)

    Grossman, L.N.; Levin, H.A.

    1975-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed. The nuclear fuel element has disposed therein an alloy having the essential components of nickel, titanium and zirconium, and the alloy reacts with water, water vapor and reactive gases at reactor ambient temperatures. The alloy is disposed in the plenum of the fuel element in the form of particles in a hollow gas permeable container having a multiplicity of openings of size smaller than the size of the particles. The openings permit gases and liquids entering the plenum to contact the particles of alloy. The container is preferably held in the spring in the plenum of the fuel element. (Official Gazette)

  4. Concept study of a hydrogen containment process during nuclear thermal engine ground testing

    Science.gov (United States)

    Wang, Ten-See; Stewart, Eric T.; Canabal, Francisco

    A new hydrogen containment process was proposed for ground testing of a nuclear thermal engine. It utilizes two thermophysical steps to contain the hydrogen exhaust. First, the decomposition of hydrogen through oxygen-rich combustion at higher temperature; second, the recombination of remaining hydrogen with radicals at low temperature. This is achieved with two unit operations: an oxygen-rich burner and a tubular heat exchanger. A computational fluid dynamics methodology was used to analyze the entire process on a three-dimensional domain. The computed flammability at the exit of the heat exchanger was less than the lower flammability limit, confirming the hydrogen containment capability of the proposed process.

  5. Phenomenological uncertainty analysis of early containment failure at severe accident of nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Su Won

    2011-02-15

    The severe accident has inherently significant uncertainty due to wide range of conditions and performing experiments, validation and practical application are extremely difficult because of its high temperature and pressure. Although internal and external researches were put into practice, the reference used in Korean nuclear plants were foreign data of 1980s and safety analysis as the probabilistic safety assessment has not applied the newest methodology. Also, it is applied to containment pressure formed into point value as results of thermal hydraulic analysis to identify the probability of containment failure in level 2 PSA. In this paper, the uncertainty analysis methods for phenomena of severe accident influencing early containment failure were developed, the uncertainty analysis that apply Korean nuclear plants using the MELCOR code was performed and it is a point of view to present the distribution of containment pressure as a result of uncertainty analysis. Because early containment failure is important factor of Large Early Release Frequency(LERF) that is used as representative criteria of decision-making in nuclear power plants, it was selected in this paper among various modes of containment failure. Important phenomena of early containment failure at severe accident based on previous researches were comprehended and methodology of 7th steps to evaluate uncertainty was developed. The MELCOR input for analysis of the severe accident reflected natural circulation flow was developed and the accident scenario for station black out that was representative initial event of early containment failure was determined. By reviewing the internal model and correlation for MELCOR model relevant important phenomena of early containment failure, the uncertainty factors which could affect on the uncertainty were founded and the major factors were finally identified through the sensitivity analysis. In order to determine total number of MELCOR calculations which can

  6. The economic potential of a cassette-type-reactor-installed nuclear ice-breaking container ship

    International Nuclear Information System (INIS)

    Kondo, Koichi; Takamasa, Tomoji

    1999-01-01

    An improved cassette-type marine reactor MRX (Marine Reactor X) which is currently researched and developed by the Japan Atomic Energy Research Institute is designed to be easily removed and transferred to another ship. If the reactor in a nuclear-powered ship, which is the reason for its higher cost, were replaced by the cassette-type-MRX, the reusability of the MRX would reduce the cost difference between nuclear-powered and diesel ships. As an investigation of one aspect of a cassette-type MRX, we attempted in this study to do an economic review of an MRX-installed nuclear-powered ice-breaking container ship sailing via the Arctic Ocean. The transportation cost between the Far East and Europe to carry one TEU (twenty-foot-equivalent container unit) over the entire life of the ship for an MRX (which is used for a 20-year period)-installed container ship sailing via the Arctic Ocean is about 70% higher than the Suez Canal diesel ship, carrying 8,000 TEU and sailing at 25 knots, and about 10% higher than the Suez Canal diesel ship carrying 4,000 TEU and sailing at 34 knots. The cost for a cassette-type-MRX (which is used for a 40-year period, removed and transferred to a second ship after being used for 20 years in the first ship)-installed nuclear-powered container ship is about 7% lower than that for the one operated for 20 years. Considering any loss or reduction in sales opportunities through the extension of the transportation period, the nuclear-powered container ship via the Arctic Sea is a more suitable means of transportation than a diesel ship sailing at 25 knots via the Suez Canal when the value of the commodities carried exceeds 2,800 dollars per freight ton. (author)

  7. Economic potential of nuclear-powered ice-breaking container ship via the northern sea route

    International Nuclear Information System (INIS)

    Takamasa, Tomoji; Kondo, Koichi

    2000-01-01

    An improved cassette-type marine reactor MRX (Marine Reactor X) which is currently researched and developed by the JAERI is designed to be easily removed and transferred to another ship. If the reactor in a nuclear-powered ship, which is the reason for its higher cost, were replaced by the cassette-type-MRX, the reusability of the MRX would reduce the cost difference between nuclear-powered and diesel ships. As an investigation of one aspect of a cassette-type MRX, we attempted in this study to do an economic review of an MRX-installed nuclear-powered ice-breaking container ship sailing via the Arctic Ocean. The transportation cost between the Far East and Europe to carry one TEU (twenty-foot-equivalent container unit) over the entire life of the ship for an MRX (which is used for a 20-year period)-installed container ship sailing via the Arctic Ocean is about 70% higher than the Suez Canal diesel ship, carrying 8,000 TEU and sailing at 25 knots, and about 10% higher than the Suez Canal diesel ship carrying 4,000 TEU and sailing at 34 knots. The cost for a cassette-type-MRX (which is used for a 40-year period, removed and transferred to a second ship after being used for 20 years in the first ship)-installed nuclear-powered container ship is about 7% lower than that for the one operated for 20 years. Considering any loss or reduction in sales opportunities through the extension of the transportation period, the nuclear-powered container ship via the Arctic Sea is a more suitable means of transportation than a diesel ship sailing at 25 knots via the Suez Canal when the value of the commodities carried exceeds 2,800 dollars per freight ton. (author)

  8. Reliability analysis of nuclear containment without metallic liners against jet aircraft crash

    Energy Technology Data Exchange (ETDEWEB)

    Siddiqui, N.A.; Iqbal, M.A.; Abbas, H. E-mail: abbas_husain@hotmail.com; Paul, D.K

    2003-09-01

    The present study presents a methodology for detailed reliability analysis of nuclear containment without metallic liners against aircraft crash. For this purpose, a nonlinear limit state function has been derived using violation of tolerable crack width as failure criterion. This criterion has been considered as failure criterion because radioactive radiations may come out if size of crack becomes more than the tolerable crack width. The derived limit state uses the response of containment that has been obtained from a detailed dynamic analysis of nuclear containment under an impact of a large size Boeing jet aircraft. Using this response in conjunction with limit state function, the reliabilities and probabilities of failures are obtained at a number of vulnerable locations employing an efficient first-order reliability method (FORM). These values of reliability and probability of failure at various vulnerable locations are then used for the estimation of conditional and annual reliabilities of nuclear containment as a function of its location from the airport. To study the influence of the various random variables on containment reliability the sensitivity analysis has been performed. Some parametric studies have also been included to obtain the results of field and academic interest.

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  10. DNA-index and stereological estimation of nuclear volume in primary and metastatic malignant melanomas

    DEFF Research Database (Denmark)

    Sørensen, Flemming Brandt; Kristensen, I B; Grymer, F

    1990-01-01

    The aim of this study was to investigate the relationship between physical nuclear volume and ploidy level in malignant melanomas, and to analyse the heterogeneity of these two parameters among primary and corresponding secondary tumours. Unbiased stereological estimates of nuclear volume can...

  11. Investigation of chloride-release of nuclear grade resin in PWR primary system coolant

    International Nuclear Information System (INIS)

    Cao Xiaoning; Li Yunde; Li Jinghong; Lin Fangliang

    1997-01-01

    A new preparation technique is developed for making the low-chloride nuclear-grade resin by commercial resin. The chloride remained in nuclear grade resin may release to PWR primary coolant. The amount of released chloride is depended on the concentration of boron, lithium, other anion impurities, and remained chloride concentration in resin

  12. Preliminary analysis of the creep behaviour of nuclear fuel-waste container materials

    International Nuclear Information System (INIS)

    Dutton, R.; Leitch, B.W.; Crosthwaite, J.L.; Kasprick, G.R.

    1996-12-01

    In the Canadian Nuclear Fuel Waste Management Program, it is proposed that nuclear fuel waste be placed in a durable container and disposed of in a deep underground vault. Consideration of various disposal-container designs has identified either titanium or copper as the material suitable for constructing the container shell. As part of the R and D program to examine the structural integrity of the container, creep tests are being conducted on commercially pure titanium and oxygen-free copper. This report presents the preliminary data obtained. It also describes the evaluation of various constitutive equations to represent the creep curves, thus providing the basis for extrapolation of the creep behaviour over the design lifetime of the container. In this regard, a specific focus is placed on equations derived from the 0-Projection Concept. Recognizing that the container lifetime will be determined by the onset of tertiary creep leading to creep rupture, we present the results of the metallographic examination of creep damage. This shows that the tertiary stage in titanium is associated with the formation of transgranular cavities within the region of localized necking of the creep specimens. In contrast, creep damage in copper is in the form of intergranular cavities uniformly distributed throughout the gauge length. These results are analyzed within the context of the extant literature, and their implications for future container design are discussed. (author)

  13. Fault tree analysis of the manufacturing process of nuclear fuel containers

    International Nuclear Information System (INIS)

    Liao Weixian; Men Dechun; Sui Yuxue

    1998-08-01

    The nuclear fuel container consists of barrel body, bottom, cover, locking ring, rubber seal ring, and so on. It should be kept sealed in transportation and storage, so keeps the fuel contained from leakage. Its manufacturing process includes blanking, forming, seam welding, assembling, derusting and painting. The seam welding and assembling of barrel body and bottom are two key procedures, and the slope grinding, barrel body flaring and deep drawing of the bottom are important procedures. Faults in the manufacturing process of the nuclear fuel containers are investigated in details as for its quality requirements. A fault tree is established with products being unqualified as the top event. Five causes resulting in process faults are classified and analysed, and some measures are suggested for controlling different failures in manufacturing. More research work should be conducted in rules how to set up fault trees for manufacturing process

  14. Fabrication and closure development of nuclear waste containers for storage at the Yucca Mountain, Nevada repository

    International Nuclear Information System (INIS)

    Russell, E.W.; Nelson, T.A.; Domian, H.A.; LaCount, D.F.; Robitz, E.S.; Stein, K.O.

    1989-04-01

    US Congress and the President have determined that the Yucca Mountain site in Nevada is to be characterized to determine its suitability for construction of the first US high-level nuclear waste repository. Work in connection with this site is carried out within the Yucca Mountain Project (YMP). Lawrence Livermore National Laboratory (LLNL) has the responsibility for designing, developing, and projecting the performance of the waste package for the permanent storage of high-level nuclear waste. Babcock ampersand Wilcox (B ampersand W) is involved with the YMP as a subcontractor to LLNL. B ampersand W's role is to recommend and demonstrate a method for fabricating the metallic waste container and a method for performing the final closure of the container after it has been filled with waste. Various fabrication and closure methods are under consideration for the production of containers. This paper presents progress to date in identifying and evaluating the candidate manufacturing processes. 2 refs., 1 fig., 7 tabs

  15. Going nuclear. Some implications of the introduction of nuclear energy as the basic primary energy supply of a developped society

    International Nuclear Information System (INIS)

    Haefele, W.; Sassin, W.

    1975-01-01

    On the basis of nuclear energy as primary energy source, the future development potentialities of secondary energies are considered; these energy forms are coal gaseification, process heat for industrial uses and district heating, and mainly hydrogen production which represents 60% of the future secondary energy demands. By using decision tree method, the eventuality of using nuclear energy as unique energy source is examined, and the successive options implied in this approach are analyzed [fr

  16. Thermal stresses at nozzles of nuclear steel containments under LOCA-conditions

    International Nuclear Information System (INIS)

    Sanchez Sarmiento, G.; Bergmann, A.N.

    1986-01-01

    During a loss of coolant accident (LOCA) of a PWR-nuclear power plant, a considerable heating of the containment atmosphere is expected to occur. Transient thermal stresses will appear at the containment as a consequence of a non-uniform rise of its temperature. Applying computer codes based on the finite element method, dimensionless general thermal stresses at nozzles of spherical steel containment have been calculated, varying the principal geometrical parameters and the Biot number for the containment internal surface. Atmosphere temperature and Biot number are assumed constant after the accident. Several plots of the maximum principal stresses are provided, which constitute general results applicable to stress analysis of any particular containment of this kind. (orig.)

  17. Conditions inside Water Pooled in a Failed Nuclear Waste Container and its Effect on Radionuclide Release

    Science.gov (United States)

    Hamdan, L. K.; Walton, J. C.; Woocay, A.

    2009-12-01

    Nuclear power use is expected to expand in the future, as part of the global clean energy initiative, to meet the world’s surging energy demand, and attenuate greenhouse gas emissions, which are mainly caused by fossil fuels. As a result, it is estimated that hundreds of thousands of metric tons of spent nuclear fuel (SNF) will accumulate. SNF disposal has major environmental (radiation exposure) and security (nuclear proliferation) concerns. Storage in unsaturated zone geological repositories is a reasonable solution for dealing with SNF. One of the key factors that determine the performance of the geological repository is the release of radionuclides from the engineered barrier system. Over time, the nuclear waste containers are expected to fail gradually due to general and localized corrosions and eventually infiltrating water will have access to the nuclear waste. Once radionuclides are released, they will be transported by water, and make their way to the accessible environment. Physical and chemical disturbances in the environment over the container will lead to different corrosion rates, causing different times and locations of penetration. One possible scenario for waste packages failure is the bathtub model, where penetrations occur on the top of the waste package and water pools inside it. In this paper the bathtub-type failed waste container is considered. We shed some light on chemical and physical processes that take place in the pooled water inside a partially failed waste container (bathtub category), and the effects of these processes on radionuclide release. Our study considers two possibilities: temperature stratification of the pooled water versus mixing process. Our calculations show that temperature stratification of the pooled water is expected when the waste package is half (or less) filled with water. On the other hand, when the waste package is fully filled (or above half) there will be mixing in the upper part of water. The effect of

  18. Stress corrosion cracking of the tubing materials for nuclear steam generators in an environment containing lead

    International Nuclear Information System (INIS)

    Kim, Kyung Mo; Kim, Uh Chul; Lee, Eun Hee; Hwang, Seong Sik

    2004-01-01

    Steam generator tube materials show a high susceptibility to stress corrosion cracking (SCC) in an environment containing lead species and some nuclear power plants currently have degradation problems associated with lead-induced stress corrosion cracking in a caustic solution. Effects of an applied potential on SCC is tested for middle-annealed Alloy 600 specimens since their corrosion potential can be changed when lead oxide coexists with other oxidizing species like copper oxide in the sludge. In addition, all the steam generator tubing materials used for nuclear power plants being operated and currently under construction in Korea are tested in a caustic solution with lead oxide. (author)

  19. The processing of boron-containing stainless steels for the nuclear industry

    International Nuclear Information System (INIS)

    Harrison, A.H.; King, K.J.; Wilkinson, J.

    1991-01-01

    Stainless steels containing boron additions of up to 2 wt% are used in the nuclear power and fuel reprocessing industries during storage and transportation of spent nuclear fuel elements. The metallurgical characteristics of these steels are described, with particular reference to the manufacture, chemical homogeneity, mechanical properties and weldability of plate products. Results are presented of tests performed on welded fabrications to demonstrate their resistance to impact loading. A neutron absorption meter for simple and rapid measurement of product boron content is described. (author)

  20. Development of CAP code for nuclear power plant containment: Lumped model

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Soon Joon, E-mail: sjhong90@fnctech.com [FNC Tech. Co. Ltd., Heungdeok 1 ro 13, Giheung-gu, Yongin-si, Gyeonggi-do 446-908 (Korea, Republic of); Choo, Yeon Joon; Hwang, Su Hyun; Lee, Byung Chul [FNC Tech. Co. Ltd., Heungdeok 1 ro 13, Giheung-gu, Yongin-si, Gyeonggi-do 446-908 (Korea, Republic of); Ha, Sang Jun [Central Research Institute, Korea Hydro & Nuclear Power Company, Ltd., 70, 1312-gil, Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of)

    2015-09-15

    Highlights: • State-of-art containment analysis code, CAP, has been developed. • CAP uses 3-field equations, water level oriented upwind scheme, local head model. • CAP has a function of linked calculation with reactor coolant system code. • CAP code assessments showed appropriate prediction capabilities. - Abstract: CAP (nuclear Containment Analysis Package) code has been developed in Korean nuclear society for the analysis of nuclear containment thermal hydraulic behaviors including pressure and temperature trends and hydrogen concentration. Lumped model of CAP code uses 2-phase, 3-field equations for fluid behaviors, and has appropriate constitutive equations, 1-dimensional heat conductor model, component models, trip and control models, and special process models. CAP can run in a standalone mode or a linked mode with a reactor coolant system analysis code. The linked mode enables the more realistic calculation of a containment response and is expected to be applicable to a more complicated advanced plant design calculation. CAP code assessments were carried out by gradual approaches: conceptual problems, fundamental phenomena, component and principal phenomena, experimental validation, and finally comparison with other code calculations on the base of important phenomena identifications. The assessments showed appropriate prediction capabilities of CAP.

  1. CFD simulations in the nuclear containment using the DES turbulence models

    International Nuclear Information System (INIS)

    Ding, Peng; Chen, Meilan; Li, Wanai; Liu, Yulan; Wang, Biao

    2015-01-01

    Highlights: • The k-ε based DES model is used in the nuclear containment simulation. • The comparison of results between different turbulent models is obtained. • The superiority of DES models is analyzed. • The computational efficiency with the DES turbulence models is explained. - Abstract: Different species of gases would be released into the containment and cause unpredicted disasters during the nuclear severe accidents. It is important to accurately predict the transportation and stratification phenomena of these gas mixtures. CFD simulations of these thermal hydraulic issues in nuclear containment are investigated in this paper. The main work is to study the influence of turbulence model on the calculation of gas transportation and heat transfer. The k-ε based DES and other frequently used turbulence models are used in the steam and helium release simulation in THAI series experiment. This paper will show the superiority of the DES turbulence model in terms of computational efficiency and accuracy with the experimental results, and analyze the necessities of DES model to simulate the large-scale containment flows with both laminar and turbulence regions

  2. Development of CAP code for nuclear power plant containment: Lumped model

    International Nuclear Information System (INIS)

    Hong, Soon Joon; Choo, Yeon Joon; Hwang, Su Hyun; Lee, Byung Chul; Ha, Sang Jun

    2015-01-01

    Highlights: • State-of-art containment analysis code, CAP, has been developed. • CAP uses 3-field equations, water level oriented upwind scheme, local head model. • CAP has a function of linked calculation with reactor coolant system code. • CAP code assessments showed appropriate prediction capabilities. - Abstract: CAP (nuclear Containment Analysis Package) code has been developed in Korean nuclear society for the analysis of nuclear containment thermal hydraulic behaviors including pressure and temperature trends and hydrogen concentration. Lumped model of CAP code uses 2-phase, 3-field equations for fluid behaviors, and has appropriate constitutive equations, 1-dimensional heat conductor model, component models, trip and control models, and special process models. CAP can run in a standalone mode or a linked mode with a reactor coolant system analysis code. The linked mode enables the more realistic calculation of a containment response and is expected to be applicable to a more complicated advanced plant design calculation. CAP code assessments were carried out by gradual approaches: conceptual problems, fundamental phenomena, component and principal phenomena, experimental validation, and finally comparison with other code calculations on the base of important phenomena identifications. The assessments showed appropriate prediction capabilities of CAP

  3. CFD simulations in the nuclear containment using the DES turbulence models

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Peng [School of Engineering, Sun Yat-Sen University, Guangzhou (China); Chen, Meilan [China Nuclear Power Technology Research Institute, Shenzhen (China); Li, Wanai, E-mail: liwai@mail.sysu.edu.cn [Sino-French Institute of Nuclear Engineering & Technology, Sun Yat-Sen University, Guangzhou (China); Liu, Yulan [School of Engineering, Sun Yat-Sen University, Guangzhou (China); Wang, Biao [Sino-French Institute of Nuclear Engineering & Technology, Sun Yat-Sen University, Guangzhou (China)

    2015-06-15

    Highlights: • The k-ε based DES model is used in the nuclear containment simulation. • The comparison of results between different turbulent models is obtained. • The superiority of DES models is analyzed. • The computational efficiency with the DES turbulence models is explained. - Abstract: Different species of gases would be released into the containment and cause unpredicted disasters during the nuclear severe accidents. It is important to accurately predict the transportation and stratification phenomena of these gas mixtures. CFD simulations of these thermal hydraulic issues in nuclear containment are investigated in this paper. The main work is to study the influence of turbulence model on the calculation of gas transportation and heat transfer. The k-ε based DES and other frequently used turbulence models are used in the steam and helium release simulation in THAI series experiment. This paper will show the superiority of the DES turbulence model in terms of computational efficiency and accuracy with the experimental results, and analyze the necessities of DES model to simulate the large-scale containment flows with both laminar and turbulence regions.

  4. The effects of age on nuclear power plant containment cooling systems

    Energy Technology Data Exchange (ETDEWEB)

    Lofaro, R.; Subudhi, M.; Travis, R.; DiBiasio, A.; Azarm, A. [Brookhaven National Lab., Upton, NY (United States); Davis, J. [Science Applications International Corp., New York, NY (United States)

    1994-04-01

    A study was performed to assess the effects of aging on the performance and availability of containment cooling systems in US commercial nuclear power plants. This study is part of the Nuclear Plant Aging Research (NPAR) program sponsored by the US Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. The effects of age were characterized for the containment cooling system by reviewing and analyzing failure data from national databases, as well as plant-specific data. The predominant failure causes and aging mechanisms were identified, along with the components that failed most frequently. Current inspection, surveillance, and monitoring practices were also examined. A containment cooling system unavailability analysis was performed to examine the potential effects of aging by increasing failure rates for selected components. A commonly found containment spray system design and a commonly found fan cooler system design were modeled. Parametric failure rates for those components in each system that could be subject to aging were accounted for in the model to simulate the time-dependent effects of aging degradation, assuming no provisions are made to properly manage it. System unavailability as a function of increasing component failure rates was then calculated.

  5. The effects of age on nuclear power plant containment cooling systems

    International Nuclear Information System (INIS)

    Lofaro, R.; Subudhi, M.; Travis, R.; DiBiasio, A.; Azarm, A.; Davis, J.

    1994-04-01

    A study was performed to assess the effects of aging on the performance and availability of containment cooling systems in US commercial nuclear power plants. This study is part of the Nuclear Plant Aging Research (NPAR) program sponsored by the US Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. The effects of age were characterized for the containment cooling system by reviewing and analyzing failure data from national databases, as well as plant-specific data. The predominant failure causes and aging mechanisms were identified, along with the components that failed most frequently. Current inspection, surveillance, and monitoring practices were also examined. A containment cooling system unavailability analysis was performed to examine the potential effects of aging by increasing failure rates for selected components. A commonly found containment spray system design and a commonly found fan cooler system design were modeled. Parametric failure rates for those components in each system that could be subject to aging were accounted for in the model to simulate the time-dependent effects of aging degradation, assuming no provisions are made to properly manage it. System unavailability as a function of increasing component failure rates was then calculated

  6. Neutron dosimetry inside the containment building of Spanish nuclear power plants with PADC based dosemeters

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Fuste, M.J. [Grup de Fisica de les Radiacions. Departament de Fisica. Edifici C. Universitat Autonoma de Barcelona, E-08193 Bellaterra (Spain); Domingo, C., E-mail: carles.domingo@uab.ca [Grup de Fisica de les Radiacions. Departament de Fisica. Edifici C. Universitat Autonoma de Barcelona, E-08193 Bellaterra (Spain); Amgarou, K.; Bouassoule, T.; Castelo, J. [Grup de Fisica de les Radiacions. Departament de Fisica. Edifici C. Universitat Autonoma de Barcelona, E-08193 Bellaterra (Spain)

    2009-10-15

    The Spanish Nuclear Safety Council (Consejo de Seguridad Nuclear, CSN) recommends performing neutron individual dose assignments at workplaces based on ambient dose equivalent measurements using area monitors and by estimating the amount of time that workers spend in the different monitored environments. In addition, some Spanish nuclear power plants estimate the neutron dose equivalent using albedo thermoluminescence dosemeters (TLD). In the period 2004-2006, our group, together with other research centers, participated in a project, funded by the CSN, with the support of the Spanish Nuclear Power Plants Association (UNESA), to investigate in situ which could be the best practical procedure for individual neutron dose monitoring in nuclear power plants. As part of this survey, several units of the UAB PADC based neutron dosemeter were exposed, on a methacrylate phantom simulating a human body, at four different places inside the containment building of the Asco I nuclear power plant. The influence of different types of calibration neutron fields is analysed and the dose equivalent for each point is estimated.

  7. Lead iron phosphate glass as a containment medium for disposal of high-level nuclear waste

    Science.gov (United States)

    Boatner, Lynn A.; Sales, Brian C.

    1989-01-01

    Lead-iron phosphate glasses containing a high level of Fe.sub.2 O.sub.3 for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90.degree. C., with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10.sup.2 to 10.sup.3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe.sub.2 O.sub.3 in forming the lead-iron phosphate glass is critical. Lead-iron phosphate nuclear waste glass can be prepared at temperatures as low as 800.degree. C., since they exhibit very low melt viscosities in the 800.degree. to 1050.degree. C. temperature range. These waste-loaded glasses do not readily devitrify at temperatures as high as 550.degree. C. and are not adversely affected by large doses of gamma radiation in H.sub.2 O at 135.degree. C. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms.

  8. Perturbation of baseline thermal stress in the Mound 9516 Shipping Package primary containment vessel

    International Nuclear Information System (INIS)

    Sansalone, K.H.F.

    1995-01-01

    Full-capacity loading of heat sources into the Mound 9516 Shipping Package primary containment vessel (PCV) results in temperature gradients which are symmetric, due to the axisymmetry of the package design. Concern over the change in thermal gradients (and therefore, stress) in the PCV due to sub-capacity loading led to the analytical examination of this phenomenon. The PCVs are cylindrical in shape and are loaded into the package such that they and all containment components are concentrically arranged along a common longitudinal axis. If the design full-capacity loading of the PCVs in this package assumes the axisymmetric (or more precisely, cyclicly symmetric) arrangement of its heat-producing contents, then sub-capacity loading implies that in many cases, the load arrangement could be asymmetric with respect to the longitudinal axis. It is then feasible that the departure from heat load axisymmetry could perturb the nominal thermal gradients so that thermally-induced stress within the PCV might increase to levels deemed unacceptable. This study applies Finite Element analysis (FEA) to the problem and demonstrates that no such unacceptable thermal stress increase occurs in the PCV material due to the asymmetric arrangement of contents. copyright 1995 American Institute of Physics

  9. Analysis CFD for the hydrogen transport in the primary containment of a BWR; Analisis CFD para el transporte de hidrogeno en la contencion primaria de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez P, D. A.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Gomez T, A. M., E-mail: guerreroazteca_69@hotmail.com [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    This study presents a qualitative and quantitative comparison among the CFD GASFLOW and OpenFOAM codes which are related with the phenomenon of hydrogen transport and other gases in the primary containment of a Boiling Water Reactor (BWR). GASFLOW is a commercial license code that is well validated and that was developed in Germany for the analysis of the gases transport in containments of nuclear reactors. On the other hand, OpenFOAM is an open source code that offers several evaluation solvers for different types of phenomena; in this case, the solver reacting-Foam is used to analyze the hydrogen transport inside the primary containment of the BWR. The results that offer the solver reacting-Foam of OpenFOAM are evaluated in the hydrogen transport calculation and the results are compared with those of the program of commercial license GASFLOW to see if is viable the use of the open source code in the case of the hydrogen transport in the primary containment of a BWR. Of the obtained results so much quantitative as qualitative some differences were identified between both codes, the differences (with a percentage of maximum error of 4%) in the quantitative results are small and they are considered acceptable for this analysis type, also, these differences are attributed mainly to the used transport models, considering that OpenFOAM uses a homogeneous model and GASFLOW uses a heterogeneous model. (Author)

  10. Containment failure modes preliminary analysis for Atucha-I nuclear power plant during severe accidents

    International Nuclear Information System (INIS)

    Baron, J.; Caballero, C.; Zarate, S.M.

