WorldWideScience

Sample records for nuclear containments confinements

  1. Confinement contains condensates

    DEFF Research Database (Denmark)

    Brodsky, S. J.; Roberts, C. D.; Shrock, R.

    2012-01-01

    Dynamical chiral symmetry breaking and its connection to the generation of hadron masses has historically been viewed as a vacuum phenomenon. We argue that confinement makes such a position untenable. If quark-hadron duality is a reality in QCD, then condensates, those quantities that have common...... evidence and incidentally expose misconceptions in a recent Comment....

  2. Reinforced confinement in a nuclear reactor

    International Nuclear Information System (INIS)

    Norman, H.

    1988-01-01

    The present invention concerns a nuclear reactor containing a reactor core, a swimming pool space that is filled and pressurized with a neutron-absorbing solution, a reactor tank, at least one heat exchanger, at least one inlet line, at least one return line and at least one circulation pump, where the said reactor tank is confined in the said swimming pool space and designed to be cooled with the aid of relatively pure water, which is fed by means of the said at least one circulating pump to the said reactor tank from the said heat exchanger via the said at least one inlet line and is returned to the heat exchanger via the said at least one return line. The problem that is to be solved by the invention is to design a reactor of the above type in such a way that a complete confinement of the primary circuit of the reactor is achieved at relatively low extra cost. This problem is solved by providing the reactor with a special confinement space that confines the heat exchanger, but not the reactor tank, with the confinement space and the swimming pool space being fashioned in the same concrete body

  3. Assessment of the feasibility of indefinite containment of canadian nuclear fuel wastes; Evaluation de la faisabilite du confinement illimite des dechets de combustible nucleaire canadiens

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-05-01

    This report presents an analysis of the expected corrosion behavior 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.

  4. Nuclear fission with inertial confinement

    CERN Document Server

    Koshkarev, D G

    2002-01-01

    The possibility of initiating the explosive fission reaction in a small quantity of fissile material through the heavy ions beam from the powerful accelerator-driver, developed for realization of the thermonuclear synthesis in the deuterium-tritium cylindrical targets with the direct ignition, is considered. The consequences of applying this method in the nuclear engineering are discussed

  5. HTR confinement/containment and the protection against aircraft crash

    International Nuclear Information System (INIS)

    Brinkmann, G.

    2005-01-01

    Does an HTR need a containment-pressure resistant-or is it possible- licensable-to have only a so called confinement. The answer depends on both the results of the safety analysis of the accidents considered in the design and the acceptance by the licensing authorities and the public of a safety approach only based on severe core damage exclusion. The safety approach to be developed for modular HTRs must describe the application of the defence in depth principle for such reactors. Whatever the requirements on the last confinement barrier could be, a convincing demonstration of the exclusion of any severe core damage is needed, relying on exhaustive and bounding considerations of severe core damage initiators and the use of non questionable arguments. The paper presents the containment issues for HTRs based on German experience background and considerations for modern modular HTR safety approach including beyond design situations. For the German HTRs (designed in the 80s), it could be shown in the licensing procedures in Germany that there was no need for a pressure retaining and gas tight containment to enclose radioactive nuclides released from the nuclear heat source. Instead, the confinement envelope acted in conjunction with other barriers to minimize the release of radioactive nuclides and the radiological impact on the environment. The confinement envelope consisted of the reactor building, a sub-atmospheric pressure system, a building pressure relief system, an HVAC systems isolation and a filtration system. During a major depressurization accident, unfiltered releases were discharged to the environment. The analyses results show that the environmental impact was far below the dose limits according to the German Radiological Protection Ordinance, even when the effect of filters was not taken into account. The demonstration strongly relied on the assumptions made for the source term definition e. g. the fuel particles failure rates (under irradiation and

  6. 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

  7. Nuclear reactor container

    International Nuclear Information System (INIS)

    Takahashi, Hiroyuki.