    1997-01-01

    The present work has the objective to analyze the containment behavior of the Atucha-I nuclear power plant during a severe accident, as part of a probabilistic safety assessment (PSA). Initially, a generic description of the containment failure modes considered in other PSAs is performed. Then, the possible containment failure modes for Atucha I are qualitatively analyzed, according to it design peculiarities. These failure modes involve some substantial differences from other PSAs, due to the particular design of Atucha I. Among others, it is studied the influence of: moderator/coolant separation, existence of cooling Zircaloy channels, existence of filling bodies inside the pressure vessel, reactor cavity geometry, on-line refueling mode, and existence of a double shell containment (steel and concrete) with an annular separation room. As a functions of the before mentioning analysis, a series of parameters to be taken into account is defined, on a preliminary basis, for definition of the plant damage states. (author) [es

  11. An assessment of the feasibility of indefinite containment of Canadian nuclear fuel wastes

    International Nuclear Information System (INIS)

    Shoesmith, D.W.; King, F.; Ikeda, B.M.

    1995-05-01

    This report presents an analysis of the expected corrosion behaviour of nuclear fuel waste containers in a conceptual Canadian disposal vault. The container materials considered are dilute Ti alloys (Grades-2, -12 and -16) and oxygen-free copper. The corrosive conditions within the disposal vault change with time as the initially trapped oxygen is consumed and as the heat and γ-radiation produced by the waste decays. This evolution of the vault environment is broadly classified into an early, warm and oxidizing period followed by a period of long-term, stable, cool and non-oxidizing conditions. The corrosion behaviour of both types of material during these two periods is discussed, and various models that have been developed to predict the lifetimes of the containers are presented. The conclusion is that indefinite containment of the waste is feasible with both copper and titanium alloys under Canadian disposal conditions. (author). refs., tabs., figs

  12. NAUAHYGROS - A code for calculating aerosol behavior in nuclear power plant containments following a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Sher, R. [Rudolph Sher Associates, Stanford, CA (United States); Li, J. [Polestar Applied Technology, Inc., Los Altos, CA (United States)

    1995-02-01

    NAUAHYGROS is a computer code to calculate the behavior of fission product and other aerosol particles in the containment of a nuclear reactor following a severe accident. It is an extension of the German code NAUA, which has been in widespread use for many years. Early versions of NAUA treated various aerosol phenomena in dry atmospheres, including aerosol agglomeration, diffusion (plateout), and settling processes. Later versions added treatments of steam condensation on particles in saturated or supersaturated containment atmospheres. The importance of these condensation effects on aerosol removal rates was demonstrated in large scale simulated containment tests. The additional features incorporated in NAUAHYGROS include principally a treatment of steam condensation on hygroscopic aerosols, which can grow as a result of steam condensation even in superheated atmospheres, and improved modelling of steam condensation on the walls of the containment. The code has been validated against the LACE experiments.

  13. Investigation into the application of polyetherimide to nuclear waste storage containers

    Energy Technology Data Exchange (ETDEWEB)

    Saboui, Y.; Bonin, H.W.; Bui, V.T. [Royal Military College of Canada, Kingston, Ontario (Canada)

    2009-07-01

    The procedure of the analysis of the effects of irradiation on the mechanical and chemical properties of the polyetherimide (PEI) is outlined. Previous research in this field at the Royal Military College of Canada is presented. Samples of PEI will be exposed to a mixed radiation field, in the pool of a SLOWPOKE-2 nuclear reactor, then changes in mechanical properties, degradation product formation, and physical property changes will be assessed. Additionally, the heat transfer in the sample will be calculated in order to model the heat transfer rate and heat diffusion profile of PEI. The purpose of the proposed research is to determine the feasibility of using PEI for spent CANDU nuclear fuel and nuclear waste storage containers. (author)

  14. Investigation into the application of polyetherimide to nuclear waste storage containers

    Energy Technology Data Exchange (ETDEWEB)

    Saboui, Y.; Bonin, H.W.; Bui, V.T. [Royal Military College, Kingston, Ontario (Canada)

    2010-07-01

    The procedure of the analysis of the effects of irradiation on the mechanical and chemical properties of the polyetherimide (PEI) is outlined. Previous research in this field at the Royal Military College of Canada is presented. Samples of PEI will be exposed to a mixed radiation field, in the pool of a SLOWPOKE-2 nuclear reactor, then changes in mechanical properties, degradation product formation, and physical property changes will be assessed. Additionally, the heat transfer in the sample will be calculated in order to model the heat transfer rate and heat diffusion profile of PEI. The purpose of the proposed research is to determine the feasibility of using PEI for spent CANDU nuclear fuel and nuclear waste storage containers. (author)

  15. Investigation into the application of polyetherimide to nuclear waste storage containers

    International Nuclear Information System (INIS)

    Saboui, Y.; Bonin, H.W.; Bui, V.T.

    2009-01-01

    The procedure of the analysis of the effects of irradiation on the mechanical and chemical properties of the polyetherimide (PEI) is outlined. Previous research in this field at the Royal Military College of Canada is presented. Samples of PEI will be exposed to a mixed radiation field, in the pool of a SLOWPOKE-2 nuclear reactor, then changes in mechanical properties, degradation product formation, and physical property changes will be assessed. Additionally, the heat transfer in the sample will be calculated in order to model the heat transfer rate and heat diffusion profile of PEI. The purpose of the proposed research is to determine the feasibility of using PEI for spent CANDU nuclear fuel and nuclear waste storage containers. (author)

  16. The remote handling of canisters containing nuclear waste in glass at the Savannah River Plant

    International Nuclear Information System (INIS)

    Callan, J.E.

    1986-01-01

    The Defense Waste Processing Facility (DWPF) is a complete production area being constructed at the Savannah River Plant for the immobilization of nuclear waste in glass. The remote handling of canisters filled with nuclear waste in glass is an essential part of the process of the DWPF at the Savannah River Plant. The canisters are filled with nuclear waste containing up to 235,000 curies of radioactivity. Handling and movement of these canisters must be accomplished remotely since they radiate up to 5000 R/h. Within the Vitrification Building during filling, cleaning, and sealing, canisters are moved using standard cranes and trolleys and a specially designed grapple. During transportation to the Glass Waste Storage Building, a one-of-a-kind, specially designed Shielded Canister Transporter (SCT) is used. 8 figs

  17. 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)

  18. Deposition of RuO 4 on various surfaces in a nuclear reactor containment

    Science.gov (United States)

    Holm, Joachim; Glänneskog, Henrik; Ekberg, Christian

    2009-07-01

    During a severe nuclear reactor accident with air ingress, ruthenium can be released from the nuclear fuel in the form of ruthenium tetroxide. Hence, it is important to investigate how the reactor containment is able to reduce the source term of ruthenium. The aim of this work was to investigate the deposition of gaseous ruthenium tetroxide on aluminium, copper and zinc, which all appear in relatively large amounts in reactor containment. The experiments show that ruthenium tetroxide is deposited on all the metal surfaces, especially on the copper and zinc surfaces. A large deposition of ruthenium tetroxide also appeared on the relatively inert glass surfaces in the experimental set-ups. The analyses of the different surfaces, with several analytical methods, showed that the form of deposited ruthenium was mainly ruthenium dioxide.

  19. Deposition of RuO{sub 4} on various surfaces in a nuclear reactor containment

    Energy Technology Data Exchange (ETDEWEB)

    Holm, Joachim, E-mail: joachim.holm@chalmers.s [Department of Nuclear Chemistry, Chalmers University of Technology, Se-412 96 Gothenburg (Sweden); Glaenneskog, Henrik [Ringhals AB, SE-430 22, Vaeroebacka (Sweden); Ekberg, Christian [Department of Nuclear Chemistry, Chalmers University of Technology, Se-412 96 Gothenburg (Sweden)

    2009-07-01

    During a severe nuclear reactor accident with air ingress, ruthenium can be released from the nuclear fuel in the form of ruthenium tetroxide. Hence, it is important to investigate how the reactor containment is able to reduce the source term of ruthenium. The aim of this work was to investigate the deposition of gaseous ruthenium tetroxide on aluminium, copper and zinc, which all appear in relatively large amounts in reactor containment. The experiments show that ruthenium tetroxide is deposited on all the metal surfaces, especially on the copper and zinc surfaces. A large deposition of ruthenium tetroxide also appeared on the relatively inert glass surfaces in the experimental set-ups. The analyses of the different surfaces, with several analytical methods, showed that the form of deposited ruthenium was mainly ruthenium dioxide.

  20. The Nuclear Car Wash: Neutron interrogation of cargo containers to detect hidden SNM

    Science.gov (United States)

    Hall, J. M.; Asztalos, S.; Biltoft, P.; Church, J.; Descalle, M.-A.; Luu, T.; Manatt, D.; Mauger, G.; Norman, E.; Petersen, D.; Pruet, J.; Prussin, S.; Slaughter, D.

    2007-08-01

    LLNL is actively involved in the development of advanced technologies for use in detecting threats in sea-going cargo containers, particularly the presence of hidden special nuclear materials (SNM). The "Nuclear Car Wash" (NCW) project presented here uses a high-energy (En ≈ 3.5-7.0 MeV) neutron probe to scan a container and then takes high-energy (Eγ ⩾ 2.5 MeV), β-delayed γ-rays emitted during the subsequent decay of any short-lived, neutron-induced fission products as a signature of fissionable material. The components of the proposed system (e.g. neutron source, gamma detectors, etc.) will be discussed along with data processing schemes, possible threat detection metrics and potential interference signals. Results from recent laboratory experiments using a prototype system at LLNL will also be presented.

  1. A novel technique for finding gas bubbles in the nuclear waste containers using Muon Scattering Tomography

    Science.gov (United States)

    Dobrowolska, M.; Velthuis, J.; Frazão, L.; Kikoła, D.

    2018-05-01

    Nuclear waste is deposited for many years in the concrete or bitumen-filled containers. With time hydrogen gas is produced, which can accumulate in bubbles. These pockets of gas may result in bitumen overflowing out of the waste containers and could result in spread of radioactivity. Muon Scattering Tomography is a non-invasive scanning method developed to examine the unknown content of nuclear waste drums. Here we present a method which allows us to successfully detect bubbles larger than 2 litres and determine their size with a relative uncertainty resolution of 1.55 ± 0.77%. Furthermore, the method allows to make a distinction between a conglomeration of bubbles and a few smaller gas volumes in different locations.

  2. The Nuclear Car Wash: Neutron interrogation of cargo containers to detect hidden SNM

    International Nuclear Information System (INIS)

    Hall, J.M.; Asztalos, S.; Biltoft, P.; Church, J.; Descalle, M.-A.; Luu, T.; Manatt, D.; Mauger, G.; Norman, E.; Petersen, D.; Pruet, J.; Prussin, S.; Slaughter, D.

    2007-01-01

    LLNL is actively involved in the development of advanced technologies for use in detecting threats in sea-going cargo containers, particularly the presence of hidden special nuclear materials (SNM). The 'Nuclear Car Wash' (NCW) project presented here uses a high-energy (E n ∼ 3.5-7.0 MeV) neutron probe to scan a container and then takes high-energy (E γ ≥ 2.5 MeV), β-delayed γ-rays emitted during the subsequent decay of any short-lived, neutron-induced fission products as a signature of fissionable material. The components of the proposed system (e.g. neutron source, gamma detectors, etc.) will be discussed along with data processing schemes, possible threat detection metrics and potential interference signals. Results from recent laboratory experiments using a prototype system at LLNL will also be presented

  3. Variable stiffness lattice support system for a condenser type nuclear reactor containment

    International Nuclear Information System (INIS)

    George, J.A.; Sutherland, J.D.

    1979-01-01

    A support structure for the lattice supporting a fusible material in the annular condenser region of a nuclear reactor containment, the flexibility of which structure can be selectively adjusted in accordance with seismic or other loading requirements. The lattice is affixed to a flexible member in a manner which allows relative movement between the two components. The flexible member is affixed to a rigid support member in a manner which selectively adjusts the resiliency of the flexible member. The support member is rigidly affixed to a wall of the containment annulus, and can also be utilized to support cooling ducts. 6 claims

  4. Fabrication and closure development of nuclear waste disposal containers for the Yucca Mountain Project: Status report

    International Nuclear Information System (INIS)

    Domian, H.A.; Robitz, E.S.; Conrardy, C.C.; LaCount, D.F.; McAninch, M.D.; Fish, R.L.; Russell, E.W.

    1991-09-01

    In GFY 89, a project was underway to determine and demonstrate a suitable method for fabricating thin-walled monolithic waste containers for service within the potential repository at Yucca Mountain. A concurrent project was underway to determine and demonstrate a suitable closure process for these containers after they have been filled with high-level nuclear waste. Phase 1 for both the fabrication and closure projects was a screening phase in which candidate processes were selected for further laboratory testing in Phase 2. This report describes the final results of the Phase 1 efforts. It also describes the preliminary results of Phase 2 efforts

  5. Influences of nuclear containment radius on the aircraft impact force based on the Riera function

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, T.; Wu, H., E-mail: abrahamhao@126.com; Fang, Q.; Gong, Z.M.

    2015-11-15

    Highlights: • A fine aircraft model of A320 was built and verified by available limited prototype impacting tests. • The influences of aircraft longitudinal crushing strength on the impact process were analyzed. • The influences of NPP containment radius on the impact force were numerically studied. • The Riera function was modified by considering the radius effect of NPP containment. - Abstract: The aircraft impact force directly influences the local failure and global response of the nuclear power plant (NPP) containment, while the existing theoretical models and the field tests were almost based on the flat target. In order to analyze the radius effect of the circular sectional containment on the impact force, a fine FE model of the commercial aircraft A320 was established and validated by the available limited full-scale F-4 Phantom impact experiment. In order to determine the force to crush the A320 FE model, the influences of aircraft longitudinal crushing strength on the impact process were analyzed based on the Riera function. Considering the containment decaying effect to aircraft impact velocity, the impact impulse was theoretically calculated, while the influences of the losses of mass and energy were not included. The numerical simulations of A320 aircrafts impacting on simplified NPP containments with different radii were conducted, which could well reproduce the airframe crushing and debris scattering. By comparison of the simulated impact impulses and the calculation values by the Riera function, the coefficients corresponding to different containment radii are derived and a fitting formula is obtained. Finally, an improved Riera function dependent on the dimensionless ratio of nuclear containment radius and aircraft wingspan is proposed.

  6. Nuclear containment steel liner corrosion workshop : final summary and recommendation report.

    Energy Technology Data Exchange (ETDEWEB)

    Erler, Bryan A. (Erler Engineering Ltd., Chicago, IL); Weyers, Richard E. (Virginia Tech University, Blacksburg, VA); Sagues, Alberto (University of South Florida, Tampa, FL); Petti, Jason P.; Berke, Neal Steven (Tourney Consulting Group, LLC, Kalamazoo, MI); Naus, Dan J. (Oak Ridge National Laboratory, Oak Ridge, TN)

    2011-07-01

    This report documents the proceedings of an expert panel workshop conducted to evaluate the mechanisms of corrosion for the steel liner in nuclear containment buildings. The U.S. Nuclear Regulatory Commission (NRC) sponsored this work which was conducted by Sandia National Laboratories. A workshop was conducted at the NRC Headquarters in Rockville, Maryland on September 2 and 3, 2010. Due to the safety function performed by the liner, the expert panel was assembled in order to address the full range of issues that may contribute to liner corrosion. This report is focused on corrosion that initiates from the outer surface of the liner, the surface that is in contact with the concrete containment building wall. Liner corrosion initiating on the outer diameter (OD) surface has been identified at several nuclear power plants, always associated with foreign material left embedded in the concrete. The potential contributing factors to liner corrosion were broken into five areas for discussion during the workshop. Those include nuclear power plant design and operation, corrosion of steel in contact with concrete, concrete aging and degradation, concrete/steel non-destructive examination (NDE), and concrete repair and corrosion mitigation. This report also includes the expert panel member's recommendations for future research.

  7. 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)

  8. Study of typical nuclear containment purge valves in an accident environment

    International Nuclear Information System (INIS)

    Watkins, J.C.; Steele, R. Jr.; Hill, R.C.; DeWall, K.G.

    1986-08-01

    This report presents the results of the containment purge and vent valve test program, conducted under the sponsorship of the United States Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research. The test program investigated butterfly valve operability and leak integrity under light-water-reactor design basis and severe accident conditions. Three nuclear-designed butterfly valves typical of those used in domestic nuclear power plant containment purge and vent applications were tested. For a comparison of response, two valve of the same size with differing internal designs were tested. For extrapolation insights, a larger-sized valve similar to one of the smaller valves was also tested. Dynamic flow tests were performed over the range of design basis accident pressures. Leak integrity testing was also performed at both design basis and severe accident temperatures and pressures. The valve experiments were performed with various piping configurations and valve orientations to the flow to simulate the various installation options in field applications. Testing was also performed in a standard ANSI test section

  9. 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)

  10. Study of Real Time Location System For Worker in Containment Building at Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S. Y.; Kim, G. S. [Samchang Enterprise Company, Ulsan (Korea, Republic of); Kim, H. S. [Ulsan Univ., Ulsan (Korea, Republic of)

    2012-03-15

    Workers are required special management to minimize radiation exposure in nuclear power plant. Especially, there are many limitation in their activities at containment building in nuclear power plant. Test personnel shall administer the workers by tracing the location of them inside containment building in nuclear power plant. They may be exposed to the unnecessary radiation due to a complex and high radiation area in the building. Test personnel needs to manage efficiently for worker's safety and work hours at containment building. Therefore, it is critical for the test personnel to notice the risk to the workers by identifying the location when the workers are facing the dangerous situation on the high area. In this paper, we introduce requirements and design method to develop the one and two dimensional RTLS(Real Time Locating System) by using CSS(Chirp Spread Spectrum) which enables precise location measurement and robust data communication even indoor environment with serious electromagnetic interference caused by complicated structure such as the inside of containment building in the nuclear power plant. In the algorithm to compute the distance, it is suggested to use SDS-TWR(Symmetrical Double-Sided Two-Way Ranging) to solve the issue of indirect routes, and develop the power circuit with 10mW of designing gain for output power to meet the KCC standard in order to increase the raging distance, in addition, communication between Anchor and distance measuring computer shall be designed to increase energy using time of Tags(nodes) by using CAN(Controller Area Network) communication.

  11. System for the sealing of containers and pipelines, especially in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Gross, R

    1976-01-22

    In order to seal containers and pipelines especially in nuclear power plants, it is suggested to incorporate hollow bodies of an elastic material in the pipeline connections which can be blown up by pressure-gas and which are placed for sealing on the inner walls of the pipe. During a longer shutdown, system parts can thus be protected from corrosion. Various forms of such cavities are shown in design examples. The sealed sections can be filled with inert gas (nitrogen).

  12. Design of containment system of nuclear fuel attacked by corrosion with leaking fission products

    International Nuclear Information System (INIS)

    Poblete Maturana, Tomas

    2015-01-01

    The following report presents the design of an innovative confinement system for the nuclear fuel attacked by corrosion, with leakage of fission products to be used in the RECH-1 nuclear experimental reactor of the Chilean Nuclear Energy Commission, is currently within the framework of the international nuclear waste management program developed by the member countries of the IAEA, including Chile. The main objective of this project is the development of a system that is capable of containing, in the smallest possible volume, the fission products that are released to the reactor coolant medium from the nuclear fuel that are attacked by corrosion. Among the tasks carried out for the development of the project are: the compilation of the necessary bibliography for the selection of the most suitable technology for the retention of the fission products, the calculation of the most important parameters to ensure that the system will operate within ranges that do not compromise the radiological safety, and the design of the hydraulic circuit of the system. The results obtained from the calculations showed that the fuel element confinement system is stable from a thermal point of view since the refrigerant does not under any circumstances reach the saturation temperature and, in addition, from a hydraulic point of view, since the rate at which the refrigerant flows through the hydraulic circuit is low enough so that the deformation of the fuel plates forming the nuclear fuel does not occur. The most appropriate technology for the extraction of fission products according to the literature consulted is by ion exchange. The calculations developed showed that with a very small volume of resins, it is possible to capture all of the non-volatile fission products of a nuclear fuel

  13. The nuclear receptor gene family in the Pacific oyster, Crassostrea gigas, contains a novel subfamily group.

    Science.gov (United States)

    Vogeler, Susanne; Galloway, Tamara S; Lyons, Brett P; Bean, Tim P

    2014-05-15

    Nuclear receptors are a superfamily of transcription factors important in key biological, developmental and reproductive processes. Several of these receptors are ligand- activated and through their ability to bind endogenous and exogenous ligands, are potentially vulnerable to xenobiotics. Molluscs are key ecological species in defining aquatic and terrestrial habitats and are sensitive to xenobiotic compounds in the environment. However, the understanding of nuclear receptor presence, function and xenobiotic disruption in the phylum Mollusca is limited. Here, forty-three nuclear receptor sequences were mined from the genome of the Pacific oyster, Crassostrea gigas. They include members of NR0-NR5 subfamilies, notably lacking any NR6 members. Phylogenetic analyses of the oyster nuclear receptors have been conducted showing the presence of a large novel subfamily group not previously reported, which is named NR1P. Homologues to all previous identified nuclear receptors in other mollusc species have also been determined including the putative heterodimer partner retinoid X receptor, estrogen receptor and estrogen related receptor. C. gigas contains a highly diverse set of nuclear receptors including a novel NR1 group, which provides important information on presence and evolution of this transcription factor superfamily in invertebrates. The Pacific oyster possesses two members of NR3, the sex steroid hormone receptor analogues, of which there are 9 in humans. This provides increasing evidence that steroid ligand specific expansion of this family is deuterostome specific. This new knowledge on divergence and emergence of nuclear receptors in C. gigas provides essential information for studying regulation of molluscan gene expression and the potential effects of xenobiotics.

  14. Heat transfer effects in vertically emplaced high level nuclear waste container

    International Nuclear Information System (INIS)

    Moujaes, S.F.; Lei, Y.M.

    1994-01-01

    Modeling free convection heat transfer in a cylindrical annular enclosure is still an active area of research and an important problem to be addressed in the high level nuclear waste repository. For the vertically emplaced waste container, the air gap which is between the container shell and the rock borehole, have an important role of dissipating heat to surrounding rock. These waste containers are vertically emplaced in the borehole 300 meters just below ground, and in a horizontal grid of 30 x 8 meters apart. The borehole will be capped after the container emplacement. The expected initial heat generated is between 3-4.74 kW per container depending on the type of waste. The goal of this study is to use a computer simulation model to find the borehole wall, air-gap and the container outer wall temperature distributions. The borehole wall temperature history has been found in the previous study, and was estimated to reach a maximum temperature of about 218 degrees C after 18 years from the emplacement. The temperature history of the rock surface is then used for the air-gap simulation. The problem includes convection and radiation heat transfer in a vertical enclosure. This paper will present the results of the convection in the air-gap over one thousand years after the containers' emplacement. During this long simulation period it was also observed that a multi-cellular air flow pattern can be generated in the air gap

  15. Primary shutdown system monitoring unit for nuclear power plants

    International Nuclear Information System (INIS)

    Khan, Tahir Kamal; Balasubramanian, R.; Agilandaeswari, K.

    2013-01-01

    Shut off rods made up of neutron absorbing material are used as Primary Shutdown System. To reduce the power of the reactor under certain abnormal operating conditions, these rods must go down into the core within a specified time. Any malfunctioning in the movement of rods cannot be tolerated and Secondary Shutdown System (SSS) must be actuated within stipulated time to reduce the reactor power. A special safety critical, hardwired electronics unit has been designed to detect failure of PSS Shut off rods movements and generate trip signals for initiating SSS. (author)

  16. Microbially influenced corrosion of copper nuclear fuel waste containers in a Canadian disposal vault

    Energy Technology Data Exchange (ETDEWEB)

    King, F

    1996-11-01

    An assessment of the potential for microbially influenced corrosion (MIC) of copper nuclear fuel waste containers in a Canadian disposal vault is presented. The assessment is based on a consideration of the microbial activity within a disposal vault, the reported cases of MIC of Cu alloys in the literature and the known corrosion behaviour of Cu. Because of the critical role of biofilms in the reported cases of MIC, their formation and properties are discussed in detail. Next, the literature on the MIC of Cu alloys is briefly reviewed. The various MIC mechanisms proposed are critically discussed and the implications for the corrosion of Cu containers considered. In the majority of literature cases, MIC depends on alternating aerated and deaerated environments, with accelerated corrosion being observed when fresh aerated water replaces stagnant water, e.g., the MIC of Cu-Ni heat exchangers in polluted seawater and the microbially influenced pitting of Cu water pipes. Finally, because of the predominance of corrosion by sulphate-reducing bacteria (SRB) in the MIC literature, the abiotic behaviour of Cu alloys in sulphide solutions is also reviewed. The effect of the evolving environment in a disposal vault on the extent and location of microbial activity is discussed. Biofilm formation on the container surface is considered unlikely throughout the container lifetime, but especially initially when the environmental conditions will be particularly aggressive. Microbial activity in areas of the vault away from the container is possible, however. Corrosion of the container could then occur if microbial metabolic by-products diffuse to the container surface. Sulphide, produced by the action of SRB are considered to be the most likely cause of container corrosion. It is concluded that the only likely form of MIC of Cu containers will result from sulphide produced by SRB diffusing to the container surface. A modelling procedure for predicting the extent of corrosion is

  17. A containment analysis for SBLOCA without ECI in the refurbished Wolsong-1 Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T.M.; Moon, B.J.; Bae, C.J.; Lee, S.H.; Choi, C.J.; Lee, D.S. [NSSS, Korea Power Engineering Company, Inc., Daejeon (Korea, Republic of); Kim, S.M. [NETEC, Korea Hydro and Nuclear Power Company, Inc., Daejeon (Korea, Republic of)

    2010-07-01

    A small break leading to loss of coolant accident (SBLOCA), being one of the topic accidents in the nuclear plant diagnosis in recent years, has been analyzed and evaluated for the refurbished Wolsong-1 Nuclear Power Plant (NPP). The industry standard toolset (IST) codes developed by CANDU Owners Group and updated models including design change parameters are applied to the event analyses. GOTHIC code has been used for the containment analysis of Wolsong-1. Also, SMART-IST code fitted in the Iodine Chemistry (IMOD-2) model has been used to predict nuclide behavior within the containment considering various aspects. IMOD-2 was incorporated into SMART-IST as a module dealing the chemical transformations and mass transfer of iodine species in containment. IMOD-2 model is very sensitive to paint and chemicals. The parameter studies for IMOD-2 model are performed to decide the analysis value set. The developed methodology and the results of SBLOCA without ECI are presented herein. Under the most heat-up conditions, the radionuclide release from the failed fuel into the containment and subsequently to the environment is such that the radioactive doses to the public are below the acceptable limits. (author)

  18. 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.