    1987-01-01

    Purpose: To improve the earthquake proofness and also increase the safety to a nuclear reactor container by preventing bucklings upon earthquake. Constitution: A device for absorbing the deformation exerted from nuclear reactor buildings is disposed to a suppression chamber constituting a reactor container. When a nclear power plant encounters earthquakes, the entire reactor buildings are shaken and deformations of buildings are transmitted by way of building shell walls to a container and the forcive deforming forces are absorbed in the deformation absorbing device. That is, bellows are formed at the base of the container, which are deformed by the deforming forces to absorb the forcive deforming amount to moderate the stresses resulted to the suppression chamber. Thus, the rigidity to the bending of the container can be reduced and allowable displacement to the bucklings can be increased to prevent the buckling, by which earthquake proofness is improved and the safety is increased. (Kamimura, M.)

  8. 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.)

  9. Inertial Confinement Fusion R and D and Nuclear Proliferation

    International Nuclear Information System (INIS)

    Goldston, Robert J.

    2011-01-01

    In a few months, or a few years, the National Ignition Facility (NIF) at the Lawrence Livermore National Laboratory may achieve fusion gain using 192 powerful lasers to generate x-rays that will compress and heat a small target containing isotopes of hydrogen. This event would mark a major milestone after decades of research on inertial confinement fusion (ICF). It might also mark the beginning of an accelerated global effort to harness fusion energy based on this science and technology. Unlike magnetic confinement fusion (ITER, 2011), in which hot fusion fuel is confined continuously by strong magnetic fields, inertial confinement fusion involves repetitive fusion explosions, taking advantage of some aspects of the science learned from the design and testing of hydrogen bombs. The NIF was built primarily because of the information it would provide on weapons physics, helping the United States to steward its stockpile of nuclear weapons without further underground testing. The U.S. National Academies' National Research Council is now hosting a study to assess the prospects for energy from inertial confinement fusion. While this study has a classified sub-panel on target physics, it has not been charged with examining the potential nuclear proliferation risks associated with ICF R and D. We argue here that this question urgently requires direct and transparent examination, so that means to mitigate risks can be assessed, and the potential residual risks can be balanced against the potential benefits, now being assessed by the NRC. This concern is not new (Holdren, 1978), but its urgency is now higher than ever before.

  10. 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.)

  11. Magnetic confinement in plasmas in nuclear devices

    International Nuclear Information System (INIS)

    Tull, C.G.

    1979-01-01

    The main emphasis of the magnetic fusion energy research program today lies in the development of two types of confinement schemes: magnetic mirrors and tokamaks. Experimental programs for both of these confinement schemes have shown steady progress toward achieving fusion power breakeven. The scaling of the current machines to a reactor operating regime and newly developed methods for plasma heating will very likely produce power breakeven within the next decade. Predictions are that the efficiency in a fusion power plant should exceed 32%

  12. 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.

  13. Twenty years of ''Nuclear Fusion''. Inertial confinement

    International Nuclear Information System (INIS)

    Yamanaka, C.

    1980-01-01

    Inertial confinement (ICF) fusion research is directed towards demonstrating the feasibility of very rapidly heating and compressing small pellets of suitable fuel until conditions exist where thermonuclear fusion can occur and useful amounts of power can be produced. Major problems which have to be solved are the following: 1) pellet design based on driver-plasma coupling; 2) the technology of energy drivers; 3) feasibility of ICF reactor systems

  14. 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.)

  15. 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

  16. 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

  17. Effects of nuclear elastic scattering in magnetic confinement plasmas

    International Nuclear Information System (INIS)

    Matsuura, Hideaki

    2015-01-01

    When the parameter regions of magnetically confined D-T and D- 3 He nuclear fusion plasma are assumed, the nuclear elastic scattering appearing in these plasmas is as follows. The following are quantitatively explained: (1) formation of knock-on tail on the ion velocity distribution function due to nuclear elastic scattering, (2) change in nuclear fusion reaction rate coefficient due to the change in velocity distribution function via nuclear elastic scattering, (3) effects of nuclear elastic scattering on the reaction product ions and neutron emission spectrum, and (4) effects of nuclear elastic scattering on the energy distribution ratio toward the ion group and electron group in the deceleration process of fast ions in the plasma. The formation of the knock-on tail is an energy addition process to the bulk ion group due to the nuclear elastic scattering of fast ions. When nuclear elastic scattering is neglected, the deceleration of reaction product protons is underestimated, and so the equilibrium distribution function of high-speed region is overestimated by up to about 40%. On the other hand, thermal component is underestimated by about 40%. Nuclear elastic scattering possibly affects the trajectory of fast ions in magnetic configuration. (A.O.)