  19. Evolution of nuclear reactor containments in India: Addressing the present day challenges

    Energy Technology Data Exchange (ETDEWEB)

    Kakodkar, Anil, E-mail: kakodkar@barc.gov.in

    2014-04-01

    Indigenously developed Pressurized Heavy Water Reactors (PHWRs) that form the backbone of current stage of nuclear power development in India have seen continuous evolution of their containment systems. This evolution that has taken place over implementation of 18 PHWRs (200/220/540 MWe) has encompassed all aspects of containment design, viz. the structural system, energy management system, radio-activity management and hydrogen management system. As a part of ongoing efforts toward strengthening of safety performance, India is also ready with the design of Advance Heavy Water Reactor (AHWR), which represents a technology demonstrator for advanced reactor systems and for thorium utilization. This reactor has a number of improved passive safety features and it is capable of meeting the demanding safety challenges that future reactor system would be expected to meet as a result of emerging expectations in the background of accidents over the past three decades viz. those at Three Mile Island (1979), Chernobyl (1986) and most recently at Fukushima (2011). In this lecture I shall focus on the evolution of nuclear reactor containments in India and highlight the design, associated structural and thermal hydraulics safety assessment made over the years for the improvement of containment performance.

  20. Primary Culture of Choroid Plexuses from Neonate Rats Containing Progenitor Cells Capable of Differentiation

    Directory of Open Access Journals (Sweden)

    Sheng-Li Huang

    2013-12-01

    Full Text Available Background: The choroid plexuses, which could secrete a number of neurotrophins, have recently been used in transplantation in central nervous system diseases. Aims: To study the mechanism of nerve regeneration in the central nervous system by grafting choroid plexus tissues. Study Design: Animal experimentation. Methods: The choroid plexuses from the lateral ventricles of neonatal rats were cultured in adherent culture, and immunocytochemical methods were used to analyse the progenitor cells on days 2, 6, and 10 after seeding. Results: Expression of both nestin and glial fibrillary acidic protein was observed in small cell aggregates on day 2 in primary culture. Most of the nestin-positive cells on day 6 were immunoreactive to glial fibrillary acidic protein antibody. No cells expressing nestin or glial fibrillary acidic protein were seen on day 10. Conclusion: These experimental results indicate that the choroid plexus contains a specific cell population – progenitor cells. Under in vitro experimental conditions, the progenitor cells differentiated into choroid plexus epithelial cells but did not form neurons or astrocytes.

  1. Corrosion of copper containers prior to saturation of a nuclear fuel waste disposal vault

    International Nuclear Information System (INIS)

    King, F.; Kolar, M.

    1997-12-01

    The buffer material surrounding the containers in a Canadian nuclear fuel waste disposal vault will partially desiccate as a result of the elevated temperature at the container surface. This will lead to a period of corrosion in a moist air atmosphere. Corrosion will either take the form of slow oxidation if the container surface remains dry or aqueous electrochemical corrosion if the surface is wetted by a thin liquid film. The relevant literature is reviewed, from which it is concluded that corrosion should be uniform in nature, except if the surface is wetted, in which case localized corrosion is a possibility. A quantitative analysis of the extent and rate of uniform corrosion during the unsaturated period is presented. Two bounding cases are considered: first, the case of slow oxidation in moist air following either logarithmic or parabolic oxide-growth kinetics and, second, the case of electrochemically based corrosion occurring in a thin liquid film uninhibited by the growth of corrosion products. (author)

  2. Development of an Ion Chamber for Monitoring the Containment of a Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Song, Tae-Yung; Kim, Han-Soo; Park, Se-Hwan; Ha, Jang-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-05-15

    Nuclear power plants need many different types of radiation detectors for different purposes. Neutron detectors are installed inside and outside of the core to check the neutron flux. Scintillation detectors are used to check the fission products included in the liquids and gases of plant system. Geiger-Mueller counters are used for the area radiation monitoring. In addition to the above-mentioned detectors, ion chambers are installed to monitor radiation level of the containment. A few ion chambers are located within the reactor containment to monitor radiation level of an accident case. Therefore, the ion chamber should be capable of monitoring high level radiation dose up to 10{sup 7} R/h. Korea Atomic Energy Research Institute (KAERI) developed an ion chamber for monitoring the radiation dose inside the containment.

  3. Progress in welding studies for Canadian nuclear fuel waste disposal containers

    International Nuclear Information System (INIS)

    Maak, P.Y.Y.

    1985-11-01

    This report describes the progress in the development of closure-welding technology for Canadian nuclear fuel waste disposal containers. Titanium, copper and Inconel 625 are being investigated as candidate materials for fabrication of these containers. Gas-tungsten-arc welding, gas metal-arc-welding, resistance-heated diffusion bonding and electron beam welding have been evaluated as candidate closure welding processes. Characteristic weldment properties, relative merits of welding techniques, suitable weld joint configurations and fit-up tolerances, and welding parameter control ranges have been identified for various container designs. Furthermore, the automation requirements for candidate welding processes have been assessed. Progress in the development of a computer-controlled remote gas-shielded arc welding system is described

  4. Proceedings of a workshop on corrosion of Nuclear fuel waste containers

    International Nuclear Information System (INIS)

    Shoesmith, D.W.

    1990-01-01

    The 23 papers presented at this conference review the technical merits, and particularly corrosion performance, of the three main materials used for nuclear fuel waste containers: titanium and its alloys, copper and its alloys, and iron and carbon steels. The specific questions posed to the Workshop were: 1) Can we predict the lifetime of container materials in a variety of vault environments? 2) Is there a limiting range of conditions beyond which a specific material cannot be used? 3) Do we have the necessary corrosion rate data and/or mechanistic models required to make predictions? 4) Can we justify the use of titanium on the basis of propagation rate measurements for crevice corrosion, or do we need to prove initiation cannot occur? 5) Will the pitting of copper be significant? 6) How thick a carbon steel container would be required, and can it be fabricated and stress-relieved? 7) Are radiation fields of any consequence at the dose rates expected?

  5. Selection criteria for container materials at the proposed Yucca Mountain high level nuclear waste repository

    International Nuclear Information System (INIS)

    Halsey, W.G.

    1989-11-01

    A geological repository has been proposed for the permanent disposal of the nation's high level nuclear waste at Yucca Mountain in the Nevada desert. The containers for this waste must remain intact for the unprecedented service lifetime of 1000 years. A combination of engineering, regulatory, and licensing requirements complicate the container material selection. In parallel to gathering information regarding the Yucca Mountain service environment and material performance data, a set of selection criteria have been established which compare candidate materials to the performance requirements, and allow a quantitative comparison of candidates. These criteria assign relative weighting to varied topic areas such as mechanical properties, corrosion resistance, fabricability, and cost. Considering the long service life of the waste containers, it is not surprising that the corrosion behavior of the material is a dominant factor. 7 refs

  6. Inspection and mitigating measures for life extension of nuclear power plant containment structures

    International Nuclear Information System (INIS)

    Meyers, B.L.; Daye, M.A.

    1989-01-01

    The subject of life extension of nuclear power plants has drawn considerable attention during the last few years. Interest in life extension initiated because of both safety and economic reasons. A number of evaluations have been performed. The main thrust of earlier work was directed toward evaluating the factors causing aging, and defining degradation sites, degradation mechanism and failure modes. At present, attention is directed toward establishing the appropriate inspection programs suitable for each of the defined aging sites. Some components and aging sites are already subjected to routine inspection programs such as the surveillance of post-tensioning system and the integrated leak rate test (ILRT) of prestressed concrete and reinforced concrete containments. The aging process affects mechanical, electrical and structural components in all types of containments. This paper addresses the structural components of PWR containments only and presents examples of inspection approaches and describes methods to prevent further degradation of select sites

  7. Investigation of cable deterioration in the containment building of the Savannah River Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gillen, K.T.; Clough, R.L.; Jones, L.H.

    1982-08-01

    This report describes an investigation of the deterioration of polyethylene and polyvinylchloride cable materials which occurred in the containment building of the Savannah River nuclear reactor located at Aiken, South Carolina. Radiation dosimetry and temperature mapping data of the containment area indicated that the maximum dose experienced by the cable materials was only 2.5 Mrad at an average operating temperature of 43/sup 0/C. Considering this relatively moderate environment, the amount of material degradation seemed surprising. To understand these findings, an experimental program was performed on the commercial polyethylene and polyvinylchloride materials used at the plant to investigate their degradation behavior under combined ..gamma..-radiation and elevated temperature conditions. It is established that the material deterioration at the plant resulted from radiation-induced oxidation and that the degradation rate can be correlated with local levels of radiation intensity in the containment area.

  8. Investigation of cable deterioration in the containment building of the Savannah River Nuclear Reactor

    International Nuclear Information System (INIS)

    Gillen, K.T.; Clough, R.L.; Jones, L.H.

    1982-08-01

    This report describes an investigation of the deterioration of polyethylene and polyvinylchloride cable materials which occurred in the containment building of the Savannah River nuclear reactor located at Aiken, South Carolina. Radiation dosimetry and temperature mapping data of the containment area indicated that the maximum dose experienced by the cable materials was only 2.5 Mrad at an average operating temperature of 43 0 C. Considering this relatively moderate environment, the amount of material degradation seemed surprising. To understand these findings, an experimental program was performed on the commercial polyethylene and polyvinylchloride materials used at the plant to investigate their degradation behavior under combined γ-radiation and elevated temperature conditions. It is established that the material deterioration at the plant resulted from radiation-induced oxidation and that the degradation rate can be correlated with local levels of radiation intensity in the containment area

  9. Nonlinear dynamic analysis of nuclear reactor primary coolant systems

    International Nuclear Information System (INIS)

    Saffell, B.F. Jr.; Macek, R.W.; Thompson, T.R.; Lippert, R.F.

    1979-01-01

    The ADINA computer code is utilized to perform mechanical response analysis of pressurized reactor primary coolant systems subjected to postulated loss-of-coolant accident (LOCA) loadings. Specifically, three plant analyses are performed utilizing the geometric and material nonlinear analysis capabilities of ADINA. Each reactor system finite element model represents the reactor vessel and internals, piping, major components, and component supports in a single coupled model. Material and geometric nonlinear capabilities of the beam and truss elements are employed in the formulation of each finite element model. Loadings applied to each plant for LOCA dynamic analysis include steady-state pressure, dead weight, strain energy release, transient piping hydraulic forces, and reactor vessel cavity pressurization. Representative results are presented with some suggestions for consideration in future ADINA code development

  10. 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.

  11. 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.

  12. Accidental sequences associated with the containment of the pressurized water nuclear installation - INAP

    International Nuclear Information System (INIS)

    Natacci, Faustina Beatriz; Correa, Francisco

    2002-01-01

    The analysis of accidental sequences associated with the Containment is one of the most important tasks during the development of the Probabilistic Safety Assessment (PSA) of nuclear plants mainly because of its importance on the mitigation of consequences of severe postulated accident initiating events. This paper presents a first approach of the Containment analysis of the INAP identifying failures and events that can compromise its performance, and outlining accidental sequences and Containment end states. The initial plant damage states, which are the input for this study, are based on the event trees developed in the PSA level 1 for the INAP. It should be emphasized that since this PSA is still in a preliminary stage it is subjected to further completion. Consequently, the Containment analysis shall also be revised in order to incorporate, in an extension as complete as possible, all initial plant damage states, the corresponding event trees, and the related Containment end states. Finally, it can be concluded that the evaluation of the qualitative analysis presented herein allows a concise and broad knowledge of the qualitative analysis presented herein allows a concise and broad knowledge of the development of accidental sequences related to the Containment of the INAP. (author)

  13. Comparative evaluation of coating techniques for the corrosion protection of disposal container for spent nuclear fuel

    International Nuclear Information System (INIS)

    Chun, Kwan Sik; Kim, Sung Soo; Park, Chong Mook; Choi, Jong Won

    2005-02-01

    To propose a suitable coating technique to prevent corrosion on metal or metal alloys of a waste container to be used for the disposal of spent nuclear fuel, several methods related to spray coating and vapor deposition techniques have been comparatively evaluated, based on some major factors recommended. From these comparative results, it can be suggested that the best coating methods among the existing techniques in Korea would be HVOF and low pressure plasma spray. Even though the surface of the container coated by these methods would be coated, pores could be remained in the coated film. And therefore post-treatment methods for eliminating the pores have been briefly introduced to keep the life time of the container. The other techniques, the cold spray and hollow cathode discharge, may become excellent coating methods in the future if they are extensively researched to apply for coating on the container. An optimal process among the recommended methods should be selected by considering the state of container, such as an empty or a loaded container, and also related coating materials. For the support to this, the characteristics of the coating materials and the coated films and the durability of this film under a repository condition should be analyzed in detail

  14. Los Alamos National Laboratory new generation standard nuclear material storage container - the SAVY4000 design

    International Nuclear Information System (INIS)

    Stone, Timothy Amos

    2010-01-01

    Incidents involving release of nuclear materials stored in containers of convenience such as food pack cans, slip lid taped cans, paint cans, etc. has resulted in defense board concerns over the lack of prescriptive performance requirements for interim storage of nuclear materials. Los Alamos National Laboratory (LANL) has shared in these incidents and in response proactively moved into developing a performance based standard involving storage of nuclear material (RD003). This RD003 requirements document has sense been updated to reflect requirements as identified with recently issued DOE M 441.1-1 'Nuclear Material Packaging Manual'. The new packaging manual was issued at the encouragement of the Defense Nuclear Facilities Safety Board with a clear directive for protecting the worker from exposure due to loss of containment of stored materials. The Manual specifies a detailed and all inclusive approach to achieve a high level of protection; from package design and performance requirements, design life determinations of limited life components, authorized contents evaluations, and surveillance/maintenance to ensure in use package integrity over time. Materials in scope involve those stored outside an approved engineered-contamination barrier that would result in a worker exposure of in excess of 5 rem Committed Effective Does Equivalent (CEDE). Key aspects of meeting the challenge as developed around the SAVY-3000 vented storage container design will be discussed. Design performance and acceptance criteria against the manual, bounding conditions as established that the user must ensure are met to authorize contents in the package (based upon the activity of heat-source plutonium (90% Pu-238) oxide, which bounds the requirements for weapons-grade plutonium oxide), interface as a safety class system within the facility under the LANL plutonium facility DSA, design life determinations for limited life components, and a sense of design specific surveillance program

  15. Device for Detection of Explosives, Nuclear and Other Hazardous Materials in Luggage and Cargo Containers

    Science.gov (United States)

    Kuznetsov, Andrey; Evsenin, Alexey; Gorshkov, Igor; Osetrov, Oleg; Vakhtin, Dmitry

    2009-12-01

    Device for detection of explosives, radioactive and heavily shielded nuclear materials in luggage and cargo containers based on Nanosecond Neutron Analysis/Associated Particles Technique (NNA/APT) is under construction. Detection module consists of a small neutron generator with built-in position-sensitive detector of associated alpha-particles, and several scintillator-based gamma-ray detectors. Explosives and other hazardous chemicals are detected by analyzing secondary high-energy gamma-rays from reactions of fast neutrons with materials inside a container. The same gamma-ray detectors are used to detect unshielded radioactive and nuclear materials. An array of several neutron detectors is used to detect fast neutrons from induced fission of nuclear materials. Coincidence and timing analysis allows one to discriminate between fission neutrons and scattered probing neutrons. Mathematical modeling by MCNP5 and MCNP-PoliMi codes was used to estimate the sensitivity of the device and its optimal configuration. Comparison of the features of three gamma detector types—based on BGO, NaI and LaBr3 crystals is presented.

  16. Device for Detection of Explosives, Nuclear and Other Hazardous Materials in Luggage and Cargo Containers

    International Nuclear Information System (INIS)

    Kuznetsov, Andrey; Evsenin, Alexey; Osetrov, Oleg; Vakhtin, Dmitry; Gorshkov, Igor

    2009-01-01

    Device for detection of explosives, radioactive and heavily shielded nuclear materials in luggage and cargo containers based on Nanosecond Neutron Analysis/Associated Particles Technique (NNA/APT) is under construction. Detection module consists of a small neutron generator with built-in position-sensitive detector of associated alpha-particles, and several scintillator-based gamma-ray detectors. Explosives and other hazardous chemicals are detected by analyzing secondary high-energy gamma-rays from reactions of fast neutrons with materials inside a container. The same gamma-ray detectors are used to detect unshielded radioactive and nuclear materials. An array of several neutron detectors is used to detect fast neutrons from induced fission of nuclear materials. Coincidence and timing analysis allows one to discriminate between fission neutrons and scattered probing neutrons. Mathematical modeling by MCNP5 and MCNP-PoliMi codes was used to estimate the sensitivity of the device and its optimal configuration. Comparison of the features of three gamma detector types--based on BGO, NaI and LaBr 3 crystals is presented.

  17. 3. Workshop for IAEA ICSP on Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System during Accidents. Presentations

    International Nuclear Information System (INIS)

    2012-04-01

    Most advanced nuclear power plant designs adopted several kinds of passive systems. Natural circulation is used as a key driving force for many passive systems and even for core heat removal during normal operation such as NuScale, CAREM, ESBWR and Indian AHWR designs. Simulation of natural circulation phenomena is very challenging since the driving force of it is weak compared to forced circulation and involves a coupling between primary system and containment for integral type reactor. The IAEA ICSP (International Collaborative Standard Problem) on 'Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System during Accidents' was proposed within the CRP on 'Natural Circulation Phenomena, Modelling, and Reliability of Passive Systems that utilize Natural Circulation'. Oregon State University (OSU) of USA offered to host this ICSP. This ICSP plans to conduct the following experiments and blind/open simulations with system codes: 1. Quasi-steady state operation with different core power levels: Conduct quasi-steady state operation with step-wise increase of core power level in order to observe single phase natural circulation flow according to power level. The experimental facility and operating conditions for an integral PWR will be used. 2. Thermo-hydraulic Coupling between Primary system and Containment: Conduct a loss of feedwater transient with subsequent ADS blowdown and long term cooling to determine the progression of a loss of feedwater transient by natural circulation through primary and containment systems. These tests would examine the blowdown phase as well as the long term cooling using sump natural circulation by coupling the primary to containment systems. This data could be used for the evaluation of system codes to determine if they model specific phenomena in an accurate manner. OSU completed planned two ICSP tests in July 2011 and real initial and boundary conditions measured from the

  18. Range of the radiation monitor for the rigid vent of primary containment during normal and emergency operation for a BWR-5 in Laguna Verde

    International Nuclear Information System (INIS)

    Tijerina S, F.; Pozos S, A. M.; Cabrera U, S.; Mata A, J. A.; Sandoval V, S.; Ovando C, R.; Vargas A, A.; Gallardo R, I.; Cruz G, M.; Amador C, C.

    2014-10-01

    The earthquake followed by a tsunami, happened in March, 2011 in the coasts of oriental Japan, caused damages in the nuclear power plants 1 at 4 of Fukushima Daiichi leading to damage of the fuel in three of the reactors and to the radiation liberation to the exterior. As consequence of those events, the regulations requires that the power plants with Primary Containment type Mark I and II evaluate to have a system of rigid vent with a monitoring equipment of radiation effluents. The present work covers the rigid vent of diameter 12 of the Primary Containment, type Mark-II, of nuclear power plant of Laguna Verde in conditions of severe accident and normal operation, low regime of Extended Power Up rate (EPU - 2317 MWt), using the codes MAAP3B, MICROSHILED 5.05 and the Bardach Black Boxes methodology. As a result the measurement range of the radiation monitor that is required for monitoring the gassy liberation to the atmosphere was determined. The conclusion is that the superior limit of the range of the radiation meter during a Severe Accident is of 8.55 E + 05 R/h (8.55 E + 08 m R/h) and the superior limit in normal operation of 1.412 E-11 at 2.540 E-7 R/h (1.412 E-14 at 2.540 E-10 m R/h). (Author)

  19. Effect on localized waste-container failure on radionuclide transport from an underground nuclear waste vault

    International Nuclear Information System (INIS)

    Cheung, S.C.H.; Chan, T.

    1983-07-01

    In the geological disposal of nuclear fuel waste, one option is to emplace the waste container in a borehole drilled into the floor of the underground vault. In the borehole, the waste container is surrounded by a compacted soil material known as the buffer. A finite-element simulation has been performed to study the effect of localized partial failure of the waste container on the steady-state radionuclide transport by diffusion from the container through the buffer to the surrounding rock and/or backfill. In this study, the radionuclide concentration at the buffer-backfill interface is assumed to be zero. Two cases are considered at the interface between the buffer and the rock. In case 1, a no-flux boundary condition is used to simulate intact rock. In case 2, a constant radionuclide concentration condition is used to simulate fractured rock with groundwater flow. The results show that the effect of localized partial failure of the waste container on the total flux is dependent on the boundary condition at the buffer-rock interface. For the intact rock condition, the total flux is mainly dependent on the location of the failure. The total flux increases as the location changes from the bottom to the top of the emplaced waste container. For a given localized failure of the waste container, the total flux remains unaffected by the area of failed surface below the top of the failure. For fractured rock, the total flux is directly proportional to the failed surface area of the waste container regardless of the failure location

  20. Experimental results from pressure testing a 1:6-scale nuclear power plant containment

    International Nuclear Information System (INIS)

    Horschel, D.S.

    1992-01-01

    This report discusses the testing of a 1:6-scale, reinforced-concrete containment building at Sandia National Laboratories, in Albuquerque, New Mexico. The scale-model, Light Water Reactor (LWR) containment building was designed and built to the American Society of Mechanical Engineers (ASME) code by United Engineers and Constructors, Inc., and was instrumented with over 1200 transducers to prepare for the test. The containment model was tested to failure to determine its response to static internal overpressurization. As part of the US Nuclear Regulatory Commission's program on containment integrity, the test results will be used to assess the capability of analytical methods to predict the performance of containments under severe-accident loads. The scaled dimensions of the cylindrical wall and hemispherical dome were typical of a full-size containment. Other typical features included in the heavily reinforced model were equipment hatches, personnel air locks, several small piping penetrations, and a ihin steel liner that was attached to the concrete by headed studs. In addition to the transducers attached to the model, an acoustic detection system and several video and still cameras were used during testing to gather data and to aid in the conduct of the test. The model and its instrumentation are briefly discussed, and is followed by the testing procedures and measured response of the containment model. A summary discussion is included to aid in understanding the significance of the test as it applies to real world reinforced concrete containment structures. The data gathered during SIT and overpressure testing are included as an appendix

  1. Experimental results from pressure testing a 1:6-scale nuclear power plant containment

    Energy Technology Data Exchange (ETDEWEB)

    Horschel, D.S. [Sandia National Labs., Albuquerque, NM (United States)

    1992-01-01

    This report discusses the testing of a 1:6-scale, reinforced-concrete containment building at Sandia National Laboratories, in Albuquerque, New Mexico. The scale-model, Light Water Reactor (LWR) containment building was designed and built to the American Society of Mechanical Engineers (ASME) code by United Engineers and Constructors, Inc., and was instrumented with over 1200 transducers to prepare for the test. The containment model was tested to failure to determine its response to static internal overpressurization. As part of the US Nuclear Regulatory Commission`s program on containment integrity, the test results will be used to assess the capability of analytical methods to predict the performance of containments under severe-accident loads. The scaled dimensions of the cylindrical wall and hemispherical dome were typical of a full-size containment. Other typical features included in the heavily reinforced model were equipment hatches, personnel air locks, several small piping penetrations, and a ihin steel liner that was attached to the concrete by headed studs. In addition to the transducers attached to the model, an acoustic detection system and several video and still cameras were used during testing to gather data and to aid in the conduct of the test. The model and its instrumentation are briefly discussed, and is followed by the testing procedures and measured response of the containment model. A summary discussion is included to aid in understanding the significance of the test as it applies to real world reinforced concrete containment structures. The data gathered during SIT and overpressure testing are included as an appendix.

  2. Sharing the load: Mex67-Mtr2 cofunctions with Los1 in primary tRNA nuclear export.

    Science.gov (United States)

    Chatterjee, Kunal; Majumder, Shubhra; Wan, Yao; Shah, Vijay; Wu, Jingyan; Huang, Hsiao-Yun; Hopper, Anita K

    2017-11-01

    Eukaryotic transfer RNAs (tRNAs) are exported from the nucleus, their site of synthesis, to the cytoplasm, their site of function for protein synthesis. The evolutionarily conserved β-importin family member Los1 (Exportin-t) has been the only exporter known to execute nuclear export of newly transcribed intron-containing pre-tRNAs. Interestingly, LOS1 is unessential in all tested organisms. As tRNA nuclear export is essential, we previously interrogated the budding yeast proteome to identify candidates that function in tRNA nuclear export. Here, we provide molecular, genetic, cytological, and biochemical evidence that the Mex67-Mtr2 (TAP-p15) heterodimer, best characterized for its essential role in mRNA nuclear export, cofunctions with Los1 in tRNA nuclear export. Inactivation of Mex67 or Mtr2 leads to rapid accumulation of end-matured unspliced tRNAs in the nucleus. Remarkably, merely fivefold overexpression of Mex67-Mtr2 can substitute for Los1 in los1 Δ cells. Moreover, in vivo coimmunoprecipitation assays with tagged Mex67 document that the Mex67 binds tRNAs. Our data also show that tRNA exporters surprisingly exhibit differential tRNA substrate preferences. The existence of multiple tRNA exporters, each with different tRNA preferences, may indicate that the proteome can be regulated by tRNA nuclear export. Thus, our data show that Mex67-Mtr2 functions in primary nuclear export for a subset of yeast tRNAs. © 2017 Chatterjee et al.; Published by Cold Spring Harbor Laboratory Press.

  3. Dose Evaluation of Neutron within Containment Building of a CE type Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Wook; Han, Jae Mun; Kim, Kyung Doek; Yun, Cheol Whan; Suh, Jang Soo; Kim, Young Jae [Nuclear Environment Technology Institute, Daejeon (Korea, Republic of)

    2005-03-15

    From measured results of the neutron fields at some principal places within the containment building in a CE type nuclear power plant in operation, the radiation exposure of a worker to the neutron at there was evaluated and the equivalent dose reflecting new recommendation (ICRP 60) was compared with that doing the old one (ICRP 26). The measured neutron field was also compared with calibration neutron field. From the analysis, the following conclusion was obtained: the average neutron radiation weighting factor according to new recommendation is 2.41 to 2.71 times higher than the old one. The average neutron radiation weighting factor at the measured place was similar to that at calibration neutron field. The average neutron energy at measured place was between 42 and 158 keV and higher than that of calibration field of 500 keV. So, the measured equivalent dose in nuclear power plant could be overestimated compared to the real equivalent dose.

  4. Effluent Containment System for space thermal nuclear propulsion ground test facilities

    International Nuclear Information System (INIS)

    1995-08-01

    This report presents the research and development study work performed for the Space Reactor Power System Division of the U.S. Department of Energy on an innovative ECS that would be used during ground testing of a space nuclear thermal rocket engine. A significant portion of the ground test facilities for a space nuclear thermal propulsion engine are the effluent treatment and containment systems. The proposed ECS configuration developed recycles all engine coolant media and does not impact the environment by venting radioactive material. All coolant media, hydrogen and water, are collected, treated for removal of radioactive particulates, and recycled for use in subsequent tests until the end of the facility life. Radioactive materials removed by the treatment systems are recovered, stored for decay of short-lived isotopes, or packaged for disposal as waste. At the end of the useful life, the facility will be decontaminated and dismantled for disposal

  5. Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

    International Nuclear Information System (INIS)

    Travelli, A.

    1988-01-01

    A nuclear fuel-containing plate structure for a nuclear reactor is described; such structure comprising a pair of malleable metallic non-fissionable matrix plates having confronting surfaces which are pressure bonded together and fully united to form a bonded surface, and elongated malleable wire-like fissionable fuel members separately confined and fully enclosed between the matrix plates along the interface to afford a high fuel density as well as structural integrity and effective retention of fission products. The plates have separate recesses formed in the confronting surfaces for closely receiving the wire-like fissionable fuel members. The wire-like fissionable fuel members are made of a maleable uranium alloy capable of being formed into elongated wire-like members and capable of withstanding pressure bonding. The wire-like fissionable fuel members are completely separated and isolated by fully united portions of the interface

  6. Engineering and innovative erection concept for the containment liner for a nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Goehring, Rainer [Babcock Noell GmbH, Wuerzburg (Germany). Abt. Nukleare Projekte; Anders, Nils; Nagelstutz, Franz [Babcock Noell GmbH, Wuerzburg (Germany). Abt. Berechnung

    2010-04-15

    The determination of the optimal design of the containment liner considering amount of material, manufacturing and erection was the challenge for the engineering team of Babcock Noell GmbH. Several load cases for normal operation and accidental conditions as well as severe accidents have been analysed. The special load cases in the vicinity of penetrations and anchor plates have been calculated. The results of theses analyses have been considered in the actual design of the liner. The construction costs of a Nuclear Power Plant are also impacted by the erection time of components. Therefore, it is necessary to optimize the erection times and consequently essentially to shorten. One possibility has been demonstrated by Babcock Noell GmbH with the erection concept of a containment liner. The liner is preassembled on the pre-assembly place in rings of up to 12 meter height, with the nominal diameter of 47 meter and a weight of approximately up to 200 tonnes. These rings, as well as the containment cup and the dome, are lifted with a heavy load crane into the reactor building. With this concept, the construction activities on the inner containment wall are only five times disrupted by the welding and coating of the circumferential weld (approx. 25 calendar days). In comparison with the common known erection concept of welding of liner shells in situ, at least 20 weeks are saved on the schedule. An integrated concept from planning, manufacturing and erection of this large component has been implemented. It could be demonstrated that within the given time frame, with the required quality and within the required tolerances the containment liner for the Nuclear Power Plant can be delivered to the Client. (orig.)

  7. Engineering and innovative erection concept for the containment liner for a nuclear power plant

    International Nuclear Information System (INIS)

    Goehring, Rainer; Anders, Nils; Nagelstutz, Franz

    2010-01-01

    The determination of the optimal design of the containment liner considering amount of material, manufacturing and erection was the challenge for the engineering team of Babcock Noell GmbH. Several load cases for normal operation and accidental conditions as well as severe accidents have been analysed. The special load cases in the vicinity of penetrations and anchor plates have been calculated. The results of theses analyses have been considered in the actual design of the liner. The construction costs of a Nuclear Power Plant are also impacted by the erection time of components. Therefore, it is necessary to optimize the erection times and consequently essentially to shorten. One possibility has been demonstrated by Babcock Noell GmbH with the erection concept of a containment liner. The liner is preassembled on the pre-assembly place in rings of up to 12 meter height, with the nominal diameter of 47 meter and a weight of approximately up to 200 tonnes. These rings, as well as the containment cup and the dome, are lifted with a heavy load crane into the reactor building. With this concept, the construction activities on the inner containment wall are only five times disrupted by the welding and coating of the circumferential weld (approx. 25 calendar days). In comparison with the common known erection concept of welding of liner shells in situ, at least 20 weeks are saved on the schedule. An integrated concept from planning, manufacturing and erection of this large component has been implemented. It could be demonstrated that within the given time frame, with the required quality and within the required tolerances the containment liner for the Nuclear Power Plant can be delivered to the Client. (orig.)

  8. Interactions of RuO4(g) with different surfaces in nuclear reactor containments

    International Nuclear Information System (INIS)

    Holm, J.; Glaenneskog, H.; Ekberg, C.

    2008-07-01

    During a severe nuclear reactor accident with air ingress, ruthenium in the form of RuO4 can be released from the nuclear fuel. Hence, it is important to investigate how the reactor containment is able to reduce the source term of ruthenium. This work has investigated the distribution of RuO4 between an aqueous and gaseous phase in the temperature interval of 20-50 deg. C by on-line measurements with an experimental set-up made of glass. The experiments showed that RuO4 is almost immediately distributed in the aqueous phase after its introduction in the set-up in the entire temperature interval. However, the deposition of ruthenium on the glass surfaces in the system was significant. The speciation of the ruthenium on the glass surfaces was studied by SEM-EDX and ESCA and was determined to be the expected RuO2. Experiments of interactions between gaseous ruthenium tetroxide and the metals aluminium, copper and zinc have been investigated. The metals were treated by RuO4 (g) at room temperature and analyzed with ESCA, SEM and XRD. The analyses show that the black ruthenium deposits on the metal surfaces were RuO2, i.e. the RuO4 (g) has been transformed on the metal surfaces to RuO2(s). The analyses showed also that there was a significant deposition of ruthenium tetroxide especially on the copper and zinc samples. Aluminium has a lower ability to deposit gaseous ruthenium tetroxide than the other metals. The conclusion that can be made from the results is that surfaces in nuclear reactor containments will likely reduce the source term in the case of a severe accident in a nuclear power plant. (au)

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

    Science.gov (United States)

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

    1991-01-01

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

  10. Effective Thermal Conductivity and Diffusivity of Containment Wall for Nuclear Power Plant OPR1000

    Directory of Open Access Journals (Sweden)

    Hyung Gyun Noh

    2017-04-01

    Full Text Available The goal of this study is to evaluate the effective thermal conductivity and diffusivity of containment walls as heat sinks or passive cooling systems during nuclear power plant (NPP accidents. Containment walls consist of steel reinforced concrete, steel liners, and tendons, and provide the main thermal resistance of the heat sinks, which varies with the volume fraction and geometric alignment of the rebar and tendons, as well as the temperature and chemical composition. The target geometry for the containment walls of this work is the standard Korean NPP OPR1000. Sample tests and numerical simulations are conducted to verify the correlations for models with different densities of concrete, volume fractions, and alignments of steel. Estimation of the effective thermal conductivity and diffusivity of the containment wall models is proposed. The Maxwell model and modified Rayleigh volume fraction model employed in the present work predict the experiment and finite volume method (FVM results well. The effective thermal conductivity and diffusivity of the containment walls are summarized as functions of density, temperature, and the volume fraction of steel for the analysis of the NPP accidents.

  11. 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.

  12. Effective thermal conductivity and diffusivity of containment wall for nuclear power plant OPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Hyung Gyun; Park, Hyun Sun [Div. of Advanced Nuclear Engineering (DANE), Pohang University of Science and Technology (POSTECH), Pohang (Korea, Republic of); Lee, Jong Hwi; Kang, Hie Chan [Mechanical Engineering Div., Kunsan National University (KNU), Gunsan (Korea, Republic of)

    2017-04-15

    The goal of this study is to evaluate the effective thermal conductivity and diffusivity of containment walls as heat sinks or passive cooling systems during nuclear power plant (NPP) accidents. Containment walls consist of steel reinforced concrete, steel liners, and tendons, and provide the main thermal resistance of the heat sinks, which varies with the volume fraction and geometric alignment of the rebar and tendons, as well as the temperature and chemical composition. The target geometry for the containment walls of this work is the standard Korean NPP OPR1000. Sample tests and numerical simulations are conducted to verify the correlations for models with different densities of concrete, volume fractions, and alignments of steel. Estimation of the effective thermal conductivity and diffusivity of the containment wall models is proposed. The Maxwell model and modified Rayleigh volume fraction model employed in the present work predict the experiment and finite volume method (FVM) results well. The effective thermal conductivity and diffusivity of the containment walls are summarized as functions of density, temperature, and the volume fraction of steel for the analysis of the NPP accidents.

  13. Method of constructing lower dry well access tunnel for nuclear reactor container

    International Nuclear Information System (INIS)

    Kume, Tadashi; Furukawa, Hedeyasu.

    1993-01-01

    The method of the present invention facilitates construction of a lower dry well access tunnel for a nuclear reactor container. The lower dry well access tunnel is constructed across the reactor container and the reactor main body foundation. In this case, the lower dry well access tunnel is divided into three sections, i.e., axial end portions and a central portion. At first, each of the end portions is attached to the walls of the reactor container and the reactor main body foundation respectively. Subsequently, the central portion is attached to each of the end portions. An adjusting margin is previously provided to the central portion upon manufacturing each of the sections for adjusting deviation from a nominal size upon construction. In such a construction method, it is possible to eliminate interference accident during construction between the end portions of the lower dry well access tunnel and the reactor container and the reactor main body foundation, to facilitate the construction. Further, this facilitates the fabricating operation for dimensional alignment between the lower dry well access tunnel, and the reactor container and the reactor main body foundation. (I.S.)

  14. 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.

  15. Probabilistic finite element investigation of prestressing loss in nuclear containment wall segments

    International Nuclear Information System (INIS)

    Balomenos, Georgios P.; Pandey, Mahesh D.

    2017-01-01

    Highlights: • Probabilistic finite element framework for assessing concrete strain distribution. • Investigation of prestressing loss based on concrete strain distribution. • Application to 3D nuclear containment wall segments. • Use of ABAQUS with python programing for Monte Carlo simulation. - Abstract: The main function of the concrete containment structures is to prevent radioactive leakage to the environment in case of a loss of coolant accident (LOCA). The Canadian Standard CSA N287.6 (2011) proposes periodic inspections, i.e., pressure testing, in order to assess the strength and design criteria of the containment (proof test) and the leak tightness of the containment boundary (leakage rate test). During these tests, the concrete strains are measured and are expected to have a distribution due to several uncertainties. Therefore, this study aims to propose a probabilistic finite element analysis framework. Then, investigates the relationship between the concrete strains and the prestressing loss, in order to examine the possibility of estimating the average prestressing loss during pressure testing inspections. The results indicate that the concrete strain measurements during the leakage rate test may provide information with respect to the prestressing loss of the bonded system. In addition, the demonstrated framework can be further used for the probabilistic finite element analysis of real scale containments.

  16. Probabilistic finite element investigation of prestressing loss in nuclear containment wall segments

    Energy Technology Data Exchange (ETDEWEB)

    Balomenos, Georgios P., E-mail: gbalomen@uwaterloo.ca; Pandey, Mahesh D., E-mail: mdpandey@uwaterloo.ca

    2017-01-15

    Highlights: • Probabilistic finite element framework for assessing concrete strain distribution. • Investigation of prestressing loss based on concrete strain distribution. • Application to 3D nuclear containment wall segments. • Use of ABAQUS with python programing for Monte Carlo simulation. - Abstract: The main function of the concrete containment structures is to prevent radioactive leakage to the environment in case of a loss of coolant accident (LOCA). The Canadian Standard CSA N287.6 (2011) proposes periodic inspections, i.e., pressure testing, in order to assess the strength and design criteria of the containment (proof test) and the leak tightness of the containment boundary (leakage rate test). During these tests, the concrete strains are measured and are expected to have a distribution due to several uncertainties. Therefore, this study aims to propose a probabilistic finite element analysis framework. Then, investigates the relationship between the concrete strains and the prestressing loss, in order to examine the possibility of estimating the average prestressing loss during pressure testing inspections. The results indicate that the concrete strain measurements during the leakage rate test may provide information with respect to the prestressing loss of the bonded system. In addition, the demonstrated framework can be further used for the probabilistic finite element analysis of real scale containments.

  17. Development of Models to Predict the Redox State of Nuclear Waste Containment Glass

    Energy Technology Data Exchange (ETDEWEB)

    Pinet, O.; Guirat, R.; Advocat, T. [Commissariat a l' Energie Atomique (CEA), Departement de Traitement et de Conditionnement des Dechets, Marcoule, BP 71171, 30207 Bagnols-sur-Ceze Cedex (France); Phalippou, J. [Universite de Montpellier II, Laboratoire des Colloides, Verres et Nanomateriaux, 34095 Montpellier Cedex 5 (France)

    2008-07-01

    Vitrification is one of the recommended immobilization routes for nuclear waste, and is currently implemented at industrial scale in several countries, notably for high-level waste. To optimize nuclear waste vitrification, research is conducted to specify suitable glass formulations and develop more effective processes. This research is based not only on experiments at laboratory or technological scale, but also on computer models. Vitrified nuclear waste often contains several multi-valent species whose oxidation state can impact the properties of the melt and of the final glass; these include iron, cerium, ruthenium, manganese, chromium and nickel. Cea is therefore also developing models to predict the final glass redox state. Given the raw materials and production conditions, the model predicts the oxygen fugacity at equilibrium in the melt. It can also estimate the ratios between the oxidation states of the multi-valent species contained in the molten glass. The oxidizing or reductive nature of the atmosphere above the glass melt is also taken into account. Unlike the models used in the conventional glass industry based on empirical methods with a limited range of application, the models proposed are based on the thermodynamic properties of the redox species contained in the waste vitrification feed stream. The thermodynamic data on which the model is based concern the relationship between the glass redox state and the oxygen fugacity in the molten glass. The model predictions were compared with oxygen fugacity measurements for some fifty glasses. The experiments carried out at laboratory and industrial scale with a cold crucible melter. The oxygen fugacity of the glass samples was measured by electrochemical methods and compared with the predicted value. The differences between the predicted and measured oxygen fugacity values were generally less than 0.5 Log unit. (authors)

  18. 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

  19. One size does not fit all: The impact of primary vaccine container size on vaccine distribution and delivery.

    Science.gov (United States)

    Haidari, Leila A; Wahl, Brian; Brown, Shawn T; Privor-Dumm, Lois; Wallman-Stokes, Cecily; Gorham, Katie; Connor, Diana L; Wateska, Angela R; Schreiber, Benjamin; Dicko, Hamadou; Jaillard, Philippe; Avella, Melanie; Lee, Bruce Y

    2015-06-22

    While the size and type of a vaccine container (i.e., primary container) can have many implications on the safety and convenience of a vaccination session, another important but potentially overlooked consideration is how the design of the primary container may affect the distribution of the vaccine, its resulting cost, and whether the vial is ultimately opened. Using our HERMES software platform, we developed a simulation model of the World Health Organization Expanded Program on Immunization supply chain for the Republic of Benin and used the model to explore the effects of different primary containers for various vaccine antigens. Replacing vaccines with presentations containing fewer doses per vial reduced vaccine availability (proportion of people arriving for vaccines who are successfully immunized) by as much as 13% (from 73% at baseline) and raised logistics costs by up to $0.06 per dose administered (from $0.25 at baseline) due to increased bottlenecks, while reducing total costs by as much as $0.15 per dose administered (from $2.52 at baseline) due to lower open vial wastage. Primary containers with a greater number of doses per vial each improved vaccine availability by 19% and reduced logistics costs by $0.05 per dose administered, while reducing the total costs by up to $0.25 per dose administered. Changes in supply chain performance were more extreme in departments with greater constraints. Implementing a vial opening threshold reversed the direction of many of these effects. Our results show that one size may not fit all when choosing a primary vaccine container. Rather, the choice depends on characteristics of the vaccine, the vaccine supply chain, immunization session size, and goals of decision makers. In fact, the optimal vial size may vary among locations within a country. Simulation modeling can help identify tailored approaches to improve availability and efficiency. Copyright © 2015 Elsevier Ltd. All rights reserved.

  20. Castor oil polyurethane as a coating option for spent nuclear fuel disposal containment

    Energy Technology Data Exchange (ETDEWEB)

    Mortley, A.; Bonin, H.W.; Bui, V.T. [Dept. of Chemistry and Chemical Engineering, Royal Military College of Canada, Kingston, Ontario (Canada)

    2009-07-01

    Castor oil polyurethane (COPU) coatings are being proposed as an additional barrier in the design of the copper containers to store spent nuclear fuel in Canada. The present work investigates the variation in the physicomechanical properties of two COPUs, based on an aliphatic and aromatic diisocyanate, as a function of ionizing radiation dose and dose rate. The changes in physicomechanical properties have shown that radiation, regardless of dose rate and isocyanate structure, increases the values of the modulus and the ultimate tensile strength when compared with those of the unirradiated samples, with aromatic based polyurethanes being more susceptible to variation than aliphatic based ones. (author)

  1. System for the sealing of containers and pipelines, especially in nuclear power plants

    International Nuclear Information System (INIS)

    Gross, R.

    1976-01-01

    In order to seal containers and pipelines especially in nuclear power plants, it is suggested to incorporate hollow bodies of an elastic material in the pipeline connections which can be blown up by pressure-gas and which are placed for sealing on the inner walls of the pipe. During a longer shutdown, system parts can thus be protected from corrosion. Various forms of such cavities are shown in design examples. The sealed sections can be filled with inert gas (nitrogen). (RW/LH) [de

  2. Device for increasing the safety in the environment of nuclear facilities in case of containment failure

    International Nuclear Information System (INIS)

    Morlock, G.; Wiesemes, J.; Bachner, D.

    1978-01-01

    In order to increase the safety in the environment of nuclear facilities, e.g. in case of containment failure, with respect to released radioactive material new or existing facilities are covered with ground. The ground material has got a consistency very much reducing the permeability for liquids and gases. In addition irrigation devices for keeping the ground wet and/or intermediate layers of films pervious to water, e.g. perforated sheets, may be provided. Additionally the ground is protected against frost. Especially suited for ground material is clay. (DG) [de

  3. Castor oil polyurethane as a coating option for spent nuclear fuel disposal containment

    International Nuclear Information System (INIS)

    Mortley, A.; Bonin, H.W.; Bui, V.T.

    2009-01-01

    Castor oil polyurethane (COPU) coatings are being proposed as an additional barrier in the design of the copper containers to store spent nuclear fuel in Canada. The present work investigates the variation in the physicomechanical properties of two COPUs, based on an aliphatic and aromatic diisocyanate, as a function of ionizing radiation dose and dose rate. The changes in physicomechanical properties have shown that radiation, regardless of dose rate and isocyanate structure, increases the values of the modulus and the ultimate tensile strength when compared with those of the unirradiated samples, with aromatic based polyurethanes being more susceptible to variation than aliphatic based ones. (author)

  4. 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.

  5. 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.

  6. Heat, mass, and momentum transport model for hydrogen diffusion flames in nuclear reactor containments

    International Nuclear Information System (INIS)

    Travis, J.R.

    1985-01-01

    It is now possible to analyze the time-dependent, fully three-dimensional behavior of hydrogen diffusion flames in nuclear reactor containments. This analysis involves coupling the full Navier-Stokes equations with multi-species transport to the global chemical kinetics of hydrogen combustion. A transport equation for the subgrid scale turbulent kinetic energy density is solved to produce the time and space dependent turbulent transport coefficients. The heat transfer coefficient governing the exchange of heat between fluid computational cells adjacent to wall cells is calculated by a modified Reynolds analogy formulation. The analysis of a MARK-III containment indicates very complex flow patterns that greatly influence fluid and wall temperatures and heat fluxes. 18 refs., 24 figs

  7. Assessment of Effective Prestressed Force of Nuclear Containment Building using SI Technique

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. P.; Jang, J. B.; Hwang, K. M.; Song, Y. C. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Bonded tendons have been used in reactor buildings of heavy water reactors and the light water reactors of some nuclear power plants operating in Korea. The assessment of prestressed forces on those bonded tendons is becoming an important issue in assuring their continuous operation beyond their design life. In order to assess the effective prestressed force on the bonded tendon, indirect assessment techniques have been applying to the test beams which were manufactured on construction time. Therefore, this research mainly forced to establish the assessment methodology to measure directly the effective prestressed force on the bonded tendon of containment buildings using System Identification (SI) technique. To accomplish this purpose, simple SI method was proposed and adapted three dimensional finite element analysis of the 1:4 scale prestressed concrete containment vessel (PCCV) tested by Sandia National Laboratory in 2000

  8. Tritium concentration reducing method in atmosphere in nuclear reactor containment facility

    International Nuclear Information System (INIS)

    Hirasawa, Yoshiya; Kigoshi, Yasutane; Yonenaga, Haruo.

    1992-01-01

    A portion of water content in an atmosphere is condensed by a condensation/evaporation device disposed in a nuclear reactor containment building and then a portion of the condensed water is evaporated in the atmosphere. A portion of hydrogen nuclides constituting the evaporated water content is subjected to isotopic exchange with tritium nuclides in the atmosphere. A portion of water content in the atmosphere applied with the isotopic exchange is condensed in the condensation/evaporation device. That is, the hydrogen nuclides in steams are applied with isotopic exchange with tritium nuclides, and steams incorporating tritium nuclides are condensed again in the condensation/evaporation device, to transfer the tritium nuclides in the atmosphere to condensed water. The condensed water is recovered without releasing the tritium nuclides to the outside of the reactor containment building, thereby enabling to reduce the tritium concentration in the atmosphere. (N.H.)

  9. HMS-burn: a model for hydrogen distribution and combustion in nuclear reactor containments

    International Nuclear Information System (INIS)

    Travis, J.R.

    1985-01-01

    It is now possible to analyze the time-dependent, fully three-dimensional behavior of hydrogen combustion in nuclear reactor containments. This analysis involves coupling the full Navier-Stokes equations with multi-species transport to the global chemical kinetics of hydrogen combustion. A transport equation for the subgrid scale turbulent kinetic energy density is solved to produce the time and space dependent turbulent transport coefficients. The heat transfer coefficient governing the exchange of heat between fluid computational cells adjacent to wall cells is calculated by a modified Reynolds analogy formulation. The analysis of a MARK-III containment indicates very complex flow patterns that greatly influence fluid and wall temperatures and heat fluxes

  10. 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

  11. Device for manufacturing methane or synthetic gas from materials containing carbon using a nuclear reactor

    International Nuclear Information System (INIS)

    Jaeger, W.

    1984-01-01

    This invention concerns a device for manufacturing methane or synthetic gas from materials containing carbon using a nuclear reactor, where part of the carbon is gasified with hydration and the remaining carbon is converted to synthetic gas by adding steam. This synthetic gas consists mainly of H 2 , CO, CO 2 and CH 4 and can be converted to methane in so-called methanising using a nickel catalyst. The hydrogen gasifier is situated in the first of two helium circuits of a high temperature reactor, and the splitting furnace is situated in the second helium circuit, where part of the methane produced is split into hydrogen at high temperature, which is used for the hydrating splitting of another part of the material containing carbon. (orig./RB) [de

  12. 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

  13. 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

  14. Passive containment system

    International Nuclear Information System (INIS)

    Kleimola, F.W.

    1977-01-01

    Disclosed is a containment system that provides complete protection entirely by passive means for the loss of coolant accident in a nuclear power plant and wherein all stored energy released in the coolant blowdown is contained and absorbed while the nuclear fuel is prevented from over-heating by a high containment back-pressure and a reactor vessel refill system. The primary containment vessel is restored to a high sub-atmospheric pressure within a few minutes after accident initiation and the decay heat is safely transferred to the environment while radiolytic hydrogen is contained by passive means. 20 claims, 14 figures

  15. Primary circuit contamination in nuclear power plants: contribution to occupational exposure

    International Nuclear Information System (INIS)

    Provens, H.

    2002-01-01

    In every country since the 80's, a clear downward trend is observed concerning the occupational doses at nuclear power plants, as shows the regularly decreasing annual collective dose per operating reactor. Even if technology and work management are improving, the reduction and the control of radiation sources remain one critical point. This paper summarizes the results of an extended study on the primary circuit contamination in nuclear power plants and its contribution to workers' exposure. The paper reviews the origin and mechanisms of radiation production and the different ways of radiation control or reduction based on physical and chemical parameters and not organisational or human factors. It underlines that chemistry control of the primary circuit is one essential component of radiation protection optimisation in nuclear power plants. Results reported come from scientific data in open literature and cannot be generalized to all the power plants

  16. Microbially influenced corrosion of copper nuclear fuel waste containers in a Canadian disposal vault

    International Nuclear Information System (INIS)

    King, F.

    1996-11-01

    An assessment of the potential for microbially influenced corrosion (MIC) of copper nuclear fuel waste containers in a Canadian disposal vault is presented. The assessment is based on a consideration of the microbial activity within a disposal vault, the reported cases of MIC of Cu alloys in the literature and the known corrosion behaviour of Cu. Because of the critical role of biofilms in the reported cases of MIC, their formation and properties are discussed in detail. Next, the literature on the MIC of Cu alloys is briefly reviewed. The various MIC mechanisms proposed are critically discussed and the implications for the corrosion of Cu containers considered. In the majority of literature cases, MIC depends on alternating aerated and deaerated environments, with accelerated corrosion being observed when fresh aerated water replaces stagnant water, e.g., the MIC of Cu-Ni heat exchangers in polluted seawater and the microbially influenced pitting of Cu water pipes. Finally, because of the predominance of corrosion by sulphate-reducing bacteria (SRB) in the MIC literature, the abiotic behaviour of Cu alloys in sulphide solutions is also reviewed. The effect of the evolving environment in a disposal vault on the extent and location of microbial activity is discussed. Biofilm formation on the container surface is considered unlikely throughout the container lifetime, but especially initially when the environmental conditions will be particularly aggressive. Microbial activity in areas of the vault away from the container is possible, however. Corrosion of the container could then occur if microbial metabolic by-products diffuse to the container surface. Sulphide, produced by the action of SRB are considered to be the most likely cause of container corrosion. It is concluded that the only likely form of MIC of Cu containers will result from sulphide produced by SRB diffusing to the container surface. A modelling procedure for predicting the extent of corrosion is

  17. Vitrification of high level nuclear waste inside ambient temperature disposal containers using inductive heating: The SMILE system

    International Nuclear Information System (INIS)

    Powell, J.; Reich, M.; Barletta, R.

    1996-01-01

    A new approach, termed SMILE (Small Module Inductively Loaded Energy), for the vitrification of high level nuclear wastes (HLW) is described. Present vitrification systems liquefy the HLW solids and associated frit material in large high temperature melters. The molten mix is then poured into small (∼1 m 3 ) disposal canisters, where it solidifies and cools. SMILE eliminates the separate, large high temperature melter. Instead, the BLW solids and frit melt inside the final disposal containers, using inductive heating. The contents then solidify and cool in place. The SMILE modules and the inductive heating process are designed so that the outer stainless can of the module remains at near ambient temperature during the process cycle. Module dimensions are similar to those of present disposal containers. The can is thermally insulated from the high temperature inner container by a thin layer of refractory alumina firebricks. The inner container is a graphite crucible lined with a dense alumina refractory that holds the HLW and fiit materials. After the SMILE module is loaded with a slurry of HLW and frit solids, an external multi-turn coil is energized with 30-cycle AC current. The enclosing external coil is the primary of a power transformer, with the graphite crucible acting as a single turn ''secondary.'' The induced current in the ''secondary'' heats the graphite, which in turn heats the HLW and frit materials. The first stage of the heating process is carried out at an intermediate temperature to drive off remnant liquid water and water of hydration, which takes about 1 day. The small fill/vent tube to the module is then sealed off and the interior temperature raised to the vitrification range, i.e., ∼1200C. Liquefaction is complete after approximately 1 day. The inductive heating then ceases and the module slowly loses heat to the environment, allowing the molten material to solidify and cool down to ambient temperature

  18. The COLIMA experiment on aerosol retention in containment leak paths under severe nuclear accidents

    Energy Technology Data Exchange (ETDEWEB)

    Parozzi, Flavio, E-mail: flavio.parozzi@rse-web.it [RSE, Power Generation Department, via Rubattino 54, I-20134 Milano (Italy); Caracciolo, Eduardo D.J., E-mail: eduardo.caracciolo@rse-web.it [RSE, Power Generation Department, via Rubattino 54, I-20134 Milano (Italy); Journeau, Christophe, E-mail: christophe.journeau@cea.fr [CEA Cadarache (France); Piluso, Pascal, E-mail: pascal.piluso@cea.fr [CEA Cadarache (France)

    2013-08-15

    Highlights: ► Experiment investigating aerosol retention within concrete containment cracks under nuclear severe accident conditions. ► Provided representative conditions of the aerosols suspended inside the containment of PWRs under a severe accident. ► Prototypical aerosol particles generated with a thermite reaction and transported through the crack sample reproducing surface characteristics, temperature, pressure drop and gas leakage. ► The results indicate the significant retention due to zig-zag path. -- Abstract: CEA and RSE managed an experimental research concerning the investigation of aerosol retention within concrete containment cracks under severe accident conditions. The main experiment was carried out in November 2008 with aerosol generated from the COLIMA facility and a sample of cracked concrete with defined geometric characteristics manufactured by RSE. The facility provided representative conditions of the aerosols suspended inside the containment of PWRs under a severe accident. Prototypical aerosol particles were generated with a thermite reaction and transported through the crack sample, where surface characteristics, temperature, pressure drop and gas leakage were properly reproduced. The paper describes the approach adopted for the preparation of the cracked concrete sample and the dimensioning of the experimental apparatus, the test procedure and the measured parameters. The preliminary results, obtained from this single test, are also discussed in the light of the present knowledge about aerosol phenomena and the theoretical analyses of particle behaviour with the crack path.