  18. 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

  19. 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.)

  20. Modeling of delayed strains of concrete under biaxial loadings. Application to the reactor containment of nuclear power plants; Modelisation des deformations differees du beton sous sollicitations biaxiales. application aux enceintes de confinement de batiments reacteurs des centrales nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Benboudjema, F

    2002-12-15

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

  1. Risk of containment sump plugging in EDF nuclear power plants; Risque de colmatage des puisards de l'enceinte de confinement des centrales nucleaires d'EDF

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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.)

  2. Dynamical nuclear polarization and confinement effects in ZnO quantum dots

    NARCIS (Netherlands)

    Baranov, P.G.; Orlinskii, S.B.; Hofmann, D.M.; de Mello Donega, C.; Meijerink, A.; Schmidt, J.

    2010-01-01

    The spatial distribution of the electronic wave function of a shallow donor (SD) in a ZnO semiconductor quantum dots (QD's) has been determined in the regime of quantum confinement by using the nuclear spins as probes. Hyperfine (HF) interactions as monitored by electron nuclear double resonance

  3. Colour confinement effects in the presence of nuclear matter

    International Nuclear Information System (INIS)

    Rek, Z.J.

    1979-01-01

    A simpleminded picture of the low psub(T) inclusive production on nuclei. The novel point is the conjecture, already investigated in the low psub(T) hadron-nucleon inclusive reactions, attributing the rising of hadron multiplicity to the necessity of the confinement of colour, the important subhadronic degree of freedom in the quancum chromodynamics. The elementary process is then viewed as proceeding in two stages: (a) The colour excitation of both incoming hadron, h → hsub(c) and target nucleon, N → Nsub(c). (b) The so called colour compensating flow. This is the abbreviation of a very complicated phenomenon caused by the necessity of the confinement of colour entering dramatically when both hsub(c) and Nsub(c) are receding from each other. (author)

  4. Confinement inertial fusion. Power reactors of nuclear fusion by lasers

    International Nuclear Information System (INIS)

    Velarde, G.; Ahnert, C.; Aragones, J.M.; Leira, G; Martinez-Val, J.M.

    1980-01-01

    The energy crisis and the need of the nuclear fusion energy are analized. The nuclear processes in the laser interation with the ablator material are studied, as well as the thermohydrodinamic processes in the implossion, and the neutronics of the fusion. The fusion reactor components are described and the economic and social impact of its introduction in the future energetic strategies.(author)

  5. Possible in-lattice confinement fusion (LCF). Dynamic application of atomic and nuclear data

    International Nuclear Information System (INIS)

    Kawarasaki, Yuuki

    1995-01-01

    New scheme of a nuclear fusion reactor system is proposed, the basic concept of which comes from ingenious combination of hitherto developed techniques and verified facts; 1) so-called cold fusion (CF), 2) plasma of both magnetic confinement fusion (MCF) and inertial confinement fusion (ICF), and 3) accelerator-based D-T(D) neutron source. Details of the LCF reactor physics require dynamics of atomic data as well as nuclear data; interaction of ions with matters in solid and the problems of radiation damage. (author)

  6. Automated analysis of cell migration and nuclear envelope rupture in confined environments.

    Science.gov (United States)