  19. Passive pH adjustment of nuclear reactor containment flood water

    International Nuclear Information System (INIS)

    Gerlowski, T.J.

    1986-01-01

    A method is described of automatically and passively adjusting the pH of the recirculating liquid used to flood the containment structure of a nuclear reactor upon the occurence of an accident in order to cool the reactor core, wherein the containment structure has a concrete floor which is provided with at least one sump from which the liquid is withdrawn for recirculation via at least one outlet pipe. The method consists of: prior to flooding and during or prior to normal operation of the reactor, providing at least one perforated basket within at least one sump with the basket containing crystals of a pH adjusting chemical which is soluble in the liquid, and covering each basket with a plastic coating which is likewise soluble in the liquid, whereby upon flooding of the containment structure the liquid in the sump will reach the level of the baskets, causing the coating and the crystals to be dissolved and the chemical to mix with the recirculating liquid to adjust the pH

  20. Predicting the effects of microbial activity on the corrosion of copper nuclear fuel waste disposal containers

    International Nuclear Information System (INIS)

    King, F.; Stroes-Gascoyne, S.

    1996-08-01

    Microbially influenced corrosion (MIC) of copper nuclear fuel waste containers may occur in a disposal vault located 500-1000 m underground in the granitic rock of the Canadian Shield. The extent and diversity of microbial activity in the vault is expected to be limited initially because of the aggressive conditions produced by γ-radiation, elevated temperatures and desiccation of the clay-based buffer in which the containers will be embedded. Experimental results on the heat- and radiation-sensitivity of the natural microbiota in buffer material are presented. The data suggest that the low water activity in the buffer material will severely limit the growth of microbes near the container. The most likely form of MIC involves sulphate-reducing bacteria (SRB). Electrochemical experiments using a clay-covered copper electrode have shown that sulphide ions produced by SRB could diffuse through buffer material and induce corrosion of the container. A method to predict the long-term corrosion behaviour is presented. (author)

  1. Severe accident consequence mitigation by filtered containment venting at Canadian nuclear power plants

    International Nuclear Information System (INIS)

    Lebel, Luke S.; Morreale, Andrew C.; Korolevych, Volodymyr; Brown, Morgan J.; Gyepi-Garbrah, Sam

    2017-01-01

    Highlights: • Use of filtered containment venting during a severe accident assessed. • Severe accident simulations performed using MAAP-CANDU and ADDAM. • Flow capacity, initiation protocols, efficiency, mass and thermal loading evaluated. • Efficient, robust system drastically reduces accident consequences. - Abstract: Having the capability to use filtered containment venting during a severe nuclear accident can significantly reduce its overall consequences. This study employs the MAAP-CANDU severe accident analysis code and the ADDAM atmospheric dispersion code to study the progression of: an unmitigated station blackout accident at a generic pressurized heavy water reactor, the release of radioactive material into the environment, the subsequent dispersion of the fission products through the atmosphere and the subsequent consequences (evacuation radius). The goal is to evaluate the application of filtered venting as an accident mitigation technology. Several aspects of filtered containment venting system design, like flow capacity, initiation protocols, filter efficiency, mass loading, and thermal loading are considered. An efficient and robust filtered containment venting system can reduce the amount of radiological materials emitted during an accident by 25 times or more, and as a result considerably reduce the off-site consequences of an accident.

  2. Application of fire-retardant treatment to the wood in Type A unirradiated nuclear fuel outer containers

    International Nuclear Information System (INIS)

    Whitlow, J.D.; Luna, R.E.

    1992-01-01

    Packagings for transporting unirradiated nuclear fuel assemblies in the United States are commonly constructed as rectangular boxes consisting of a metal inner container, a wooden outer container, and cushioning material separating the two. The wood in the outer container is a potential source of fuel for fire. Use of a fire-retardant treatment on the wood may reduce or eliminate the damage to nuclear fuel assemblies in some types of accidents involving fire. The applicability of using fire-retardant treatments on the wood of outer containers is addressed. An approximate cost-benefit analysis to determine if fire-retardant treatments are economically justified is presented. (Author)

  3. Fabrication development for high-level nuclear waste containers for the tuff repository; Phase 1 final report

    Energy Technology Data Exchange (ETDEWEB)

    Domian, H.A.; Holbrook, R.L.; LaCount, D.F. [Babcock and Wilcox Co., Lynchburg, VA (USA). Nuclear Power Div.]|[Babcock and Wilcox Co., Alliance, OH (USA). Research and Development Div.

    1990-09-01

    This final report completes Phase 1 of an engineering study of potential manufacturing processes for the fabrication of containers for the long-term storage of nuclear waste. An extensive literature and industry review was conducted to identify and characterize various processes. A technical specification was prepared using the American Society of Mechanical Engineers Boiler & Pressure Vessel Code (ASME BPVC) to develop the requirements. A complex weighting and evaluation system was devised as a preliminary method to assess the processes. The system takes into account the likelihood and severity of each possible failure mechanism in service and the effects of various processes on the microstructural features. It is concluded that an integral, seamless lower unit of the container made by back extrusion has potential performance advantages but is also very high in cost. A welded construction offers lower cost and may be adequate for the application. Recommendations are made for the processes to be further evaluated in the next phase when mock-up trials will be conducted to address key concerns with various processes and materials before selecting a primary manufacturing process. 43 refs., 26 figs., 34 tabs.

  4. Nuclear power plant Olkiluoto 3. Containment leakage test under extreme conditions

    Energy Technology Data Exchange (ETDEWEB)

    Fleckenstein, Tobias [TUEV SUED Industrie Service GmbH, Munich (Germany). Measaruement Technology Dept.

    2015-01-15

    Modern nuclear power plants place high demands on the design and execution of safety checks. TUEV SUED supported the containment leakage test for the largest- capacity third generation nuclear power plant in the world - Olkiluoto 3 in Finland. The experts successfully met the challenges presented by exceptional parameters of the project. The containment of Olkiluoto 3 is unique in that the vessel's volume is 80,000 m{sup 3} while measurements were carried out over a period of ten days. To execute the test, 75 temperature and 15 humidity sensors had to be installed and correctly interlinked by more than ten kilometres of cable. These instruments also needed to withstand an absolute pressure of 6 bar, ambient temperatures of 30 C and high levels of humidity. These conditions required comprehensive preparation and a high amount of qualification tests. Parts of the qualifications were carried out at the autoclave system of the Technical University in Munich, Germany, where the project test conditions could be simulated. The software required to determine the tests was developed by TUEV SUED and verified by German's national accreditation body DAkkS under ISO 17025. TUEV SUED enabled the test schedule to continue without delay by analysing all recorded data continuously on site, including pressure, temperature, humidity and leakage mass flow curves. With the comprehensive preparation, data acquisition system recording measurements continuously and the on-time result calculation, all components of the leak-tightness assessment were successfully completed in accordance with requirements.

  5. 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.

  6. 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.

  7. A model for predicting the lifetimes of Grade-2 titanium nuclear waste containers

    Energy Technology Data Exchange (ETDEWEB)

    Shoesmith, D W; Ikeda, B M; Bailey, M G; Quinn, M J; LeNeveu, D M [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

    1995-08-01

    The development of a model to predict the lifetimes of Grade-2 titanium containers for nuclear fuel waste is described. This model assumes that the corrosion processes most likely to lead to container failure are crevice corrosion, hydrogen-induced cracking and general corrosion. Because of the expected evolution of waste vault conditions from initially warm (<{approx} 100 deg C) and oxidizing to eventually cool (<30 deg C) and non-oxiding, the period for which crevice corrosion can propagate will be limited by repassivation, and long container lifetimes will be achieved since the rate of general corrosion is extremely low. However, in the model presented, not only is it assumed that crevices will initiate rapidly on all containers, but also that the propagation of these crevices will continue indefinitely since conditions will remain sufficiently oxiding for repassivation to be avoided. The mathematical development of the model is described in detail. A simple ramped distribution is used to describe the failures due to the presence of initial defects. For crevice corrosion the propagation rates are assumed to be normally distributed and to be determined predominantly by temperature. The temperature dependence of the crevice propagation rate is determined from the calculated cooling profiles for the containers and an experimentally determined Arrhenius relationship for crevice propagation rates. The cooling profiles are approximated by double or single step functions, depending on the location of the container within the vault. The experimental data upon which this model is based is extensively reviewed. This review includes descriptions of the available data to describe and quantify the processes of general corrosion, crevice corrosion and hydrogen-induced cracking. For crevice corrosion and hydrogen-induced cracking the results of studies on both Grades-2 and -12 are presented. Also, the effects of impurities in the Grade-2 material are discussed.

  8. A model for predicting the lifetimes of Grade-2 titanium nuclear waste containers

    International Nuclear Information System (INIS)

    Shoesmith, D.W.; Ikeda, B.M.; Bailey, M.G.; Quinn, M.J.; LeNeveu, D.M.

    1995-08-01

    The development of a model to predict the lifetimes of Grade-2 titanium containers for nuclear fuel waste is described. This model assumes that the corrosion processes most likely to lead to container failure are crevice corrosion, hydrogen-induced cracking and general corrosion. Because of the expected evolution of waste vault conditions from initially warm (<∼ 100 deg C) and oxidizing to eventually cool (<30 deg C) and non-oxiding, the period for which crevice corrosion can propagate will be limited by repassivation, and long container lifetimes will be achieved since the rate of general corrosion is extremely low. However, in the model presented, not only is it assumed that crevices will initiate rapidly on all containers, but also that the propagation of these crevices will continue indefinitely since conditions will remain sufficiently oxiding for repassivation to be avoided. The mathematical development of the model is described in detail. A simple ramped distribution is used to describe the failures due to the presence of initial defects. For crevice corrosion the propagation rates are assumed to be normally distributed and to be determined predominantly by temperature. The temperature dependence of the crevice propagation rate is determined from the calculated cooling profiles for the containers and an experimentally determined Arrhenius relationship for crevice propagation rates. The cooling profiles are approximated by double or single step functions, depending on the location of the container within the vault. The experimental data upon which this model is based is extensively reviewed. This review includes descriptions of the available data to describe and quantify the processes of general corrosion, crevice corrosion and hydrogen-induced cracking. For crevice corrosion and hydrogen-induced cracking the results of studies on both Grades-2 and -12 are presented. Also, the effects of impurities in the Grade-2 material are discussed. Special attention is

  9. Lead-iron phosphate glass as a containment medium for the disposal of high-level nuclear wastes

    Science.gov (United States)

    Boatner, L.A.; Sales, B.C.

    1984-04-11

    Disclosed are lead-iron phosphate glasses containing a high level of Fe/sub 2/O/sub 3/ for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste

  10. Problems of interaction between water and fuel containing masses inside the object 'Shelter' of Chernobyl nuclear power plant

    International Nuclear Information System (INIS)

    Yukhnovs'kij, Yi.R.; Kobrin, O.Je.; Tokarchuk, M.V.; Tokarevs'kij, V.V.

    1997-01-01

    The main forms of the existence of nuclear fuel and major concomitant factors of nuclear and ecological danger of the object 'Shelter' are presented. The processes of interaction between water and fuel containing materials have been analysed on the basis of experimental data

  11. Problems of the processing of nuclear magnetic logging signals (identification of fluid-containing strata from a number of measurements)

    International Nuclear Information System (INIS)

    Aliev, T.M.; Orlov, G.L.; Lof, V.M.; Mityushin, E.M.; Ragimova, E.K.

    1978-01-01

    Problems of the processing of nuclear magnetic logging signals to identification of fluid-containing strata from a number of measurements. Problems of application statistical decision theory to discovery of fluid-containing beds from a number of measurements are considered. Using the technique possibilities of nuclear magnetic logging method the necessary volume of samples is motivated, the rational algorithm for processing of sequential measurements is obtained

  12. Numerical indices for quantification of hydrogen mixing and deflagration potential in the nuclear reactor containment

    Energy Technology Data Exchange (ETDEWEB)

    Agrawal, Nilesh, E-mail: nilesh_agrawal@igcar.gov.in [Safety Research Institute, Atomic Energy Regulatory Board, Government of India, Kalpakkam (India); Das, Sarit K. [Department of Mechanical Engineering, Indian Institute of Technology, Madras, Chennai 600036 (India)

    2013-06-15

    Highlights: • Four indices for mixing and deflagration potential of H{sub 2} distributions are presented. • Theoretical basis and significance are explained through illustrations. • The influence of steam condensation on H{sub 2} distributions is studied numerically. • The indices depict the rate of mixing and changes in deflagration potential. • Results show that the indices can give useful integral information for comparison. -- Abstract: Studies on hydrogen distribution in the nuclear reactor containment and the effect of a distribution on subsequent combustion are important to nuclear safety. Contour plots, concentration profiles and ternary diagrams are routinely used to represent a distribution. The significance to safety has to be qualitatively inferred from these representations. Thus, there is a need to quantify distributions in terms of gross parameters that are important to safety. In the present study, four numerical indices are developed to obtain quantitative information on the mixing and deflagration potential of a distribution of hydrogen, steam and air. Two indices, namely, Average mole fraction and Non-uniformity index can be used to give the state of mixing of hydrogen in an enclosure at any instant of time. Similarly, the other two indices, namely, deflagration volume fraction and deflagration pressure ratio can be used to indicate the relative size of combustible cloud and the expected pressure rise in case of deflagration in the cloud at any instant of time. The significance and utility of the indices are brought forth through simple illustrations and numerical studies on the influence of steam condensation on hydrogen distributions. The indices depict the rate of mixing and changes in deflagration potential for the situations considered. Results form the simple studies show that the indices can give useful integral information for comparison of mixing and mitigation measures deployed in the nuclear reactor containment.

  13. Damage evaluation of 500 MWe Indian pressurized heavy water reactor nuclear containment for air craft impact

    International Nuclear Information System (INIS)

    Kukreja, Mukesh; Singh, R.K; Vaze, K.K; Kushwaha, H.S.

    2003-01-01

    Non-linear transient dynamic analysis of 500 MWe Indian Pressurized Heavy Water Reactor (PHWR) nuclear containment has been carried out for the impact of Boeing and Airbus category of aircraft operated in India. The impulsive load time history is generated based on the momentum transfer of the crushable aircraft (soft missiles) of Boeing and Airbus families on the containment structure. The case studies include the analyses of outer containment wall (OCW) single model and the combined model with outer and inner containment wall (ICW) for impulsive loading due to aircraft impact. Initially the load is applied on OCW single model and subsequently the load is transferred to ICW after the local perforation of the OCW is noticed in the transient simulation. In the first stage of the analysis it is demonstrated that the OCW would suffer local perforation with a peak local deformation of 117 mm for impact due to B707-320 and 196 mm due to impact of A300B4 without loss of the overall integrity. However, this first barrier (OCW) cannot absorb the full impulsive load. In the second stage of the analysis of the combined model, the ICW is subjected to lower impulse duration as the load is transferred after 0.19 sec for B707-320 and 0.24 sec for A300B4 due to the local perforation of OCW. This results in the local deformation of approx. 115 mm for B707-320 and 124 mm for A300B4 in ICW and together both the structures (OCW and ICW) are capable of absorbing the full impulsive load. The analysis methodology evolved in the present work would be useful for studying the behaviour of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due commercial aircraft operated in India. (author)

  14. 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)

  15. 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

  16. Pressure and temperature analyses using GOTHIC for Mark I containment of the Chinshan Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Yen-Shu, E-mail: yschen@iner.org.t [Nuclear Engineering Division, Institute of Nuclear Energy Research, 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Yuann, Yng-Ruey; Dai, Liang-Che; Lin, Yon-Pon [Nuclear Engineering Division, Institute of Nuclear Energy Research, 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China)

    2011-05-15

    Research highlights: The Chinshan Mark I containment pressure-temperature responses are analyzed. GOTHIC is used to calculate the containment responses under three pipe break events. This study is used to support the Chinshan Stretch Power Uprate (SPU) program. The calculated peak pressure and temperature are still below the design values. The Chinshan containment integrity can be maintained under SPU condition. - Abstract: Chinshan Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/4 plant with original licensed thermal power (OLTP) of 1775 MWt for each unit. Recently, the Stretch Power Uprate (SPU) program for the Chinshan plant is being conducted to uprate the core thermal power to 1858 MWt (104.66% OLTP). In this study, the Chinshan Mark I containment pressure/temperature responses during LOCA at 105% OLTP (104.66% OLTP + 0.34% OLTP power uncertainty = 105% OLTP) are analyzed using the containment thermal-hydraulic program GOTHIC. Three kinds of LOCA (Loss of Coolant Accident) scenarios are investigated: Recirculation Line Break (RCLB), Main Steam Line Break (MSLB), and Feedwater Line Break (FWLB). In the short-term analyses, blowdown data generated by RELAP5 transient analyses are provided as boundary conditions to the GOTHIC containment model. The calculated peak drywell pressure and temperature in the RCLB event are 217.2 kPaG and 137.1 {sup o}C, respectively, which are close to the original FSAR results (219.2 kPaG and 138.4 {sup o}C). Additionally, the peak drywell temperature of 155.3 {sup o}C calculated by MSLB is presented in this study. To obtain the peak suppression pool temperature, a long-term RCLB analysis is performed using a simplified RPV (Reactor Pressure Vessel) volume to calculate blowdown flow rate. One RHR (Residual Heat Removal) heat exchanger is assumed to be inoperable for suppression pool cooling mode. The calculated peak suppression pool temperature is 93.2 {sup o}C, which is below the pool temperature used for evaluating the

  17. Pressure and temperature analyses using GOTHIC for Mark I containment of the Chinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Chen, Yen-Shu; Yuann, Yng-Ruey; Dai, Liang-Che; Lin, Yon-Pon

    2011-01-01

    Research highlights: → The Chinshan Mark I containment pressure-temperature responses are analyzed. → GOTHIC is used to calculate the containment responses under three pipe break events. → This study is used to support the Chinshan Stretch Power Uprate (SPU) program. → The calculated peak pressure and temperature are still below the design values. → The Chinshan containment integrity can be maintained under SPU condition. - Abstract: Chinshan Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/4 plant with original licensed thermal power (OLTP) of 1775 MWt for each unit. Recently, the Stretch Power Uprate (SPU) program for the Chinshan plant is being conducted to uprate the core thermal power to 1858 MWt (104.66% OLTP). In this study, the Chinshan Mark I containment pressure/temperature responses during LOCA at 105% OLTP (104.66% OLTP + 0.34% OLTP power uncertainty = 105% OLTP) are analyzed using the containment thermal-hydraulic program GOTHIC. Three kinds of LOCA (Loss of Coolant Accident) scenarios are investigated: Recirculation Line Break (RCLB), Main Steam Line Break (MSLB), and Feedwater Line Break (FWLB). In the short-term analyses, blowdown data generated by RELAP5 transient analyses are provided as boundary conditions to the GOTHIC containment model. The calculated peak drywell pressure and temperature in the RCLB event are 217.2 kPaG and 137.1 o C, respectively, which are close to the original FSAR results (219.2 kPaG and 138.4 o C). Additionally, the peak drywell temperature of 155.3 o C calculated by MSLB is presented in this study. To obtain the peak suppression pool temperature, a long-term RCLB analysis is performed using a simplified RPV (Reactor Pressure Vessel) volume to calculate blowdown flow rate. One RHR (Residual Heat Removal) heat exchanger is assumed to be inoperable for suppression pool cooling mode. The calculated peak suppression pool temperature is 93.2 o C, which is below the pool temperature used for evaluating the

  18. Development of a method for detecting nuclear fuel debris and water leaks at a nuclear reactor/containment vessel by flow visualization

    International Nuclear Information System (INIS)

    Umezawa, Shuichi; Tanaka, Katsuhiko

    2013-01-01

    It is the important issue to fill up each nuclear reactor/containment vessel with water and to take out debris of damaged fuel from them for decommissioning of Fukushima Daiichi nuclear power plants. It is necessary to detect the debris and water leaks at a nuclear reactor/containment vessel for the purpose. However, the method is not completely developed in the present stage. Accordingly, we have developed a method for detecting debris and water leaks at a nuclear reactor/containment vessel by flow visualization. Experiments of the flow visualization were conducted using two types of water tanks. An optical fiber and a collimator lens were employed for modifying a straight laser beam into a sheet projection. Some visualized images were obtained through the experiments. Particle Image Velocimetry, i.e. PIV, analysis was applied to the images for quantitative flow rate analysis. Consequently, it is considered that the flow visualization method has a possibility for the practical use. (author)

  19. Detecting special nuclear materials in suspect containers using high-energy gamma rays emitted by fission products

    Science.gov (United States)

    Norman, Eric B [Oakland, CA; Prussin, Stanley G [Kensington, CA

    2009-01-27

    A method and a system for detecting the presence of special nuclear materials in a suspect container. The system and its method include irradiating the suspect container with a beam of neutrons, so as to induce a thermal fission in a portion of the special nuclear materials, detecting the gamma rays that are emitted from the fission products formed by the thermal fission, to produce a detector signal, comparing the detector signal with a threshold value to form a comparison, and detecting the presence of the special nuclear materials using the comparison.

  20. Survey of Regulations Applicable to the Finned Containment in Korean Nuclear Power Plant for Light Water Reactor

    International Nuclear Information System (INIS)

    Noh, Hyung Gyun; Kang, Hie Chan

    2016-01-01

    In severe accident, the molten corium would discharge into the reactor cavity and interact with water and concrete of cavity. Molten corium includes non-oxidation metals such as Zr, Fe and Cr. These metal species reacted with water emit hydrogen gas. In addition to this, a mount of steam can be emitted to the containment such as steam line break accident. As a result, steam and hydrogen gas can pressurize containment over the design pressure and threaten its integrity. For this reasons, a concept equipped with finned on the containment building was proposed for coping with prolonged accident. Finned containment can enhance heat transfer to the ambient, and the building itself is working as a heat sink. Multiple metal fins and metal rod are penetrated into containment wall, and the rods are working as an additional path of heat removal. To be accepted in the nuclear power plants, this configuration should satisfy the requirement of heat removal and follow all regulations related with containment also. For applying to Korean nuclear power plants, the finned containment should follow all regulations specialized in Korea such as Nuclear regulatory criteria for light water reactor and Guidelines of nuclear safety examination for light water reactor. A concept of containment as a passive cooling system has been proposed. Furthermore, the new containment concept can be applied on the real containment which satisfies the various regulations. Finned containment would be expected positive effects on heat removal from the containment. If the fins are properly welded to the liner, finned containment could satisfy the leak tightness and prevention of external influences. Finned containment could be favorable to protect external impact like aircraft crash because of the additional structural integrity by the fins

  1. Survey of Regulations Applicable to the Finned Containment in Korean Nuclear Power Plant for Light Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Hyung Gyun [Pohang University, Pohang (Korea, Republic of); Kang, Hie Chan [Kunsan University, Gunsan (Korea, Republic of)

    2016-05-15

    In severe accident, the molten corium would discharge into the reactor cavity and interact with water and concrete of cavity. Molten corium includes non-oxidation metals such as Zr, Fe and Cr. These metal species reacted with water emit hydrogen gas. In addition to this, a mount of steam can be emitted to the containment such as steam line break accident. As a result, steam and hydrogen gas can pressurize containment over the design pressure and threaten its integrity. For this reasons, a concept equipped with finned on the containment building was proposed for coping with prolonged accident. Finned containment can enhance heat transfer to the ambient, and the building itself is working as a heat sink. Multiple metal fins and metal rod are penetrated into containment wall, and the rods are working as an additional path of heat removal. To be accepted in the nuclear power plants, this configuration should satisfy the requirement of heat removal and follow all regulations related with containment also. For applying to Korean nuclear power plants, the finned containment should follow all regulations specialized in Korea such as Nuclear regulatory criteria for light water reactor and Guidelines of nuclear safety examination for light water reactor. A concept of containment as a passive cooling system has been proposed. Furthermore, the new containment concept can be applied on the real containment which satisfies the various regulations. Finned containment would be expected positive effects on heat removal from the containment. If the fins are properly welded to the liner, finned containment could satisfy the leak tightness and prevention of external influences. Finned containment could be favorable to protect external impact like aircraft crash because of the additional structural integrity by the fins.

  2. Development of web-based integrity evaluation system for primary components in a nuclear power plant

    International Nuclear Information System (INIS)

    Lee, S.M.; Kim, J.C.; Choi, J.B.; Kim, Y.J.; Choi, S.N.; Jang, K.S.; Hong, S.Y.

    2004-01-01

    A nuclear power plant is composed of a number of primary components. Maintaining the integrity of these components is one of the most critical issues in nuclear industry. In order to maintain the integrity of these primary components, a complicated procedure is required including periodical in-service inspection, failure assessment, fracture mechanics analysis, etc. Also, experts in different fields have to co-operate to resolve the integrity issues on the basis of inspection results. This integrity evaluation process usually takes long, and thus, is detrimental for the plant productivity. Therefore, an effective safety evaluation system is essential to manage integrity issues on a nuclear power plant. In this paper, a web-based integrity evaluation system for primary components in a nuclear power plant is proposed. The proposed system, which is named as WEBIES (web-based integrity evaluation system), has been developed in the form of 3-tier system architecture. The system consists of three servers; application program server, user interface program server and data warehouse server. The application program server includes the defect acceptance analysis module and the fracture mechanics analysis module which are programmed on the basis of ASME sec. XI, appendix A. The data warehouse server provides data for the integrity evaluation including material properties, geometry information, inspection data and stress data. The user interface program server provides information to all co- workers in the field of integrity evaluation. The developed system provides engineering knowledge-based information and concurrent and collaborative working environment through internet, and thus, is expected to raise the efficiency of integrity evaluation procedures on primary components of a nuclear power plant. (orig.)

  3. Development of web-based integrity evaluation system for primary components in a nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S.M.; Kim, J.C.; Choi, J.B.; Kim, Y.J. [SAFE Research Center, Sungkyunkwan Univ., Suwon (Korea); Choi, S.N.; Jang, K.S.; Hong, S.Y. [Korea Electronic Power Research Inst., Daejeon (Korea)

    2004-07-01

    A nuclear power plant is composed of a number of primary components. Maintaining the integrity of these components is one of the most critical issues in nuclear industry. In order to maintain the integrity of these primary components, a complicated procedure is required including periodical in-service inspection, failure assessment, fracture mechanics analysis, etc. Also, experts in different fields have to co-operate to resolve the integrity issues on the basis of inspection results. This integrity evaluation process usually takes long, and thus, is detrimental for the plant productivity. Therefore, an effective safety evaluation system is essential to manage integrity issues on a nuclear power plant. In this paper, a web-based integrity evaluation system for primary components in a nuclear power plant is proposed. The proposed system, which is named as WEBIES (web-based integrity evaluation system), has been developed in the form of 3-tier system architecture. The system consists of three servers; application program server, user interface program server and data warehouse server. The application program server includes the defect acceptance analysis module and the fracture mechanics analysis module which are programmed on the basis of ASME sec. XI, appendix A. The data warehouse server provides data for the integrity evaluation including material properties, geometry information, inspection data and stress data. The user interface program server provides information to all co- workers in the field of integrity evaluation. The developed system provides engineering knowledge-based information and concurrent and collaborative working environment through internet, and thus, is expected to raise the efficiency of integrity evaluation procedures on primary components of a nuclear power plant. (orig.)

  4. Ultrasonic Fingerprinting of Structural Materials: Spent Nuclear Fuel Containers Case-Study

    Science.gov (United States)

    Sednev, D.; Lider, A.; Demyanuk, D.; Kroening, M.; Salchak, Y.

    Nowadays, NDT is mainly focused on safety purposes, but it seems possible to apply those methods to provide national and IAEA safeguards. The containment of spent fuel in storage casks could be dramatically improved in case of development of so-called "smart" spent fuel storage and transfer casks. Such casks would have tamper indicating and monitoring/tracking features integrated directly into the cask design. The microstructure of the containers material as well as of the dedicated weld seam is applied to the lid and the cask body and provides a unique fingerprint of the full container, which can be reproducibly scanned by using an appropriate technique. The echo-sounder technique, which is the most commonly used method for material inspection, was chosen for this project. The main measuring parameter is acoustic noise, reflected from material's artefacts. The purpose is to obtain structural fingerprinting. Reference measurement and additional measurement results were compared. Obtained results have verified the appliance of structural fingerprint and the chosen control method. The successful authentication demonstrates the levels of the feature points' compliance exceeding the given threshold which differs considerably from the percentage of the concurrent points during authentication from other points. Since reproduction or doubling of the proposed unique identification characteristics is impossible at the current state science and technology, application of this technique is considered to identify the interference into the nuclear materials displacement with high accuracy.

  5. Corrosion susceptibility of steel drums containing cemented intermediate level nuclear wastes

    Science.gov (United States)

    Duffó, Gustavo S.; Farina, Silvia B.; Schulz, Fátima M.; Marotta, Francesca

    2010-10-01

    Cementation processes are used as immobilization techniques for low or intermediate level radioactive waste for economical and safety reasons and for being a simple operation. In particular, ion-exchange resins commonly used for purification of radioactive liquid waste from nuclear reactors are immobilized before being stored to improve the leach resistance of the waste matrix and to maintain mechanical stability. Combustible solid radioactive waste can be incinerated and the resulting ashes can also be immobilized before storage. The immobilized resins and ashes are then contained in steel drums that may undergo corrosion depending on the presence of certain contaminants. The work described in this paper was aimed at evaluating the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins and incineration ashes containing different concentrations of aggressive species (mostly chloride and sulphate ions). A special type of specimen was designed to simulate the cemented waste in the drum. The evolution of the corrosion potential and the corrosion current density of the steel, as well as the electrical resistivity of the matrix were monitored over a time period of 1 year. The results show the deleterious effect of chloride on the expected lifespan of the waste containers.

  6. Standard practice for sampling special nuclear materials in multi-container lots

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1987-01-01

    1.1 This practice provides an aid in designing a sampling and analysis plan for the purpose of minimizing random error in the measurement of the amount of nuclear material in a lot consisting of several containers. The problem addressed is the selection of the number of containers to be sampled, the number of samples to be taken from each sampled container, and the number of aliquot analyses to be performed on each sample. 1.2 This practice provides examples for application as well as the necessary development for understanding the statistics involved. The uniqueness of most situations does not allow presentation of step-by-step procedures for designing sampling plans. It is recommended that a statistician experienced in materials sampling be consulted when developing such plans. 1.3 The values stated in SI units are to be regarded as the standard. 1.4 This standard does not purport to address all of the safety problems, if any, associated with its use. It is the responsibility of the user of this standar...

  7. Computational concept for the containment liner for a nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Nagelstutz, Franz; Anders, Nils [Babcock Noell GmbH, Wuerzburg (Germany). Abt. Berechnung

    2010-05-15

    The determination of the optimal design of the Containment Liner considering amount of material, manufacturing and erection was the challenge for the engineering team of Babcock Noell GmbH. Several load cases for normal operation and accidental conditions as well as severe accidents have been analyzed. A realistic consideration of impacts by accidents was especially difficult. The special load cases in the vicinity of penetrations and anchor plates have been calculated. The results of theses analyses have been considered in the actual design of the liner. An integrated concept from planning, manufacturing and erection of this large component has been implemented, which is the topic of the speech 'ENGINEERING AND INNOVATIVE ERECTION CONCEPT FOR THE CONTAINMENT LINER FOR AN EPR trademark ' given by Dr. Rainer Goehring, Babcock Noell GmbH, Division Nuclear Technology Projects, Wuerzburg. He demonstrates that within the given time frame, with the required quality and within the required tolerances the containment liner can be erected. (orig.)

  8. Corrosion susceptibility of steel drums containing cemented intermediate level nuclear wastes

    International Nuclear Information System (INIS)

    Duffo, Gustavo S.; Farina, Silvia B.; Schulz, Fatima M.; Marotta, Francesca

    2010-01-01

    Cementation processes are used as immobilization techniques for low or intermediate level radioactive waste for economical and safety reasons and for being a simple operation. In particular, ion-exchange resins commonly used for purification of radioactive liquid waste from nuclear reactors are immobilized before being stored to improve the leach resistance of the waste matrix and to maintain mechanical stability. Combustible solid radioactive waste can be incinerated and the resulting ashes can also be immobilized before storage. The immobilized resins and ashes are then contained in steel drums that may undergo corrosion depending on the presence of certain contaminants. The work described in this paper was aimed at evaluating the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins and incineration ashes containing different concentrations of aggressive species (mostly chloride and sulphate ions). A special type of specimen was designed to simulate the cemented waste in the drum. The evolution of the corrosion potential and the corrosion current density of the steel, as well as the electrical resistivity of the matrix were monitored over a time period of 1 year. The results show the deleterious effect of chloride on the expected lifespan of the waste containers.

  9. Contribution to the simulation of the behavior of containment buildings in nuclear power plants

    International Nuclear Information System (INIS)

    Jason, Ludovic

    2016-01-01

    In this document, containment buildings of nuclear power plants in reinforced and prestressed concrete are studied. The mechanical behavior is first investigated. It includes studies at different scales, from the industrial scale at which it is difficult to obtain detailed information about the crack openings, to the structural scale at which it becomes possible to develop more refined approaches (steel-concrete bond, modeling of the different heterogeneities, constitutive models for concrete,...). When it is possible, the developments are based on the combination between modeling, simulation and experiments. A new steel-concrete bond model is especially developed. Its interest, compared to the classical 'perfect bond' hypothesis is particularly discussed. An adaptive condensation technique is finally proposed to bridge the gap between the scale of the containment and the scale of Structural Representative Volumes. As the tightness of the containment buildings is related to the fluid flow through concrete, two approaches are proposed and compared. The first one is based on a chained hydro mechanical simulation and tends to show that the mechanical damage is only influent when it crosses the whole concrete section. In this case, a localized approach is preferred from which a reference crack opening is defined. (author) [fr

  10. Passive 3D imaging of nuclear waste containers with Muon Scattering Tomography

    Science.gov (United States)

    Thomay, C.; Velthuis, J.; Poffley, T.; Baesso, P.; Cussans, D.; Frazão, L.

    2016-03-01

    The non-invasive imaging of dense objects is of particular interest in the context of nuclear waste management, where it is important to know the contents of waste containers without opening them. Using Muon Scattering Tomography (MST), it is possible to obtain a detailed 3D image of the contents of a waste container on reasonable timescales, showing both the high and low density materials inside. We show the performance of such a method on a Monte Carlo simulation of a dummy waste drum object containing objects of different shapes and materials. The simulation has been tuned with our MST prototype detector performance. In particular, we show that both a tungsten penny of 2 cm radius and 1 cm thickness, and a uranium sheet of 0.5 cm thickness can be clearly identified. We also show the performance of a novel edge finding technique, by which the edges of embedded objects can be identified more precisely than by solely using the imaging method.

  11. Effects of secondary containment air cleanup system leakage on the accident offsite dose as determined during preop tests of the Sequoyah Nuclear Plant

    International Nuclear Information System (INIS)

    Klaes, L.J.; Nass, S.A.; Proctor, L.D.

    1981-01-01

    The Sequoyah Nuclear Plant has two secondary containments. One is the annular region between the primary containment and the shield building surrounding the primary containment. The second is the auxiliary building secondary containment enclosure which is potentially subject to direct airborne radioactivity. Two air cleanup systems are provided to serve these areas. The emergency gas treatment system (EGTS) serves the annulus between the primary containment and the shield building, and the auxiliary building gas treatment system (ABGTS) serves the area inside of the auxiliary building secondary containment enclosure. The major function served by these air cleanup systems is that of controlling and processing airborne contamination released in these areas during any accident up to a design basis accident. This is accomplished by (1) creating a negative pressure in the areas served to ensure that no unprocessed air is released to the atmosphere, (2) providing filtration units to process all air exhausted from the secondary containment spaces, and (3) providing a low-leakage enclosure to limit exhaust flows. Offsite dose effects due to secondary containment release rates, bypass leakage, and duct and damper leakages are presented and parameter variations are considered. For the EGTS, a recirculation system, the most important parameter is the total inleakage of the system which causes an increase in both whole body (gamma) and thyroid (iodine) doses. For the ABGTS, a once-through system, the most important paramter is the inleakage which bypasses the filters resulting in an increase in the thyroid dose only. Actual preoperational test data are utilized. Problems encountered during the preop test are summarized. Solutions incorporated to bring the EGTS and ABGTS air cleanup systems within the test acceptance criteria required to meet offsite dose limitations are discussed and the resultant calculated offsite dose is presented

  12. A proposed structural, risk-informed approach to the periodicity of CANDU-6 nuclear containment integrated leak rate testing

    Energy Technology Data Exchange (ETDEWEB)

    Saliba, N. [McGill Univ., Dept. of Civil Engineering and Applied Mechanics, Montreal, Quebec (Canada); Komljenovic, D. [Hydro-Quebec, Gentilly-2 Nuclear Power Plant, Becancour, Quebec (Canada); Chouinard, L. [McGill Univ., Dept. of Civil Engineering and Applied Mechanics, Montreal, Quebec (Canada); Vaillancourt, R.; Chretien, G. [Hydro-Quebec, Gentilly-2 Nuclear Power Plant, Becancour, Quebec (Canada); Gocevski, V. [Hydro-Quebec Equipements, Montreal, Quebec (Canada)

    2010-07-01

    As ultimate lines of defense against leakage of large amounts of radioactive material to the environment in case of major reactor accidents, containments have been monitored through well designed periodic tests to ensure their proper performance. Regulatory organizations have imposed types and frequencies of containment tests based on highly-conservative deterministic approaches, and judgments of knowledgeable experts. Recent developments in the perception and methods of risk evaluation have been applied to rationalize the leakage-rate testing frequencies while maintaining risks within acceptable levels, preserving the integrity of containments, and respecting the defense-in-depth philosophy. The objective of this paper is to introduce a proposed risk-informed decision making framework on the periodicity of nuclear containment ILRTs for CANDU-6 nuclear power plants based on five main decision criteria, namely: 1) the containment structural integrity; 2) inputs from PSA Level-2; 3) the requirements of deterministic safety analyses and defense-in-depth concepts; 4- the obligations under regulatory and standard requirements; and 5) the return of experience from nuclear containments historic performance. The concepts of dormant reliability and structural fragility will guide the assessment of the containment structural integrity, within the general context of a global containment life cycle management program. This study is oriented towards the requirements of CANDU-6 reactors, in general, and Hydro-Quebec's Gentilly-2 nuclear power plant, in particular. The present article is the first part in a series of papers that will comprehensively detail the proposed research. (author)

  13. 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.

  14. A filter system for steam-gas mixture ejections from under a nuclear reactor containment following a severe accident

    International Nuclear Information System (INIS)

    Dulepov, Ju. N.; Sharygin, L. M.; Tretjakov, S. Ja.; Shtin, A.P.; Glushko, V. V.; Babenko, E. A.; Kurakov, Ju. A.

    1997-01-01

    In this paper newly built NPPs obligatory incorporate a containment having a filter system for removing radioactive materials ejections under severe accidents including nuclear fuel melting is described. The system prevents a containment failure and provides ejected radioactive materials decontamination to permissible levels. The physical-chemical and chemical characteristics of Termoxid-58 sorbent (TiO 5 based sorbent) are presented

  15. A proposed structural, risk-informed approach to the periodicity of CANDU-6 nuclear containment integrated leak rate testing

    International Nuclear Information System (INIS)

    Saliba, N.; Komljenovic, D.; Chouinard, L.; Vaillancourt, R.; Chretien, G.; Gocevski, V.

    2010-01-01

    As ultimate lines of defense against leakage of large amounts of radioactive material to the environment in case of major reactor accidents, containments have been monitored through well designed periodic tests to ensure their proper performance. Regulatory organizations have imposed types and frequencies of containment tests based on highly-conservative deterministic approaches, and judgments of knowledgeable experts. Recent developments in the perception and methods of risk evaluation have been applied to rationalize the leakage-rate testing frequencies while maintaining risks within acceptable levels, preserving the integrity of containments, and respecting the defense-in-depth philosophy. The objective of this paper is to introduce a proposed risk-informed decision making framework on the periodicity of nuclear containment ILRTs for CANDU-6 nuclear power plants based on five main decision criteria, namely: 1) the containment structural integrity; 2) inputs from PSA Level-2; 3) the requirements of deterministic safety analyses and defense-in-depth concepts; 4- the obligations under regulatory and standard requirements; and 5) the return of experience from nuclear containments historic performance. The concepts of dormant reliability and structural fragility will guide the assessment of the containment structural integrity, within the general context of a global containment life cycle management program. This study is oriented towards the requirements of CANDU-6 reactors, in general, and Hydro-Quebec's Gentilly-2 nuclear power plant, in particular. The present article is the first part in a series of papers that will comprehensively detail the proposed research. (author)

  16. Comparison of different LMFBR primary containment codes applied to a Benchmark problem

    International Nuclear Information System (INIS)

    Benuzzi, A.

    1986-01-01

    The Cont Benchmark calculation exercise is a project sponsored by the Containment Loading and Response Group, a subgroup of the Safety Working Group of the Fast Reactor Coordinating Committee - CEC. A full-size typical Pool type LMFBR undergoing a postulated Core Disruptive Accident (CDA) has been defined by Belgonucleaire-Brussels under a study contract financed by the CEC and has been submitted to seven containment code calculations. The results of these calculations are presented and discussed in this paper

  17. Imaging Nuclear-Cytoplasmic Dynamics in Primary and Metastatic Colon Cancer in Nude Mice.

    Science.gov (United States)

    Hasegawa, Kosuke; Suetsugu, Atsushi; Nakamura, Miki; Matsumoto, Takuro; Aoki, Hitomi; Kunisada, Takahiro; Bouvet, Michael; Shimizu, Masahito; Hoffman, Robert M

    2016-05-01

    Colon cancer frequently results in metastasis to the liver, where it becomes the main cause of death. However, the cell cycle in primary tumors and metastases is poorly understood. We developed a mouse model of liver metastasis using the human colon cancer cell line HCT-116, which expresses green fluorescent protein (GFP) in the nucleus and red fluorescent protein (RFP) in the cytoplasm (HCT-116-GFP-RFP). HCT-116 GFP-RFP cells were injected into the spleen of nu/nu nude mice. HCT-116-GFP-RFP cells subsequently formed primary tumors in the spleen, as well as metastatic colonies in the liver and retroperitoneum by 28 days after cell transplantation. Using an Olympus FV1000 confocal microscope, it was possible to clearly image mitosis of the dual-colored colon cancer cells in the primary tumor as well as liver and other metastases. Multi-nucleate cancer cells, in addition to mono-nucleate cancer cells and their mitosis, were observed in the primary tumor and metastasis. Multi-nucleate HCT-116-GFP-RFP cells were also observed after culture of the primary and metastatic tumors. A similar ratio of mono-nucleate, multi-nucleate, and mitotic cells grew from the primary and metastatic tumors in culture, suggesting similarity of the nuclear-cytoplasmic dynamics of primary and metastatic cancer cells, further emphasizing the stochastic nature of metastasis. Our results demonstrate a similar heterogeneity of nuclear-cytoplasmic dynamics within primary tumors and metastases, which may be an important factor in the stochastic nature of metastasis. Copyright© 2016 International Institute of Anticancer Research (Dr. John G. Delinassios), All rights reserved.

  18. Comparing the analgesic effect of heat patch containing iron chip and ibuprofen for primary dysmenorrhea: a randomized controlled trial

    Directory of Open Access Journals (Sweden)

    Navvabi Rigi Shahindokht

    2012-08-01

    Full Text Available Abstract Background Primary dysmenorrhea is a common and sometimes disabling condition. In recent years, some studies aimed to improve the treatment of dysmenorrhea, and therefore, introduced several therapeutic measures. This study was designed to compare the analgesic effect of iron chip containing heat wrap with ibuprofen for the treatment of primary dysmenorrhea. Methods In this randomized (IRCT201107187038N2 controlled trial, 147 students (18–30 years old with the diagnosis of primary dysmenorrhea were enrolled considering the CONSORT guideline. Screening for primary dysmenorrhea was done by a two-question screening tool. The participants were randomly assigned into one of the intervention groups (heat Patch and ibuprofen. Data regarding the severity and emotional impact of the pain were recorded by a shortened version of McGill Pain Questionnaire (SF-MPQ. Student's t test was used for statistical analysis. Results The maximum and minimum pain severities were observed at 2 and 24 hours in both groups. The severity of sensual pain at 8, 12, and 24 hours was non-significantly less in the heat Patch group. There was also no significant difference between the groups regarding the emotional impact of pain at the first 2, 4, 8, 12 and 12 hours of menstruation. Conclusions Heat patch containing Iron chip has comparable analgesic effects to ibuprofen and can possibly be used for primary dysmenorrhea. Trial registration IRCT201107187038N2

  19. 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.

  20. Design and analysis of new prestressed concrete containment and its passive cooling system for nuclear power plants

    International Nuclear Information System (INIS)

    Tan Xiaoshi; Li Xiaowei; Li Xiaotian; He Shuyan

    2014-01-01

    A new nuclear power plant prestressed concrete containment and its passive cooling system design were proposed for CAP1700 nuclear power plant as an example. The thermal-hydraulic calculation method for the new passive containment cooling system of CAP1700 was introduced and the operating parameters in accident condition were obtained. The result shows that the design of passive containment cooling system for CAP1700 is feasible and can meet the cooling demand in accident condition. Reservoir capacity of tank has a big margin and can be further optimized by calculation. (authors)

  1. Structural review of the Palisades Nuclear Power Plant Unit 1 containment structure under combined loads for the Systematic Evaluation Program

    International Nuclear Information System (INIS)

    Liaw, C.Y.; Debeling, A.; Tsai, N.C.

    1981-12-01

    A structural reassessment of the containment structure of the Palisades Nuclear Power Plant Unit 1 was performed for the Nuclear Regulatory Commission as part of the Systematic Evaluation Program. Conclusions about the ability of the containment structure to withstand the Abnormal/Extreme Environment are presented. The reassessment focused mainly on the overall structural integrity of the containment building for the Abnormal/Extreme Environment. In this case, the Abnormal Environmental condition is caused by the worst case of either a Loss-of-Coolant Accident or a main steam line break. The Extreme Environmental condition is the Safe Shutdown Earthquake

  2. Floor Response Spectra of Nuclear Containment Building with Soil-Structure Interaction

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Choon Gyo; Ryu, Jeong Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-10-15

    This paper presents a seismic analysis technique for a 3D soil-structure interaction(SSI) system in frequency domain, based on the finite element formulation incorporating frequency-dependent dynamic infinite elements for the far field soil region. Earthquake input motions are regarded as traveling SV-wave which is vertically incident from a far-field soil region. In which, the equivalent earthquake forces in the frequency domain are calculated using the exterior rigid boundary method and the free field response analysis. For the application, floor response spectra analyses for nuclear containment building on a soil medium is carried out, the obtained results are compared with the free field response by other solution.

  3. Effect of prestress force on longitudinal vibration of bonded tendons embedded in a nuclear containment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byeong Hwa, E-mail: bhkim@kyungnam.ac.k [Dept. of Civil Engineering, Kyungnam University, 449 Wolyoung, Masan, Kyungnam 631-701 (Korea, Republic of); Jang, Jung Bum, E-mail: jbjang@kepri.re.k [Nuclear Power Laboratory, Korea Electric Power Research Institute, 103-16 Munji-Dong, Yuseong-Gu, Daejeon 305-380 (Korea, Republic of); 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); Lee, Do Hyung, E-mail: dohlee@pcu.ac.k [Dept. of Civil Environmental and Railroad Engineering, Paichai University, Yonja 1 gil (439-6 Doma 2 dong), Seo-gu, Daejeon 302-735 (Korea, Republic of)

    2010-06-15

    In this paper, evaluation of tensile force on bonded seven-wire strand embedded in nuclear reactor containments has been nondestructively investigated. For this purpose, a total of eight prestressed concrete beam specimens with different tension force levels have been fabricated and tested in terms of using an impact pulse at one end of the strand. Subsequently, various longitudinal vibration characteristics excited by the pulse have been analyzed in detailed. In short, the present study has demonstrated that longitudinal frequency, elastic wave velocity, and elastic modulus are nonlinearly increased as the tensile force level is increased. It is thus postulated that the longitudinal vibration characteristics of the existing bonded tendons can be of promising indicators for the inspection of a tensile force loss.

  4. The adhesion characteristics of protective coating materials for the containment structure in nuclear power plants

    International Nuclear Information System (INIS)

    Lee, Sang-Kook; Shin, Jae-Chul

    2003-01-01

    Protective coating materials used in the containment structures should be durable for the designed 30 to 40 year lifetime of a nuclear power plant. At the present, these materials have not yet been developed. Therefore it is very important to keep the durability of the protective coating materials through persistent maintenance, and in order to achieve this, understanding the adhesion characteristics of the coating materials is of utmost importance. Therefore, this study attempts to find any methods for durability maintenance of these protective coating materials. To accomplish these aims, this study applied an experimental deterioration environment condition relevant to Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB), categorized as of Design Basis Accident (DBA), onto steel liner plate specimens covered with protective coating materials. Adhesion tests were performed on these deteriorated coating materials to characterize the physical properties and through these tests, the quantitative adhesion characteristics according to the history of deterioration environment were found

  5. Containment aerosol behaviour simulation studies in the BARC nuclear aerosol test facility

    International Nuclear Information System (INIS)

    Mayya, Y.S.; Sapra, B.K.; Khan, Arshad; Sunny, Faby; Nair, R.N.; Raghunath, Radha; Tripathi, R.M.; Markandeya, S.G.; Puranik, V.D.; Ghosh, A.K.; Kushwaha, H.S.; Shreekumar, K.P.; Padmanabhan, P.V.A.; Murthy, P.S.S.; Venlataramani, N.

    2005-02-01

    A Nuclear Aerosol Test Facility (NATF) has been built and commissioned at Bhabha Atomic Research Centre to carry out simulation studies on the behaviour of aerosols released into the reactor containment under accident conditions. This report also discusses some new experimental techniques for estimation of density of metallic aggregates. The experimental studies have shown that the dynamic densities of aerosol aggregates are far lower than their material densities as expected by the well-known fractal theory of aggregates. In the context of codes, this has significant bearing in providing a mechanistic basis for the input density parameter used in estimating the aerosol evolution characteristics. The data generated under the quiescent and turbulent conditions and the information on aggregate densities are now being subjected to the validation of the aerosol behaviour codes. (author)

  6. Alternative technology of containment construction for WWER 1000 nuclear power plant

    International Nuclear Information System (INIS)

    Chalus, Z.

    1982-01-01

    A number of alternatives was assessed for the assembly of the steel elements of the cylindrical part of containment. Alternative 1 is based on the common technology of manufacture, transport and assembly of reinforced concrete blocks of ca. 3x12 m in size, used for building leak-proof walls of WWER 440 nuclear power plants. Alternative 2 is based on reinforced concrete blocks using 12x12 m blocks assembled from individual elements on the building. site. Alternative 3 is a specific variant of the previous alternative. Alternative 4 envisages the assembly of a prefabricated support structure made of steel. Alternative 5 is based on the gradual assembly of partial elements mounted onto a support structure. Alternative 6 only differs from 5 in the method of assembly and manufacture of the support structure. All alternatives are shown in diagrams. (J.B.)

  7. Mechanistic modelling of the corrosion behaviour of copper nuclear fuel waste containers

    Energy Technology Data Exchange (ETDEWEB)

    King, F; Kolar, M

    1996-10-01

    A mechanistic model has been developed to predict the long-term corrosion behaviour of copper nuclear fuel waste containers in a Canadian disposal vault. The model is based on a detailed description of the electrochemical, chemical, adsorption and mass-transport processes involved in the uniform corrosion of copper, developed from the results of an extensive experimental program. Predictions from the model are compared with the results of some of these experiments and with observations from a bronze cannon submerged in seawater saturated clay sediments. Quantitative comparisons are made between the observed and predicted corrosion potential, corrosion rate and copper concentration profiles adjacent to the corroding surface, as a way of validating the long-term model predictions. (author). 12 refs., 5 figs.

  8. Development of mobile manipulator for maintenance work in containment vessels of nuclear power plants

    International Nuclear Information System (INIS)

    Omichi, Takeo; Nishihara, Masatoshi; Hosaka, Shigetaka; Nakayama, Junji; Sato, Masatoshi; Ishida, Michiyasu

    1985-01-01

    The teleoperation system with robot is described for in the containment vessels of nuclear power plants. We have developed a high level robot system as the practical use level. The robot is designed to execute the locomotions and manipulations required for closing and opening the valve, tightening the bolt and others. The robot consists of a locomotor with four legs and two driving wheels, an articulated manipulator with seven joints, and an ITV arm with stereo-camera. The size of the robot is small, that is about 500 mm in length, 500 mm in width, 1200 mm in height and 420 kg in weight. The robot can be operated in a hostile environment, which has a 10 6 R gamma ray dose, 70 deg C temperature, 100 % relative humidity. We have added an advanced control method in order to reduce the operator's load. Also, an interlock and a fail-safe control are installed in the robot system. (author)

  9. Method of chemical reprocessing of irradiated nuclear fuels (especially fuels containing uranium)

    International Nuclear Information System (INIS)

    Koch, G.

    1975-01-01

    The invention deals with a method for the extraction especially of fast breeder fuels of high burn-up. A quaternary ammonium nitrate of high molecular weight is put into an organic diluting medium as extraction agent, corresponding to the general formula NRR'R''R'''NO 3 where R,R' and R'' are aliphatic radicals, R''' a methyl radical and the sum of the C atoms is greater than 16. After the extraction of the aqueous nitric acid containing nuclear fuel solution with this extracting agent, uranium, plutonium (or also thorium) can be found to a very high percentage in the organic phase and can be practically quantitatively back-extracted by means of diluted nitric acid, sulphuric acid or acetic acid. By using 30 volume percent tricapryl methyl ammonium nitrate in diethyl benzene for example, a distribution coefficient of 10.3 is obtained for uranium. (RB/LH) [de

  10. Critical stresses in pintle, weldment and top head of nuclear waste container

    International Nuclear Information System (INIS)

    Ladkany, S.G.; Kniss, B.R.