    Elacqua, Joshua J; McGregor, Alexandra L; Lammerding, Jan

    2018-01-01

    Recent in vitro and in vivo studies have highlighted the importance of the cell nucleus in governing migration through confined environments. Microfluidic devices that mimic the narrow interstitial spaces of tissues have emerged as important tools to study cellular dynamics during confined migration, including the consequences of nuclear deformation and nuclear envelope rupture. However, while image acquisition can be automated on motorized microscopes, the analysis of the corresponding time-lapse sequences for nuclear transit through the pores and events such as nuclear envelope rupture currently requires manual analysis. In addition to being highly time-consuming, such manual analysis is susceptible to person-to-person variability. Studies that compare large numbers of cell types and conditions therefore require automated image analysis to achieve sufficiently high throughput. Here, we present an automated image analysis program to register microfluidic constrictions and perform image segmentation to detect individual cell nuclei. The MATLAB program tracks nuclear migration over time and records constriction-transit events, transit times, transit success rates, and nuclear envelope rupture. Such automation reduces the time required to analyze migration experiments from weeks to hours, and removes the variability that arises from different human analysts. Comparison with manual analysis confirmed that both constriction transit and nuclear envelope rupture were detected correctly and reliably, and the automated analysis results closely matched a manual analysis gold standard. Applying the program to specific biological examples, we demonstrate its ability to detect differences in nuclear transit time between cells with different levels of the nuclear envelope proteins lamin A/C, which govern nuclear deformability, and to detect an increase in nuclear envelope rupture duration in cells in which CHMP7, a protein involved in nuclear envelope repair, had been depleted

  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. Achieving competitive excellence in nuclear energy: The threat of proliferation; the challenge of inertial confinement fusion

    International Nuclear Information System (INIS)

    Nuckolls, J.H.

    1994-06-01

    Nuclear energy will have an expanding role in meeting the twenty-first-century challenges of population and economic growth, energy demand, and global warming. These great challenges are non-linearly coupled and incompletely understood. In the complex global system, achieving competitive excellence for nuclear energy is a multi-dimensional challenge. The growth of nuclear energy will be driven by its margin of economic advantage, as well as by threats to energy security and by growing evidence of global warming. At the same time, the deployment of nuclear energy will be inhibited by concerns about nuclear weapons proliferation, nuclear waste and nuclear reactor safety. These drivers and inhibitors are coupled: for example, in the foreseeable future, proliferation in the Middle East may undermine energy security and increase demand for nuclear energy. The Department of Energy's nuclear weapons laboratories are addressing many of these challenges, including nuclear weapons builddown and nonproliferation, nuclear waste storage and burnup, reactor safety and fuel enrichment, global warming, and the long-range development of fusion energy. Today I will focus on two major program areas at the Lawrence Livermore National Laboratory (LLNL): the proliferation of nuclear weapons and the development of inertial confinement fusion (ICF) energy

  9. Cooling equipments in nuclear reactor containers

    International Nuclear Information System (INIS)

    Miyagawa, Takehisa; Sonoda, Takayuki.

    1986-01-01

    Purpose: To maintain a sound circumstance to stainless steel materials in a reactor container by passing intensely cooling gases to the upper portion in the container where the temperature tends to be higher, while gases with a relatively low cooling degree before cooling are passed from the middle to the lower portion to unify the temperature. Constitution: In a nuclear power facility having ducts contained in a reactor container in which a first opening is disposed to the space inside the pedestal, a second opening is disposed to the space inside the shielding wall and other openings are disposed above the shielding walls respectively, an air intake port communicating from the outer circumference to the inside of the pedestal is disposed to the pedestal, a blower is disposed at the midway of the duct between the first and second openings and a cooling coil is disposed at the midway of the duct between the blower and other openings, whereby the circumstantial integrity to the stainless steels materials can be ensured without supplying the cold air in the container to the space inside the pedestal directly. (Takahashi, M.)

  10. 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.

  11. 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.)

  12. 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

  13. WWER-440/V-230 Confinement modernization to upgrade the critical safety function 'Containment integrity' in case of severe accident

    International Nuclear Information System (INIS)

    Sartmadjiev, A.

    1999-01-01

    In this lecture the WWER-440/V-230 confinement modernization to upgrade the critical safety function 'Containment integrity' in case of severe accident is presented. There are discussed: design limitations of the location system; consequence from these design limitations; a few confinement reconstruction concepts of this type of units worldwide; and purpose of the confinement reconstruction - to improve significantly the original design, ensuring (1) localization for all possible primary breaks and (2) limitation of the radiological consequences for the personnel, the population and the environment below the regulatory requirements

  14. MEASUREMENTS OF THE CONFINEMENT LEAKTIGHTNESS AT THE KOLA NUCLEAR POWER STATION (UNIT 2) IN RUSSIA

    International Nuclear Information System (INIS)

    GREENE, G.A.; GUPPY, J.G.