    1992-01-01

    Critical stresses in the pintle, the weldment, and the top heads (flat and curved), of a high level nuclear waste container are evaluated under an annular loading. This loading is three times larger than the expected normal operating load. Results show that the shape and the thickness of the pintle and the top head, along with the thickness of the weldment, substantially affect the magnitude of the critical stresses and distortions in the various components (i.e. pintle, shell, and heads) when they are supporting a load. Stiffer top heads and pintles and larger weldment sizes reduce the critical stresses in all welded joints. Various shapes of curved top heads were investigated. In this paper an ASME flanged and dished top head, which has the same thickness as the canister, is analyzed

  11. Foreign research reactor irradiated nuclear fuel inventories containing HEU and LEU of United States origin

    International Nuclear Information System (INIS)

    Matos, J.E.

    1994-12-01

    This report provides estimates of foreign research reactor inventories of aluminum-based and TRIGA irradiated nuclear fuel elements containing highly enriched and low enriched uranium of United States origin that are anticipated in January 1996, January 2001, and January 2006. These fuels from 104 research reactors in 41 countries are the same aluminum-based and TRIGA fuels that were eligible for receipt under the Department of Energy's Offsite Fuels Policy that was in effect in 1988. All fuel inventory and reactor data that were available as of December 1, 1994, have been included in the estimates of approximately 14,300 irradiated fuel elements in January 1996, 18,800 in January 2001, and 22,700 in January 2006

  12. 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

  13. The nuclear waste containment and some aspects of the deep disposal concept

    International Nuclear Information System (INIS)

    Felix, B.; Thorner, P.; Raimbault, P.; Beaulieu, F.

    1995-01-01

    The French agency for the management of nuclear waste, ANDRA, is in charge of investigating the feasibility of deep disposal of high level waste in at least two types of geologic formation, leading to the validation of disposal concepts with and without retrievability. Plans to build two underground laboratories are afoot. Meanwhile, parametric modelling studies have been performed, with interesting results, some of which are shown here in graphic form. It is proved that if the overpack surrounding waste containers can be made to last for a thousand years, the dose resulting from Sr-90 and Cs-137 is nil. Conversely, the dose from actinides such as americium and Th-229 is largely unaffected by the package, being determined by their own low solubilities and underground water flow. Temperature rise in a granite host formation was modelled as a function of the distance between disposal boreholes. Finite element two dimensional calculations of water flow through backfill were also performed. 1 ref., 8 figs

  14. 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)

  15. Thermal-Hydraulic Analysis of the Nuclear Power Engineering Corporation Containment Experiments with GOTHIC

    International Nuclear Information System (INIS)

    Wiles, Lawrence E.; George, Thomas L.

    2003-01-01

    GOTHIC version 7.0 was used to model five tests that were conducted in the Nuclear Power Engineering Corporation facility in Japan. The tests involved steam and helium injection into a preheated, spray-moderated, 1/4-scale model of a pressurized water reactor dry containment. Comparison of GOTHIC predictions to measured data for pressure, vapor temperatures, structure surface temperatures, and helium concentrations provided the opportunity to evaluate methods for modeling gas dispersion, drop heat and mass transfer, and surface heat transfer.The test facility includes three floors. The lower two floors are partitioned into a variety of rooms that simulate the lower regions of the modeled containment. On the upper floor, rooms that simulate the steam generator enclosures and the pressurizer enclosure extend into the dome, which represents about two-thirds of the total volume of the containment.The GOTHIC model was defined with 30 control volumes using a mix of lumped parameter volumes and subdivided volumes that employ a three-dimensional mesh. Each volume included several thermal conductors to model the various structures. More than 100 flow paths were used to model the hydraulic connections.Comparison of predictions to data showed that enhanced grid resolution in the vicinity of the steam-helium release point served to limit dispersion of the steam-helium plume. The data comparisons also suggested that spray effectiveness was reduced by drop impact with the containment wall and by the high drop concentration. The data comparisons further suggested that the presence of condensation, sprays, splashing, and other wetting mechanisms should be considered to obtain a reasonable estimate of the effect of liquid films on the structure surfaces

  16. 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)

  17. Composition - structure - properties relationships of peraluminous glasses for nuclear waste containment

    International Nuclear Information System (INIS)

    Piovesan, Victor

    2016-01-01

    Part of the Research and Development program concerning high level nuclear waste conditioning aims to assess new glass formulations able to incorporate a high waste content with enhanced properties in terms of homogeneity, thermal stability, long term behavior and process ability. This study focuses on peraluminous glasses, defined by an excess of aluminum ions Al"3"+ in comparison with modifier elements such as Na"+, Li"+ or Ca"2"+. A Design of Experiment approach has been employed to determine relationships between composition of simplified peraluminous glasses (SiO_2 - B_2O_3 - Al_2O_3 - Na_2O - Li_2O - CaO - La_2O_3) and their physical properties such as viscosity, glass transition temperature and glass homogeneity. Moreover, some structural investigation (NMR) was performed in order to better understand the structural role of Na"+, Li"+ and Ca"2"+ and the structural organization of peraluminous glasses. Then, physical and chemical properties of fully simulated peraluminous glasses were characterized to evaluate transposition between simplified and fully simulated glasses and also to put forward the potential of peraluminous glasses for nuclear waste containment. (author) [fr

  18. Optimization of a primary circuit of the nuclear power plant from the vibration point of view

    International Nuclear Information System (INIS)

    Dupal, J.; Zeman, V.

    2003-01-01

    The primary circuit of the nuclear power plant (NPP) as a dynamical vibrating system can be disturbed by various excitation including earthquake or pressure pulsation generated by main circulation pumps (MCP). Especially, unpleasant pulsation vibration growth can be caused by the small differences of revolutions between main circulation pumps of the individual coolant loops. This growth corresponds to the well known beats. The paper deals with an approach to the improving and optimization of dynamical properties of the whole primary circuit system including the reactor and coolant loops under pressure pulsation. (author)

  19. High temperature alloys for the primary circuit of a prototype nuclear process heat plant

    International Nuclear Information System (INIS)

    Ennis, P.J.; Schuster, H.

    1979-01-01

    As part of a comprehensive materials test programme for the High Temperature Reactor Project 'Prototype Plant for Nuclear Process Heat' (PNP), high temperature alloys are being investigated for primary circuit components operating at temperatures above 750 0 C. On the basis of important material parameters, in particular corrosion behaviour and mechanical properties in primary coolant helium, the potential of candidate alloys is discussed. By comparing specific PNP materials data with the requirements of PNP and those of conventional plant, the implications for the materials programme and component design are given. (orig.)

  20. Geotechnical studies relevant to the containment of underground nuclear explosions at the Nevada Test Site

    Energy Technology Data Exchange (ETDEWEB)

    Heuze, F.E.

    1982-05-01

    The Department of Energy and the Department of Defense are actively pursuing a program of nuclear weapons testing by underground explosions at the Nevada Test Site (NTS). Over the past 11 years, scores of tests have been conducted and the safety record is very good. In the short run, emphasis is put on preventing the release of radioactive materials into the atmosphere. In the long run, the subsidence and collapse of the ground above the nuclear cavities also are matters of interest. Currently, estimation of containment is based mostly on empiricism derived from extensive experience and on a combination of physical/mechanical testing and numerical modeling. When measured directly, the mechanical material properties are obtained from short-term laboratory tests on small, conventional samples. This practice does not determine the large effects of scale and time on measured stiffnesses and strengths of geological materials. Because of the limited data base of properties and in situ conditions, the input to otherwise fairly sophisticated computer programs is subject to several simplifying assumptions; some of them can have a nonconservative impact on the calculated results. As for the long-term, subsidence and collapse phenomena simply have not been studied to any significant degree. This report examines the geomechanical aspects of procedures currently used to estimate containment of undergroung explosions at NTS. Based on this examination, it is concluded that state-of-the-art geological engineering practice in the areas of field testing, large scale laboratory measurements, and numerical modeling can be drawn upon to complement the current approach.

  1. Geotechnical studies relevant to the containment of underground nuclear explosions at the Nevada Test Site

    International Nuclear Information System (INIS)

    Heuze, F.E.

    1982-05-01

    The Department of Energy and the Department of Defense are actively pursuing a program of nuclear weapons testing by underground explosions at the Nevada Test Site (NTS). Over the past 11 years, scores of tests have been conducted and the safety record is very good. In the short run, emphasis is put on preventing the release of radioactive materials into the atmosphere. In the long run, the subsidence and collapse of the ground above the nuclear cavities also are matters of interest. Currently, estimation of containment is based mostly on empiricism derived from extensive experience and on a combination of physical/mechanical testing and numerical modeling. When measured directly, the mechanical material properties are obtained from short-term laboratory tests on small, conventional samples. This practice does not determine the large effects of scale and time on measured stiffnesses and strengths of geological materials. Because of the limited data base of properties and in situ conditions, the input to otherwise fairly sophisticated computer programs is subject to several simplifying assumptions; some of them can have a nonconservative impact on the calculated results. As for the long-term, subsidence and collapse phenomena simply have not been studied to any significant degree. This report examines the geomechanical aspects of procedures currently used to estimate containment of undergroung explosions at NTS. Based on this examination, it is concluded that state-of-the-art geological engineering practice in the areas of field testing, large scale laboratory measurements, and numerical modeling can be drawn upon to complement the current approach

  2. Corrosion product behaviour in the primary circuit of the KNK nuclear reactor facility

    International Nuclear Information System (INIS)

    Stamm, H.H.; Stade, K.Ch.

    1976-01-01

    During nuclear operation of the KNK facility from 1972 until September 1974 the composition and behaviour of radionuclides occuring in the primary circuit were investigated. Besides traces of 140 Ba/ 140 La, no fission product activity was detectable in the KNK primary circuit. The fuel element purification from sodium deposits (prior to transport to the reprocessing plant) did not yield any indication of a fuel element failure during KNK-I operation. The activity inventory of the primary loop was exclusively made up of activated corrosion products and 22 Na. The main activity was due to 65 Zn, followed by 54 Mn, 22 Na, sup(110m)Ag, 182 Ta, 60 Co and 124 Sb. It was found that the sorption of 65 Zn and 54 Mn on crucibles made from nickel was condiserably higher than on vessels made from other materials. This observation was confirmed both in tests with material samples from the primary circuit and for disks of gate valves of the primary circuit. sup(110m)Ag did hardly exhibit any sorption effects and had been dissolved largely homogeneously in the hot primary coolant. In the first primary cold trap which was removed from the circuit after some 20,000 hours of operation, only 65 Zn and 54 Mn were detected in addition to traces of 60 Co and 182 Ta. (author)

  3. Comparative research of finite element methods for perforated structures of nuclear power plant primary equipment

    International Nuclear Information System (INIS)

    Xiong Guangming; Deng Xiaoyun; Jin Ting

    2013-01-01

    Many perforated structures are used for nuclear power plant primary equipment, and they are complex, and have various forms. In order to explore the analysis and evaluation method, this paper used finite element method and equivalent analytic method to do the comparative analysis of perforated structures. The paper considered the main influence factors (including perforated forms, arrangements, and etc.), obtaining the systematic analysis methods of perforated structures. (authors)

  4. Proceedings of the Canadian Nuclear Society international conference on containment design

    International Nuclear Information System (INIS)

    Black, R.K.

    1984-01-01

    The subject of containment was discussed under the major topic headings of: USA views on containment design operation and regulation, containment structure analysis, containment envelope and energy suppression systems, European views on containment design operation and regulation, containment atmosphere control, Canadian views on containment design operation and regulation, containment analysis, and regulatory requirements, testing, and periodic inspection

  5. GASFLOW: the theoretical model to analyze accidents in nuclear containments, confinements, and facility buildings

    International Nuclear Information System (INIS)

    Travis, J.R.; Wilson, T.L.

    1993-01-01

    This report documents the governing physical equations for GASFLOW, a finite-volume computer code for solving transient, three-dimensional, compressible, Navier-Stokes equations for multiple gas species. The code is designed to be a best-estimate tool for predicting the transport, mixing, and combustion of hydrogen and other gases in nuclear reactor containments, confinements, and other facility buildings. An analysis with GASFLOW will result in time-dependent gas species concentrations throughout the structure analyzed, and in the event of combustion the pressure and temperature loadings on the walls and internal structures. GASFLOW can model geometrically complex containment systems with multiple compartments and internal structures. It can calculate gas behavior of low-speed buoyancy-driven flows, of diffusion-dominated flows, and during deflagrations. The code can model condensation heat transfer to walls and internal structures by natural and forced convection; chemical kinetics of combustion of hydrogen or hydrocarbon fuels; and fluid turbulence. Heat conduction within walls and structures is considered one-dimensional. 17 refs., 1 fig., 1 tab

  6. Chemical durability of borosilicate glasses containing simulated high-level nuclear wastes, 1

    International Nuclear Information System (INIS)

    Hara, Shigeo; Terai, Ryohei; Yamanaka, Hiroshi

    1983-01-01

    The Soxhlet-type leaching test apparatus has been developed to evaluate the chemical durability of some borosilicate glasses containing simulated High-Level nuclear Wastes, HLW. After the leaching over the temperature range of 50 0 -95 0 C, the weight loss of specimens with time was determined on both the samples of blocks and grains, and various components dissolved into water were analyzed by atomic absorption and colorimetry technique. It was found that Soxhlet-type test method was more useful than JIS test method, because the specimens in Soxhlet type apparatus were forced always to react with pure water and the mechanism of leaching could be evaluate accurately. The chemical durability of commercial glasses decreases generally with increasing of alkali contents in glasses. In the case of these borosilicate glasses containing HLW, however, the leachability was apparently independent on the alkali contents because of the complexity of these glass compositions. The variation of leaching rate with temperature suggests that dissolution mechanism changes with temperature. (author)

  7. Selection of candidate canister materials for high-level nuclear waste containment in a tuff repository

    International Nuclear Information System (INIS)

    McCright, R.D.; Weiss, H.; Juhas, M.C.; Logan, R.W.

    1983-11-01

    A repository located at Yucca Mountain at the Nevada Test Site is a potential site for permanent geological disposal of high-level nuclear waste. The repository can be located in a horizon in welded tuff, a volcanic rock, which is above the static water level at this site. The environmental conditions in this unsaturated zone are expected to be air and water vapor dominated for much of the containment period. Type 304L stainless steel is the reference material for fabricating canisters to contain the solid high-level wastes. Alternative stainless alloys are considered because of possible susceptibility of 304L to localized and stress forms of corrosion. For the reprocessed glass wastes, the canisters serve as the recipient for pouring the glass with the result that a sensitized microstructure may develop because of the times at elevated temperatures. Corrosion testing of the reference and alternative materials has begun in tuff-conditioned water and steam environments. 21 references, 8 figures, 8 tables

  8. Gamma radiolysis effects on leaching behavior of ceramic materials for nuclear fuel waste immobilization containers

    International Nuclear Information System (INIS)

    Onofrei, M.; Raine, D.K.; Hocking, W.H.; George, K.; Betteridge, J.S.

    1986-01-01

    The leaching behavior of ceramic materials for nuclear fuel waste immobilization containers, under the influence of a moderate gamma dose rate (4 Gy/h), has been investigated. Samples of Al/sub 2/O/sub 3/, stabilized ZrO/sub 2/, TiO/sub 2/, cermet (70% Al/sub 2/O-30% TiC), porcelain (with high Al/sub 2/O/sub 3/ content), and concrete (with sulfate-resisting portland cement plus silica fume) have been leached in Standard Canadian Shield Saline Solution (SCSSS), and SCSSS plus clay and sand (components of the disposal system), at 100 0 and 150 0 C for 231 and 987 days, respectively. Leaching solutions were analyzed and the surfaces of the leached samples were investigated by scanning electron microscopy in conjunction with energy dispersive X-ray spectroscopy and secondary ion mass spectrometry. Radiolysis did not appear to enhance the leaching, with or without bentonite and sand in the system. Analysis of the gas phase from sealed capsules showed O/sub 2/ depletion and production of CO/sub 2/ in all experiments containing bentonite. The decrease in O/sub 2/ is attributed to the leaching from the clay of Fe(II) species, which can participate in redox reactions with radicals generated by radiolysis. The CO/sub 2/ is produced from either the organic or inorganic fraction in the bentonite

  9. Properties and solubility of chrome in iron alumina phosphate glasses containing high level nuclear waste

    International Nuclear Information System (INIS)

    Huang, W.; Day, D.E.; Ray, C.S.; Kim, C.W.; Reis, S.T.D.

    2004-01-01

    Chemical durability, glass formation tendency, and other properties of iron alumina phosphate glasses containing 70 wt% of a simulated high level nuclear waste (HLW), doped with different amounts of Cr 2 O 3 , have been investigated. All of the iron alumina phosphate glasses had an outstanding chemical durability as measured by their small dissolution rate (1 . 10 -9 g/(cm 2 . min)) in deionized water at 90 C for 128 d, their low normalized mass release as determined by the product consistency test (PCT) and a barely measurable corrosion rate of 2 . d) after 7 d at 200 C by the vapor hydration test (VHT). The solubility limit for Cr 2 O 3 in the iron phosphate melts was estimated at 4.1 wt%, but all of the as-annealed melts contained a few percent of crystalline Cr 2 O 3 that had no apparent effect on the chemical durability. The chemical durability was unchanged after deliberate crystallization, 48 h at 650 C. These iron phosphate waste forms, with a waste loading of at least 70 wt%, can be readily melted in commercial refractory crucibles at 1250 C for 2 to 4 h, are resistant to crystallization, meet all current US Department of Energy requirements for chemical durability, and have a solubility limit for Cr 2 O 3 which is at least three times larger than that for borosilicate glasses. (orig.)

  10. Development of containment system for application to decommissioning of nuclear facilities. 2

    International Nuclear Information System (INIS)

    Mizuno, Oichi; Iwasaki, Yukio; Miyao, Hidehiko; Uchikoshi, Tadaaki; Furuya, Hirotaka; Kamata, Hirofumi

    1998-01-01

    New greenhouse/containment was developed to apply for confining the radioactive materials and preventing the dispersion of radioactive contamination during the maintenance and dismantling of the nuclear facility. The pressed-air tubes for columns and beams of the structure frame were applied and the vinyl chloride sheet reinforced with the polyester fiber was used as a canvas of wall and ceiling. This canvas is possible to use repeatedly and has high efficiency of safe enclosing. The containment can be easily assembled and disassembled by charge and discharge of the pressed air in the tubes of columns and beams. Two standard units (2.5mL x 2.5mW x 2.5mH, 5mL x 5mW x 2.5mH) are prepared, and lateral connection of these standard units makes it applicable to the wide working area. Expansion model up to 5m in height is also available. (author)

  11. Selected safety aspects of containments for nuclear power plants with WWER-440 reactors

    International Nuclear Information System (INIS)

    Jankowski, M.W.; Kulig, M.J.; Strupczewski, A.; Balabanov, E.D.

    1996-01-01

    Considerable attention has been and continues to be focused on the design and operational features that prevent the release of radioactive materials to the environment for a spectrum of accidents for the two classes of WWER-440 reactors: the older 230 model and the more recently designed 213 models. This paper, based on published and unpublished information, aims to clarify the perceptions of the Russian WWER-440 models 230 and 213 nuclear power plant containment system designs and their relevance to selected aspects of accident mitigation. It should be noted that these are unclearly and often negatively perceived, primarily because of a lack of reliable information and a poorly assembled experimental database. Conflicting statements have been made regarding the nature and the features of the plant's containment system. The paper presents a brief outline of the design of both WWER-440 models with respect to their confinement functions. Selected safety-related aspects of the accident localization systems are discussed, and the recognized shortcomings and safety merits are pointed out. The older 230 units experience high leak rates and are designed to withstand medium-size pipe breaks. The possible implications for safety are pointed out in the paper. The on going studies that concentrate on improving the system are highlighted. (orig.)

  12. Failure by void coalescence in metallic materials containing primary and secondary voids subject to intense shearing

    DEFF Research Database (Denmark)

    Nielsen, Kim Lau; Tvergaard, Viggo

    2011-01-01

    Failure under intense shearing at close to zero stress triaxiality is widely observed for ductile metallic materials, and is identified in experiments as smeared-out dimples on the fracture surface. Numerical cell-model studies of equal sized voids have revealed that the mechanism governing...... this shear failure mode boils down to the interaction between primary voids which rotate and elongate until coalescence occurs under severe plastic deformation of the internal ligaments. The objective of this paper is to analyze this failure mechanism of primary voids and to study the effect of smaller...... secondary damage that co-exists with or nucleation in the ligaments between larger voids that coalesce during intense shearing. A numerical cell-model study is carried out to gain a parametric understanding of the overall material response for different initial conditions of the two void populations...

  13. Primary breast cancer tumours contain high amounts of IgA1 immunoglobulin

    DEFF Research Database (Denmark)

    Welinder, Charlotte; Baldetorp, Bo; Blixt, Klas Ola

    2013-01-01

    seen in the percentage of stained cells and in the staining pattern in the different breast cancers analysed. Anti-Tn antibody and HPA were also shown to specifically bind to a number of possible constellations of the Tn antigen in the hinge region of IgA1. Both reagents could also detect the presence....... The short O-glycan that forms the antigen is carried by a number of different proteins. One potential carrier of the Tn antigen is immunoglobulin A1 (IgA1), which we surprisingly found in tumour cells of the invasive parts of primary breast carcinoma. Conventional immunohistochemical analysis of paraffin......-embedded sections from primary breast cancers showed IgA1 to be present in the cytoplasm and plasma membrane of 35 out of 36 individual primary tumours. The immunohistochemical staining of HPA and anti-Tn antibody (GOD3-2C4) did to some extent overlap with the presence of IgA1 in the tumours, but differences were...

  14. Corrosion of steel drums containing cemented ion-exchange resins as intermediate level nuclear waste

    Science.gov (United States)

    Duffó, G. S.; Farina, S. B.; Schulz, F. M.

    2013-07-01

    Exhausted ion-exchange resins used in nuclear reactors are immobilized by cementation before being stored. They are contained in steel drums that may undergo internal corrosion depending on the presence of certain contaminants. The objective of this work is to evaluate the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins with different aggressive species. The corrosion potential and the corrosion rate of the steel, and the electrical resistivity of the matrix were monitored for 900 days. Results show that the cementation of ion-exchange resins seems not to pose special risks regarding the corrosion of the steel drums. The corrosion rate of the steel in contact with cemented ion-exchange resins in the absence of contaminants or in the presence of 2.3 wt.% sulphate content remains low (less than 0.1 μm/year) during the whole period of the study (900 days). The presence of chloride ions increases the corrosion rate of the steel at the beginning of the exposure but, after 1 year, the corrosion rate drops abruptly reaching a value close to 0.1 μm/year. This is probably due to the lack of water to sustain the corrosion process. When applying the results obtained in the present work to estimate the corrosion depth of the steel drums containing the cemented radioactive waste after a period of 300 years, it is found that in the most unfavourable case (high chloride contamination), the corrosion penetration will be considerably lower than the thickness of the wall of the steel drums. Cementation of ion-exchange resins does not seem to pose special risks regarding the corrosion of the steel drums that contained them; even in the case the matrix is highly contaminated with chloride ions.

  15. GASFLOW simulations for containment venting system of Emsland nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Zhanjie; Jordan, Thomas

    2015-07-01

    In a hypothetical core melt severe accident in nuclear power plants, a large amount of burnable gas, such as hydrogen and carbon monoxide, with certain amount of oxygen is released into the containment. In case of high containment pressure a filtered venting is designed as a mitigation measure. After aerosols and condensable portions of the gas mixture from the containment are purged, the remaining burnable gases flows into a chamber connected to horizontal venting pipes and further to a vertical stack, which is open to free atmosphere at the top. The main goal of the study is to investigate the chemical sensitivity of the burnable and potentially detonable gas mixture in the venting system by means of computational fluid dynamic computer simulations. Based on the calculations with different boundary conditions for different scenarios, it is concluded that the venting system has no risk of flame acceleration and detonation if the power supply of the ventilation fans can be maintained. In case of station blackout, the whole system has no risk of detonation if the scrubber - purging water pool - can get boiling in 10 minutes, owing to the heating effect resulted from the latent heat of the condensable gases into the water pool. It is because the injected steam from the boiling scrubber inerts the mixture of hydrogen, carbon monoxide plus oxygen, effectively and significantly. In this case, a gas mixture only in a very local region in the stack, as small as about 0.16 cubic meter, is in a risk of flame acceleration. It is in principle ignorable owing to the tiny volume and thus only a small amount of hydrogen and carbon monoxide.

  16. A Remote-operated System to Map Radiation Dose in the Fukushima Daiichi Primary Containment Vessel

    Science.gov (United States)

    Nancekievill, M.; Jones, A. R.; Joyce, M. J.; Lennox, B.; Watson, S.; Katakura, J.; Okumura, K.; Kamada, S.; Katoh, M.; Nishimura, K.

    2018-01-01

    This paper describes the development of a submersible system based on a remote-operated vehicle coupled with radiation detectors to map the interior of the reactors at the Fukushima Daiichi nuclear power station. It has the aim oflocating fuel debris. The AVEXIS submersible vehicle used in this study has been designed as a low-cost, potentially disposable, inspection platform that is the smallest of its class and is capable of being deployed through a 150 mm diameter access pipe. To map the gamma-ray environment, a cerium bromide scintillator detector with a small form factor has been incorporated into the AVEXIS to identify radioactive isotopes via gamma-ray spectroscopy. This provides the combined system with the potential to map gamma-ray spectra and particle locations throughout submerged, contaminated facilities, such as Units 1, 2 and 3 of the Fukushima Daiichi nuclear power plant. The hypothesis of this research is to determine the sensitivity of the combined system in a submerged environment that replicates the combination of gamma radiation and water submersion but at lower dose rates.

  17. Effects of container material on PCT leach test results for high-level nuclear waste glasses

    International Nuclear Information System (INIS)

    Xing, S.B.; Pegg, I.L.

    1994-01-01

    A glass-based waste form used for the immobilization of high-level nuclear wastes should exhibit good resistance to aqueous corrosion since typically this is the primary process by which radionucleides could be released into the environment upon failure of other barriers. In the USA, the Waste Acceptance Product Specifications (WAPS) provides a set of requirements to ensure the consistency of the waste forms produced and specifies the Product Consistency Test (PCT) as a measure of relative chemical durability. While the PCT procedure permits usage of both Teflon and stainless steel vessels for testing of simulated development glasses, Teflon is not permitted for testing of production glasses due to radiative degradation. The results presented in this paper indicate that there are very significant differences between tests conducted in the two types of vessels due to the well-known permeability of Teflon to atmospheric carbon dioxide which results in lowering of the solution pH and a consequent reduction in the leach rate of silicate glasses. A wide range of nuclear waste glass compositions was subjected to the PCT procedure using both Teflon and stainless steel vessels. The magnitude of the effect (up to a factor of four for B, Na, Li concentrations) depends strongly on glass composition, therefore the isolated checks performed previously were inconclusive. The permeability to CO, of two types of Teflon vessels specified in the PCT procedure was directly measured using buffer solutions: ingress of CO, is linear in time, strongly pH-dependent, and was as high as 100 ppm after 7 days. In actual PCT tests in Teflon vessels, the total CO, content was 560 ppm after 87 days and 1930 ppm after one year

  18. Methodology of containment response analysis for the Probabilistic Safety Assessment -PSA of the CAREM-25 nuclear power plant

    International Nuclear Information System (INIS)

    Baron, Jorge

    1996-01-01

    This work is part of the Probabilistic Safety Assessment actually under development for the CAREM-25 Nuclear Power Station, and departs from the accident sequences already obtained and quantified by the Event Trees/Fault Trees techniques. At first, the potential containment failure modes for nuclear stations are listed, based on the experience. Then, the CAREM-25 design peculiarities are analyzed, on their possible influence on the containment behavior during, severe accidents. Then Plan Damage States are then defined. Furthermore, Containment Damage States are also defined, and Containment Event Trees are built for each Plant Damage State. Those sequences considered representative from the annual probability (those which exceed or equal a probability of 1E-09 per year, are used to quantify the combinations of Plant Damage States/Containment Damage States, based on the estimation of a Vulnerability Matrix. (author)

  19. Total cross-sections assessment of neutron reaction with stainless steel SUS-310 contained in various nuclear data files

    International Nuclear Information System (INIS)

    Suwoto

    2002-01-01

    The integral testing of neutron cross-sections for Stainless Steel SUS-310 contained in various nuclear data files have been performed. The shielding benchmark calculations for Stainless Steel SUS-310 has been analysed through ORNL-Broomstick Experiment calculation which performed by MAERKER, R.E. at ORNL - USA ( 1) . Assessment with JENDL-3.1, JENDL-3.2, ENDF/B-IV, ENDF/B-VI nuclear data files and data from GEEL have also been carried out. The overall calculation results SUS-310 show in a good agreement with the experimental data, although, underestimate results appear below 3 MeV for all nuclear data files. These underestimation tendencies clearly caused by presented of iron nuclide which more than half in Stainless Steel compound. The total neutron cross-sections of iron nuclide contained in various nuclear data files relatively lower on that energy ranges

  20. Management of primary-to-secondary leaks at Loviisa nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Mohnsen, B.; Jaenkaelae, K. [IVO International Ltd, Vantaa (Finland)

    1995-12-31

    The Loviisa Nuclear power plant consisting of two VVER-440 type press water reactor units has been in commercial operation since the late 1970`s. Specific features for VVER-440 reactors are six primary loops with horizontal steam generators and main gate valves. The structure of the horizontal steam generators construction may cause a large primary to secondary leak in case of a break in the cover of the primary collector. An accident where two primary collector covers opened totally and two covers opened partly took place in Rovno, Ukraine January 1982. Primary to secondary leaks are one of the main contributors to the core melt frequency in VVER reactors according to the Loviisa 1 Probabilistic Safety Assessment. The high core damage contribution has set requirements for the development of effective means to cope with all sizes of primary to secondary leaks in the steam generator. A concept for all leak sizes has been developed for Loviisa 1 and 2. The solution includes four main areas which are a new steam generator leakage monitoring system based on nitrogen-16 measurement, an upgraded pressurizer spray system, an increased emergency cooling water reserve and an automated isolation of the defected steam generator.

  1. Management of primary-to-secondary leaks at Loviisa nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Mohnsen, B; Jaenkaelae, K [IVO International Ltd, Vantaa (Finland)

    1996-12-31

    The Loviisa Nuclear power plant consisting of two VVER-440 type press water reactor units has been in commercial operation since the late 1970`s. Specific features for VVER-440 reactors are six primary loops with horizontal steam generators and main gate valves. The structure of the horizontal steam generators construction may cause a large primary to secondary leak in case of a break in the cover of the primary collector. An accident where two primary collector covers opened totally and two covers opened partly took place in Rovno, Ukraine January 1982. Primary to secondary leaks are one of the main contributors to the core melt frequency in VVER reactors according to the Loviisa 1 Probabilistic Safety Assessment. The high core damage contribution has set requirements for the development of effective means to cope with all sizes of primary to secondary leaks in the steam generator. A concept for all leak sizes has been developed for Loviisa 1 and 2. The solution includes four main areas which are a new steam generator leakage monitoring system based on nitrogen-16 measurement, an upgraded pressurizer spray system, an increased emergency cooling water reserve and an automated isolation of the defected steam generator.

  2. Management of primary-to-secondary leaks at Loviisa nuclear power plant

    International Nuclear Information System (INIS)

    Mohnsen, B.; Jaenkaelae, K.

    1995-01-01

    The Loviisa Nuclear power plant consisting of two VVER-440 type press water reactor units has been in commercial operation since the late 1970's. Specific features for VVER-440 reactors are six primary loops with horizontal steam generators and main gate valves. The structure of the horizontal steam generators construction may cause a large primary to secondary leak in case of a break in the cover of the primary collector. An accident where two primary collector covers opened totally and two covers opened partly took place in Rovno, Ukraine January 1982. Primary to secondary leaks are one of the main contributors to the core melt frequency in VVER reactors according to the Loviisa 1 Probabilistic Safety Assessment. The high core damage contribution has set requirements for the development of effective means to cope with all sizes of primary to secondary leaks in the steam generator. A concept for all leak sizes has been developed for Loviisa 1 and 2. The solution includes four main areas which are a new steam generator leakage monitoring system based on nitrogen-16 measurement, an upgraded pressurizer spray system, an increased emergency cooling water reserve and an automated isolation of the defected steam generator

  3. Development of filtered containment venting system and application for Kashiwazaki-Kariwa Nuclear Power Station Unit 6, 7

    International Nuclear Information System (INIS)

    Murai, Soutarou; Hiranuma, Naoki; Kimura, Takeo; Omori, Shuichi; Watanabe, Fumitoshi; Sasa, Daisuke

    2014-01-01

    The Fukushima Dai-ichi Nuclear Power Station (1F) of Tokyo Electric Power Company (TEPCO) had experienced severe radio-active release to the environment in the Tohoku Region Pacific Coast Earthquake (alias: the Great East Japan Earthquake) in 2011. Under the Station Black-Out (SBO) conditions caused by tsunami with the earthquake, the 1F operators had tried to vent the gasses in the Primary Containment Vessels (PCVs) of the unit 1, 2 and 3 to the environment through the water pools in the suppression chambers of the PCVs. Its venting, however, was imperfect and, as a result, major direct radio-active release to the environment was caused. After this disaster, TEPCO launched a project to develop the Filtered Containment Venting System (FCVS), in which our very bitter experiences in the 1F accident as described above are reflected. One of the main purposes of the development of the FCVS is to enhance operability of venting under the severe plant conditions such as the SBO during progressing of severe core damage, and another is to enhance removal performance of radio-nuclides with the newly added filtering equipment, which is installed in the venting line from the PCV to the outer. The Kashiwazaki-Kariwa NPS unit 6 and 7 will be the first reactors applied the FCVSs. In this paper, we show the design concept of the TEPCO's FCVS, the brief overview of the system design and the summary of experiment which has been performed for getting the performance data of the FCVS such as decontamination factor in various conditions. (author)

  4. Fuel containment and damage tolerance for large composite primary aircraft structures. Phase 1: Testing

    Science.gov (United States)

    Sandifer, J. P.

    1983-01-01

    Technical problems associated with fuel containment and damage tolerance of composite material wings for transport aircraft were identified. The major tasks are the following: (1) the preliminary design of damage tolerant wing surface using composite materials; (2) the evaluation of fuel sealing and lightning protection methods for a composite material wing; and (3) an experimental investigation of the damage tolerant characteristics of toughened resin graphite/epoxy materials. The test results, the test techniques, and the test data are presented.

  5. Leaching assessment of road materials containing primary lead and zinc slags.

    Science.gov (United States)

    Barna, R; Moszkowicz, P; Gervais, C

    2004-01-01

    Characterisation of the leaching behaviour of waste-containing materials is a crucial step in the environmental assessment for reuse scenarios. In our research we applied the multi-step European methodology ENV 12-920 to the leaching assessment of road materials containing metallurgical slag. A Zn slag from an imperial smelting furnace (ISF) and a Pb slag from a lead blast furnace (LBF) are investigated. The two slags contain up to 11.2 wt% of lead and 3.5 wt% of zinc and were introduced as a partial substitute for sand in two road materials, namely sand-cement and sand-bitumen. At the laboratory scale, a leaching assessment was performed first through batch equilibrium leaching tests. Second, the release rate of the contaminants was evaluated using saturated leaching tests on monolithic material. Third, laboratory tests were conducted on monolithic samples under intermittent wetting conditions. Pilot-scale tests were conducted for field testing of intermittent wetting conditions. The results show that the release of Pb and Zn from the materials in a saturated scenario was controlled by the pH of the leachates. For the intermittent wetting conditions, an additional factor, blocking of the pores by precipitation during the drying phase is proposed. Pilot-scale leaching behaviour only partially matched with the laboratory-scale test results: new mass transfer mechanisms and adapted laboratory leaching tests are discussed.

  6. Leak detection system with distributed microprocessor in the primary containment vessel

    International Nuclear Information System (INIS)

    Inahara, K.; Yoshioka, K.; Tomizawa, T.

    1980-01-01

    Responding to the demand for greater improvements of the safety monitoring system, less public radiation exposure, and increase of plant availability, measuring and control systems in nuclear power plants have undergone many improvements. Leak detection systems are also required to give earlier warning, additional accuracy, and continuous monitoring function. This paper describes the drywell sump leakage detection system utilizing a distributed microprocessor, which is a successful application owing to its versatile function and ease of installation. The microprocessor performs various functions such as a rate of level change computation, conversion to leakage flow rate, initiation of alarm, and sump pump control. This system has already been applied to three operating BWR plants that demonstrate its efficiency. (auth)

  7. Isotope distributions in primary heat transport and containment systems during a severe accident in CANDU type reactor

    International Nuclear Information System (INIS)

    Constantin, M.

    2005-01-01

    The paper is intended to analyse the distribution of the fission products (FPs) in CANDU Primary Heat Transport (PHT) and CANDU Containment Systems by using the ASTEC code. The complexity of the data required by ASTEC and the complexity both of CANDU PHT and Containment System were strong motivation to begin with a simplified model. The data related to the nodes' definitions, temperatures and pressure conditions were chosen as possible as real data from CANDU loss of coolant accident sequence (CATHENA code results). The source term of FPs introduced into the PHT was estimated by ORIGEN code. The FPs distribution in the nodes of the circuit and the FPs mass transfer per isotope and chemical species were obtained by using SOPHAEROS module. The distributions within the containment are obtained by the CPA module (thermalhydraulic calculations in the containment and FPs aerosol transport). The results consist of mass distributions in the nodes of the circuit and the transferred mass to the containment through the break for different species (FPs and chemical species) and mass distributions in the different parts of the containment and different hosts. (authors)

  8. 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.)

  9. Aircraft crash upon a containment structure of a nuclear power plant

    International Nuclear Information System (INIS)

    Paul, D.K.; Abbas, H.; Godbole, P.N.; Nayak, G.C.

    1993-01-01

    The reinforced concrete outer containment of a Nuclear Power Plant (NPP) is required to be designed to withstand the impact of aircraft and aircrash debris etc. The problem is of strategic significance because the damage caused to the structure by these missiles may lead to the leakage of nuclear radiations. The safety design of NPP against aircraft crash requires the evaluation of crash probability. If the probability is smaller than the allowable value then the aircraft crash is neglected as design basis item. Otherwise adequate measures are taken to bring the released radioactive material within the permissible limit. The aircrafts and their striking velocity to be considered in the design of a structure are decided by the accident analysis. The probabilistic aspect of the problem has not been covered in the present work. The non-affordability of coupled analysis of large problems like aircraft crash on NPP, automobile impact on a structure etc. due to the requirement of excessive amount of manual as well as computer time and storage compels us to switch over to the uncoupled analysis. Moreover, the results of coupled analysis are heavily influenced by the analyst's modelling technique and choice of increment size. It is uncoupling of the missile and the target which converts the impact load to the impulse load. This impulse can be found by taking into consideration only the inertial and stiffness properties of missile and considering the target to be rigid. Though the impulse load so obtained disregards the inertial and stiffness characteristics of the target but its effect can be incorporated by modifying it for the inertial and stiffness properties of target at different time steps as we march in time domain during the analysis of the target

  10. PARIS project: Radiolytic oxidation of molecular iodine in containment during a nuclear reactor severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Bosland, L. [Institut de Radioprotection et de Surete Nucleaire, DPAM, SEMIC, LETR - CEN Cadarache, BP 3, 13115 Saint Paul Lez Durance (France)], E-mail: loic.bosland@irsn.fr; Funke, F. [AREVA NP GmbH, Technical Center, P.O. Box 1109, D-91001 Erlangen (Germany); Girault, N. [Institut de Radioprotection et de Surete Nucleaire, DPAM, SEMIC, LETR - CEN Cadarache, BP 3, 13115 Saint Paul Lez Durance (France); Langrock, G. [AREVA NP GmbH, Technical Center, P.O. Box 1109, D-91001 Erlangen (Germany)

    2008-12-15

    In case of a hypothetical severe accident in a nuclear LWR (light water reactor), the high radiation fields reached in the reactor containment building due to the release of fission products from the reactor core could induce air radiolysis. The air radiolysis products could, in turn, oxidise gaseous molecular iodine into aerosol-borne iodine-oxygen-nitrogen compounds. Thereby, this reaction involves a change of iodine speciation and a decrease of iodine volatility in the reactor containment atmosphere. Kinetic data were produced within the PARIS project on the air radiolysis products formation and destruction, and on their reaction with molecular iodine, with the objective of developing and validating existing kinetic models. The current paper includes the non-iodine tests of the PARIS project whose objective was to determine the rates of formation and destruction of air radiolysis products in the presence of both structural containment surfaces (decontamination coating ('paint') and stainless steel), aerosol particles such as silver rich particles (issued from the control rods) in boundary conditions representative for LWR or PHEBUS facility containments. It is found that the air radiolysis products concentration increases with dose and tend to approach saturation levels at doses higher than about 1 kGy. This behaviour is more evident in oxygen/steam atmospheres, producing ozone, than in air/30% (v/v) steam atmospheres, the latter favouring the model-predicted on-going production of nitrogen dioxide even at very high doses. No significant effect of temperature, dose rate and hydrogen addition (4%, v/v) was observed. Furthermore, the inserted surfaces do not exhibit significant effects on the air radiolysis concentrations. However, these 'non-noticeable influence' could be due to a masking of small effects by the appreciable scattering of the experimental air radiolysis product concentrations. The PARIS results are then analysed using two

  11. PARIS project: Radiolytic oxidation of molecular iodine in containment during a nuclear reactor severe accident

    International Nuclear Information System (INIS)

    Bosland, L.; Funke, F.; Girault, N.; Langrock, G.

    2008-01-01

    In case of a hypothetical severe accident in a nuclear LWR (light water reactor), the high radiation fields reached in the reactor containment building due to the release of fission products from the reactor core could induce air radiolysis. The air radiolysis products could, in turn, oxidise gaseous molecular iodine into aerosol-borne iodine-oxygen-nitrogen compounds. Thereby, this reaction involves a change of iodine speciation and a decrease of iodine volatility in the reactor containment atmosphere. Kinetic data were produced within the PARIS project on the air radiolysis products formation and destruction, and on their reaction with molecular iodine, with the objective of developing and validating existing kinetic models. The current paper includes the non-iodine tests of the PARIS project whose objective was to determine the rates of formation and destruction of air radiolysis products in the presence of both structural containment surfaces (decontamination coating ('paint') and stainless steel), aerosol particles such as silver rich particles (issued from the control rods) in boundary conditions representative for LWR or PHEBUS facility containments. It is found that the air radiolysis products concentration increases with dose and tend to approach saturation levels at doses higher than about 1 kGy. This behaviour is more evident in oxygen/steam atmospheres, producing ozone, than in air/30% (v/v) steam atmospheres, the latter favouring the model-predicted on-going production of nitrogen dioxide even at very high doses. No significant effect of temperature, dose rate and hydrogen addition (4%, v/v) was observed. Furthermore, the inserted surfaces do not exhibit significant effects on the air radiolysis concentrations. However, these 'non-noticeable influence' could be due to a masking of small effects by the appreciable scattering of the experimental air radiolysis product concentrations. The PARIS results are then analysed using two different kinetic models

  12. DEVELOPING AND QUANTIFYING PARAMETERS FOR CLOSURE WELDING OVERPACKS CONTAINING RESEARCH REACTOR SPENT NUCLEAR FUEL AT HANFORD

    International Nuclear Information System (INIS)

    CANNELL GR

    2007-01-01

    Fluor engineers developed a Gas Tungsten Arc Welding (GTAW) technique and parameters, demonstrated requisite weld quality and successfully closure-welded packaged spent nuclear fuel (SNF) overpacks at the Hanford Site. This paper reviews weld development and qualification activities associated with the overpack closure-welding and provides a summary of the production campaign. The primary requirement of the closure weld is to provide leaktight confinement of the packaged material against release to the environment during interim storage (40-year design term). Required weld quality, in this case, was established through up-front development and qualification, and then verification of parameter compliance during production welding. This approach was implemented to allow for a simpler overpack design and more efficient production operations than possible with approaches using routine post-weld testing and nondestructive examination (NDE). . A series of welding trials were conducted to establish the desired welding technique and parameters. Qualification of the process included statistical evaluation and American Society of Mechanical Engineers (ASME) Section IX testing. In addition, pull testing with a weighted mockup, and thermal calculation/physical testing to identify the maximum temperature the packaged contents would be subject to during welding, was performed. Thirteen overpacks were successfully packaged and placed into interim storage. The closure-welding development activities (including pull testing and thermal analysis) provided the needed confidence that the packaged SNF overpacks could be safely handled and placed into interim storage, and remain leaktight for the duration of the storage term

  13. The Flooding Water Source Analysis following the Feed Line Break at the Compartment outside Containment for Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Song, Dong Soo [Korea Electric Power Research Institute, Daejeon (Korea, Republic of); Park, Young Chan [ACT, Daejeon (Korea, Republic of)

    2007-07-01

    The Periodic Safety Review(PSR) has been performing for the operating nuclear power plant in Korea. One of the PSR evaluation items is environmental qualification. Flooding issue for nuclear power plants designed and built in 1970 is extremely severe for main steam header compartment and main feed water line region of intermediate building and lower floor. This study presents to analyze flood level of feed water line breaks for the Westinghouse nuclear power plant. This analyses provides the mass and energy releases using the developed methodology for a break outside containment. For the analyses RETRAN-3D computer program is used.

  14. Research programme related to the influence of wind on contamination containment in nuclear installations

    Energy Technology Data Exchange (ETDEWEB)

    Soares, S.; Gelain, T.; Laborde, J.C.; Ricciardi, L. [Institut de Radioprotection et de Surete Nucleaire (IRSN/DSU/SERAC/LEMAC), 91 - Gif-sur-Yvette (France)

    2006-07-01

    In nuclear industry, the response of a ventilation network to accidental disturbances, either mechanical (fan failure, damper blockage,..) or thermal (fire..) is difficult to evaluate when the network becomes complex. In order to determine and analyze the consequences of these disturbances on the radioactive materials containment, a computer code called SIMEVENT has been developed. However, among the external parameters likely to affect a ventilation network, the wind effect is actually basically modeled, due to a lack of qualified data concerning the wind impact on complex building geometries and the interaction between wind and chimney exhaust. In view of the networks complexity and the facilities diversity, a research program including experimental and model studies has been launched to assess the wind influence on contamination containment. 1. step: improvement of data (2005-2006): The diversity of facilities geometries needs the use of a qualified multi-D code for pressure coefficients Cpi assessment, characterizing the wind effect on building walls and the interaction between wind and chimney exhaust. Different chimney terminals have then been placed in a wind tunnel (the parameters are the incline angle a, the wind velocity U and the air flow in the duct W); for each angle, the evolution of the pressure coefficient versus wind velocity is determined and is characteristic of a chimney terminal geometry. Furthermore, two types of scale-model have been chosen for representing either nuclear power plants (NPP) or plants and laboratories buildings. The different values of wind pressure coefficients have been measured on both scale-models placed in a wind-tunnel. The experimental data obtained are compared with CFD simulations (CFX code), in order to qualify such code for the assessment of pressure coefficient on complex geometries. The results are quite encouraging. 2. step - wind tunnel tests on a ventilated scale model (2007-2009): Wind tunnel tests will be

  15. Research programme related to the influence of wind on contamination containment in nuclear installations

    International Nuclear Information System (INIS)

    Soares, S.; Gelain, T.; Laborde, J.C.; Ricciardi, L.

    2006-01-01

    In nuclear industry, the response of a ventilation network to accidental disturbances, either mechanical (fan failure, damper blockage,..) or thermal (fire..) is difficult to evaluate when the network becomes complex. In order to determine and analyze the consequences of these disturbances on the radioactive materials containment, a computer code called SIMEVENT has been developed. However, among the external parameters likely to affect a ventilation network, the wind effect is actually basically modeled, due to a lack of qualified data concerning the wind impact on complex building geometries and the interaction between wind and chimney exhaust. In view of the networks complexity and the facilities diversity, a research program including experimental and model studies has been launched to assess the wind influence on contamination containment. 1. step: improvement of data (2005-2006): The diversity of facilities geometries needs the use of a qualified multi-D code for pressure coefficients Cpi assessment, characterizing the wind effect on building walls and the interaction between wind and chimney exhaust. Different chimney terminals have then been placed in a wind tunnel (the parameters are the incline angle a, the wind velocity U and the air flow in the duct W); for each angle, the evolution of the pressure coefficient versus wind velocity is determined and is characteristic of a chimney terminal geometry. Furthermore, two types of scale-model have been chosen for representing either nuclear power plants (NPP) or plants and laboratories buildings. The different values of wind pressure coefficients have been measured on both scale-models placed in a wind-tunnel. The experimental data obtained are compared with CFD simulations (CFX code), in order to qualify such code for the assessment of pressure coefficient on complex geometries. The results are quite encouraging. 2. step - wind tunnel tests on a ventilated scale model (2007-2009): Wind tunnel tests will be

  16. Structural response of nuclear containment shield buildings with unanticipated construction openings

    Science.gov (United States)

    Mac Namara, Sinead Caitriona

    As Nuclear Power Plants age many require steam generator replacement. There is a nickel alloy in the steam generator tubes that is susceptible to stress cracking and although these cracks can be sealed the generator becomes uneconomical without 10%-15% of the tubes. The steam generator in a typical nuclear power plant is housed in the containment structure next to the reactor. The equipment hatch is not big enough to facilitate steam generator replacement, thus construction openings in the dome of the containment structure are required. To date the structural consequences of construction openings in the dome have not been examined. This thesis examines the effects of such openings. The prototype concrete dome is made up of a 2 ft thick dome atop 3 ft thick and 170 ft high cylindrical walls (radius 65.5 ft) with a tension ring 15 ft high and 8 ft thick in between. The dome of the building is cast in two layers; a lower 9 inch layer that serves as the formwork for an upper 15 inch layer. The weight of the dome is carried in axial compression along the hoops and meridians of the dome. The first finite element model uses shell elements and considers two limiting load cases; where the two layers act as one, and where the lower layer carries the weight of both. The openings interrupt the hoops and meridians and the weight of the dome must be redistributed around the openings. Without openings, the stresses due to dead load in the structure are very low when compared to the material strength. The impact of the openings is increased compression stresses near the opening. The maximum stresses are approximately four times larger than in the original structure. These results are confirmed by the second model which is made from layers of solid elements. This model shows a significant difference between the compression on the top surface of the dome, in the affected areas, and that on the bottom surface, leading to shear stresses. These shear stresses are largest around the

  17. Analysis of primary containment response using an arbitrary Langrangian-Eulerian method

    Energy Technology Data Exchange (ETDEWEB)

    Chungyi, Wang

    1988-02-01

    This paper describes an advanced arbitrary Lagrangian-Eulerian method and its associated computer program, ALICE-II, for calculating the response of liquid metal reactor containment to core energetics. It is a versatile numerical algorithm with more flexibility and efficiency than other methods used to treat reactor containment with complex internals such as internal thin shells, upper internal structures, perforated plates, core-support diagrid, shield baffles, and deflector plates. The algorithm uses a two-dimensional, hybrid Lagrangian-Eulerian finite-difference technique to calculate the hydrodynamics and fluid-structure interactions, together with a purely Eulerian finite-difference approach to analyze the free-surface and material interface motions. It has significant advantages in treating complex phenomena such as flow through perforated structures, large material distortions, multi-dimensional sliding interfaces, flow around corners, highly contorted fluid boundaries, outflow boundary conditions, and coolant spillage. Numerical calculations for the hydrodynamic solutions are separated into three phases. The first phase consists of an explicit Lagrangian calculation. The second phase, which is options, contains an implicit iteration. The third phase, which is also optional, rezones the mesh vertices to prescribed positions. The structural response is computed by a library of elastic-plastic elements formulated in corotational coordinates in conjunction with an explicit time-integration scheme. Interaction between fluid and structure is accounted for by rigorously enforcing the interface boundary conditions. Many sample problems are given to illustrate the code effectiveness. Results demonstrate that these complex fluid-structure interaction problems can be analyzed with the ALICE-II code in a relatively natural and straightforward manner.

  18. Studies of heat transfer having relevance to nuclear reactor containment cooling by buoyancy-driven air flow

    Energy Technology Data Exchange (ETDEWEB)

    Jackson, J. D.; Li, J.; Wang, J. [The Univ., of Manchester, Manchester (United Kingdom)

    2003-07-01

    Two separate effects experiments concerned with buoyancy-influenced convective heat transfer in vertical passages which have relevance to the problem of nuclear reactor containment cooling by means of buoyancy-driven airflow are described. A feature of each is that local values of heat transfer coefficient are determined on surfaces maintained at uniform temperature. Experimental results are presented which highlight the need for buoyancy-induced impairment of turbulent convective heat transfer to be accounted for in the design of such passive cooling systems. A strategy is presented for predicting the heat removal by combined convective and radiative heat transfer from a full scale nuclear reactor containment shell using such experimental results.

  19. Development of containers sealing system like part of surveillance program of the vessel in nuclear power plants

    International Nuclear Information System (INIS)

    Romero C, J.; Hernandez C, R.; Fernandez T, F.; Rocamontes A, M.; Perez R, N.

    2009-10-01

    The owners of nuclear power plants should be demonstrate that the embrittlement effects by neutronic radiation do not commit the structural integrity from the pressure vessel of nuclear reactors, during conditions of routine operation and below postulate accident. For this reason, there are surveillance programs of vessels of nuclear power plants, in which are present surveillance capsules. A surveillance capsule is compound by the support, six containers for test tubes and dosimeters. The containers for test tubes are of two types: rectangular container for test tubes, Charpy V and Cylindrical Container for tension test tubes. These test tubes are subject to a same or bigger neutronic flow to that of vessel, being representative of vessel mechanical conditions. The test tubes are rehearsed to watch over the increase of embrittlement that presents the vessel. This work describes the development of welding system to seal the containers for test tubes, these should be filled with helium of ultra high purity, to a pressure of an atmosphere. In this system the welding process Gas Tungsten Arc Welding is used, a hermetic camera that allows to place the containers with three grades of freedom, a vacuum subsystem and pressure, high technology equipment's like: power source with integrated computer, arc starter of high frequency, helium flow controller, among others. Finally, the advances in the inspection system for the qualification of sealing system are mentioned, system that should measure the internal pressure of containers and the helium purity inside these. (Author)

  20. Summary Report of the Technical Meeting on Primary Radiation Damage: From Nuclear Reaction to Point Defects

    International Nuclear Information System (INIS)

    Stoller, R. E.; Nordlund, K.; Simakov, S.P.

    2012-11-01

    The Meeting was convened to bring together the experts from both the nuclear data and materials research communities because of their common objective of accurately characterizing irradiation environments and resulting material damage. The meeting demonstrated that significant uncertainties remain regarding both the status of nuclear data and the use of these data by the materials modeling community to determine the primary damage state obtained in irradiated materials. At the conclusion of the meeting, the participants agreed that there is clear motivation to initiate a CRP that engages participants from the nuclear data and materials research communities. The overall objective of this CRP would be to determine the best possible parameter (or a few parameters) for correlating damage from irradiation facilities with very different particle types and energy spectra, including fission and fusion reactors, charged particle accelerators, and spallation irradiation facilities. Regarding progress achieved during the last decade in the atomistic simulation of primary defects in crystalline materials, one of the essential and quantitative outcomes from the CRP is expected to be cross sections for point defects left after recoil cascade quenching. (author)