    1998-01-01

    This is the final report on the INSP project entitled, ''Kola Confinement Leaktightness'' conducted by BNL under the authorization of Project Work Plan WBS 1.2.2.1. This project was initiated in February 1993 to assist the Russians to reduce risks associated with the continued operation of older Soviet-designed nuclear power plants, specifically the Kola VVER-440/230 Units 1 and 2, through upgrades in the confinement performance to reduce the uncontrolled leakage rate. The major technical objective of this-project was to improve the leaktightness of the Kola NPP VVER confinement boundaries, through the application of a variety of sealants to penetrations, doors and hatches, seams and surfaces, to the extent that current technology permitted. A related objective was the transfer, through training of Russian staff, of the materials application procedures to the staff of the Kola NPP. This project was part of an overall approach to minimizing uncontrolled releases from the Kola NPP VVER440/230s in the event of a serious accident, and to thereby significantly mitigate the consequences of such an accident. The US provided materials, application technology, and applications equipment for application of sealant materials, surface coatings, potting materials and gaskets, to improve the confinement leaktightness of the Kola VVER-440/23Os. The US provided for training of Russian personnel in the applications technology

  15. MEASUREMENTS OF THE CONFINEMENT LEAKTIGHTNESS AT THE KOLA NUCLEAR POWER STATION (UNIT 2) IN RUSSIA

    Energy Technology Data Exchange (ETDEWEB)

    GREENE,G.A.; GUPPY,J.G.

    1998-08-01

    This is the final report on the INSP project entitled, ``Kola Confinement Leaktightness'' conducted by BNL under the authorization of Project Work Plan WBS 1.2.2.1. This project was initiated in February 1993 to assist the Russians to reduce risks associated with the continued operation of older Soviet-designed nuclear power plants, specifically the Kola VVER-440/230 Units 1 and 2, through upgrades in the confinement performance to reduce the uncontrolled leakage rate. The major technical objective of this-project was to improve the leaktightness of the Kola NPP VVER confinement boundaries, through the application of a variety of sealants to penetrations, doors and hatches, seams and surfaces, to the extent that current technology permitted. A related objective was the transfer, through training of Russian staff, of the materials application procedures to the staff of the Kola NPP. This project was part of an overall approach to minimizing uncontrolled releases from the Kola NPP VVER440/230s in the event of a serious accident, and to thereby significantly mitigate the consequences of such an accident. The US provided materials, application technology, and applications equipment for application of sealant materials, surface coatings, potting materials and gaskets, to improve the confinement leaktightness of the Kola VVER-440/23Os. The US provided for training of Russian personnel in the applications technology.

  16. 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.

  17. Study of deuterons induced nuclear reactions on light elements (N, Al and Si): Application to containment materials of radioactive wastes; Etude des reactions nucleaires induites par des deuterons sur des elements legers (N, Al, Si): application aux materiaux de confinement des dechets radioactifs

    Energy Technology Data Exchange (ETDEWEB)

    Pellegrino, St

    2004-03-01

    Nuclear reaction analysis is well adapted to the quantification of light element. Profiles of concentration in order to follow elements migration into materials can be undertaken. This technique is used to study the behavior of the future matrices for nuclear waste containment. This technique is isotopic, characterized by a good signal-to-background ratio and a very low detection limit. The probability of a nuclear reaction is linked to a parameter called 'cross section' we have to know in order to carry out quantitative analysis. We have determined excitation curves for nitrogen, aluminium and silicon. These experiments were done with deuterons from 0.5 to 2 MeV. Two methods for the cross section characterization are presented and are in agreement with each other. The second one reduces uncertainty. Data are incorporated in the simulation software SIMNRA. We have compared the results obtained on different samples when we use data in literature or data of the study. We have noticed a great fit improvement with the data of this study. The new cross sections of this work will be integrated in the general data base SIGMABASE. Applications on materials such as Si{sub 3}N{sub 4}, nano-metric powders, WCN and nuclear glass YLaMgSiAlON studied for radioactive waste containment are also presented. (author)

  18. 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

  19. Escape rates and biting histories of dogs confined to their owner's property through the use of various containment methods.

    Science.gov (United States)

    Starinsky, Nicole S; Lord, Linda K; Herron, Meghan E

    2017-02-01

    OBJECTIVE To determine escape rates for dogs confined to their owner's property by various containment methods and determine whether biting history was associated with containment method. DESIGN Cross-sectional survey. SAMPLE 974 owners of 1,053 dogs. PROCEDURES Individuals patronizing pet stores in Columbus, Ohio, were recruited to complete a survey on the method they used to confine their dogs to their property and their dogs' behavior history. RESULTS Dogs were confined to their owner's property by a physical fence (821/1,053 [78.0%]), electronic fence (150/1,053 [14.2%]), or tether system (82/1,053 [7.8%]). Dogs confined by an electronic fence were more likely to have escaped (66/150 [44.0%]) than were dogs confined by a see-through fence (153/658 [23.3%]), privacy fence (38/163 [23.3%]), or tether (22/82 [26.8%]). Forty-eight (4.6%) dogs had reportedly bitten a person in the past, and 81 (7.7%) had reportedly bitten another dog, but containment method was not significantly associated with whether dogs had ever bitten a person or another dog. Greeting behavior (growling, snarling, or trying to bite) was significantly associated with a history of biting a person or another dog. CONCLUSIONS AND CLINICAL RELEVANCE Results suggested that escape rate, but not biting history, was associated with the method owners used to confine dogs to their properties. Greeting behavior was associated with biting history, suggesting that owners of dogs that growl, snarl, or attempt to bite when meeting an unfamiliar person or dog should seek assistance to prevent future bites.

  20. 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)

  1. 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

  2. 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

  3. Calculation code evaluating the confinement of a nuclear facility in case of fires

    Energy Technology Data Exchange (ETDEWEB)

    Laborde, J.C.; Prevost, C.; Vendel, J. [and others

    1995-02-01

    Accident events involving fire are quite frequent and could have a severe effect on the safety of nuclear facilities. As confinement must be maintained, the ventilation and filtration systems have to be designed to limit radioactive release to the environment. To determine and analyse the consequences of a fire on the contamination confinement, IPSN, COGEMA and SGN are participating in development of a calculation code based on introduction, in the SIMEVENT ventilation code, of various models associated to fire risk and mass transfer in the ventilation networks. This calculation code results from the coupling of the SIMEVENT code with several models describing the temperature in a room resulting of a fire, the temperatures along the ventilation ducts, the contamination transfers through out the ventilation equipments (ducts, dampers, valves, air cleaning systems) and the High Efficiency Particulate Air (HEPA) filters clogging. The paper proposed presents the current level of progress in development of this calculation code. It describes, in particular, the empirical model used for the clogging of HEPA filters by the aerosols derived from the combustion of standard materials used in the nuclear industry. It describes, also, the specific models used to take into account the mass transfers and resulting from the basic mechanisms of aerosols physics. In addition, an assessment of this code is given using the example of a simple laboratory installation.

  4. 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.

  5. 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

  6. Nondestructive Inspection System for Special Nuclear Material Using Inertial Electrostatic Confinement Fusion Neutrons and Laser Compton Scattering Gamma-Rays

    Science.gov (United States)

    Ohgaki, H.; Daito, I.; Zen, H.; Kii, T.; Masuda, K.; Misawa, T.; Hajima, R.; Hayakawa, T.; Shizuma, T.; Kando, M.; Fujimoto, S.

    2017-07-01

    A Neutron/Gamma-ray combined inspection system for hidden special nuclear materials (SNMs) in cargo containers has been developed under a program of Japan Science and Technology Agency in Japan. This inspection system consists of an active neutron-detection system for fast screening and a laser Compton backscattering gamma-ray source in coupling with nuclear resonance fluorescence (NRF) method for precise inspection. The inertial electrostatic confinement fusion device has been adopted as a neutron source and two neutron-detection methods, delayed neutron noise analysis method and high-energy neutron-detection method, have been developed to realize the fast screening system. The prototype system has been constructed and tested in the Reactor Research Institute, Kyoto University. For the generation of the laser Compton backscattering gamma-ray beam, a race track microtron accelerator has been used to reduce the size of the system. For the NRF measurement, an array of LaBr3(Ce) scintillation detectors has been adopted to realize a low-cost detection system. The prototype of the gamma-ray system has been demonstrated in the Kansai Photon Science Institute, National Institutes for Quantum and Radiological Science and Technology. By using numerical simulations based on the data taken from these prototype systems and the inspection-flow, the system designed by this program can detect 1 kg of highly enriched 235U (HEU) hidden in an empty 20-ft container within several minutes.

  7. Study of the confined solutions properties: case of gel formed during nuclear glass alteration

    International Nuclear Information System (INIS)

    Matar-Briman, I.

    2012-01-01

    In this study, we have investigated the thermodynamic properties, the structure and the dynamics of confined solutions in model gels and in a gel coming from glass alteration. The first step was to determine the structure and the dynamics of pure confined water in porous materials by using nuclear magnetic resonance and neutron scattering. Meso-porous silica was elaborated and grafted by sol-gel route to decrease the pore sizes from 2.7 to 2 nm and to modify pore surfaces to have Si-OH, Zr-OH and Al-OH. The second step involved determining the dynamics of water in leachate confined in the model gels and in the gel of altered glass by using neutron scattering. In the model gels and at a 10 -12 -10 -9 second timescale, two kinds of waters were highlighted: first, an interfacial water linked to the pore surfaces and second, a free water in the pore core. Their ratio depends on the pore size and pore surface composition. Whatever the pore surface, when the pore size decreases the free water ratio in the pore center also decreases. For pores smaller than 2.3 nm and pore surfaces with Zr-OH or Al-OH surfaces, water is strongly linked to the surface and few water molecules are mobile. This is due to the ability of alumina and zirconia to immobilize water molecules through chemical coordination bonds stronger than the physical bonds established between silica and water. The result also highlight that pore surface composition could be the predominant parameter affecting the fixed proton content. Moreover, the mobility of water confined in a leachate is not modified. The study of the water dynamics in a gel formed during alteration of glass constituted of SiO 2 , Al 2 O 3 and CaO, and having a porosity between 2 and 7 nm showed the same behavior as water confined in pores presenting an Al-OH surface. (author) [fr

  8. Advanced energy systems: 2XIIB: heating and containing magnetically confined plasmas

    International Nuclear Information System (INIS)

    Coensgen, F.H.

    1975-01-01

    Recent experiments on the 2XIIB mirror machine have produced encouraging results: a buildup of hot ion densities to 4 x 10 13 cm -3 , ion temperatures of 13 keV (the highest ever observed in a major fusion experiment), and a confinement time exceeding 5 ms. Two major factors in these achievements were the injection of twelve 20-keV neutral beams to increase plasma temperature and the introduction of warm streaming plasma to suppress microinstabilities. With them, near-classical confinement of a hot plasma was demonstrated. We are now doubling the injected neutral beam energy to see if plasma stability and energy scaling of plasma confinement persist at higher ion temperatures

  9. 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

  10. Development of an inertial confinement fusion platform to study charged-particle-producing nuclear reactions relevant to nuclear astrophysics

    Science.gov (United States)

    Gatu Johnson, M.; Zylstra, A. B.; Bacher, A.; Brune, C. R.; Casey, D. T.; Forrest, C.; Herrmann, H. W.; Hohenberger, M.; Sayre, D. B.; Bionta, R. M.; Bourgade, J.-L.; Caggiano, J. A.; Cerjan, C.; Craxton, R. S.; Dearborn, D.; Farrell, M.; Frenje, J. A.; Garcia, E. M.; Glebov, V. Yu.; Hale, G.; Hartouni, E. P.; Hatarik, R.; Hohensee, M.; Holunga, D. M.; Hoppe, M.; Janezic, R.; Khan, S. F.; Kilkenny, J. D.; Kim, Y. H.; Knauer, J. P.; Kohut, T. R.; Lahmann, B.; Landoas, O.; Li, C. K.; Marshall, F. J.; Masse, L.; McEvoy, A.; McKenty, P.; McNabb, D. P.; Nikroo, A.; Parham, T. G.; Paris, M.; Petrasso, R. D.; Pino, J.; Radha, P. B.; Remington, B.; Rinderknecht, H. G.; Robey, H.; Rosenberg, M. J.; Rosse, B.; Rubery, M.; Sangster, T. C.; Sanchez, J.; Schmitt, M.; Schoff, M.; Séguin, F. H.; Seka, W.; Sio, H.; Stoeckl, C.; Tipton, R. E.

    2017-04-01

    This paper describes the development of a platform to study astrophysically relevant nuclear reactions using inertial-confinement fusion implosions on the OMEGA and National Ignition Facility laser facilities, with a particular focus on optimizing the implosions to study charged-particle-producing reactions. Primary requirements on the platform are high yield, for high statistics in the fusion product measurements, combined with low areal density, to allow the charged fusion products to escape. This is optimally achieved with direct-drive exploding pusher implosions using thin-glass-shell capsules. Mitigation strategies to eliminate a possible target sheath potential which would accelerate the emitted ions are discussed. The potential impact of kinetic effects on the implosions is also considered. The platform is initially employed to study the complementary T(t,2n)α, T(3He,np)α and 3He(3He,2p)α reactions. Proof-of-principle results from the first experiments demonstrating the ability to accurately measure the energy and yields of charged particles are presented. Lessons learned from these experiments will be used in studies of other reactions. The goals are to explore thermonuclear reaction rates and fundamental nuclear physics in stellar-like plasma environments, and to push this new frontier of nuclear astrophysics into unique regimes not reachable through existing platforms, with thermal ion velocity distributions, plasma screening, and low reactant energies.

  11. Seal welded cast iron nuclear waste container

    Science.gov (United States)

    Filippi, Arthur M.; Sprecace, Richard P.

    1987-01-01

    This invention identifies methods and articles designed to circumvent metallurgical problems associated with hermetically closing an all cast iron nuclear waste package by welding. It involves welding nickel-carbon alloy inserts which are bonded to the mating plug and main body components of the package. The welding inserts might be bonded in place during casting of the package components. When the waste package closure weld is made, the most severe thermal effects of the process are restricted to the nickel-carbon insert material which is far better able to accommodate them than is cast iron. Use of nickel-carbon weld inserts should eliminate any need for pre-weld and post-weld heat treatments which are a problem to apply to nuclear waste packages. Although the waste package closure weld approach described results in a dissimilar metal combination, the relative surface area of nickel-to-iron, their electrochemical relationship, and the presence of graphite in both materials will act to prevent any galvanic corrosion problem.

  12. 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

  13. 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

  14. 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.

  15. Design of the reactor containment systems in nuclear power plants

    International Nuclear Information System (INIS)

    1985-01-01

    The present Guide deals with the following design aspects of containment systems: The considerations which make containment necessary; The purpose of, and functional requirements for, the containment system; The design basis for the containment system; The specific design requirements and features needed to implement the functional requirements of the containment system; The tests and inspections needed to ensure that the functional requirements of the system can be met throughout the operating life of a nuclear power plant; The general quality assurance and documentation requirements for design of the containment system. This Safety Guide covers the method of limiting the radionuclide releases resulting from accident conditions

  16. 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

  17. Multiscale nuclear magnetic relaxation dispersion of complex liquids in bulk and confinement.

    Science.gov (United States)

    Korb, Jean-Pierre

    2018-02-01

    The nuclear magnetic relaxation dispersion (NMRD) technique consists of measurement of the magnetic-field dependence of the longitudinal nuclear-spin-lattice relaxation rate 1/T 1 . Usually, the acquisition of the NMRD profiles is made using a fast field cycling (FFC) NMR technique that varies the magnetic field and explores a very large range of Larmor frequencies (10 kHz relaxation is sensitive. The FFC technique thus offers opportunities on multiple scales of both time and distance for characterizing the molecular dynamics and transport properties of complex liquids in bulk or embedded in confined environments. This review presents the principles, theories and applications of NMRD for characterizing fundamental properties such as surface correlation times, diffusion coefficients and dynamical surface affinity (NMR wettability) for various confined liquids. The basic longitudinal and transverse relaxation equations are outlined for bulk liquids. The nuclear relaxation of a liquid confined in pores is considered in detail in order to find the biphasic fast exchange relations for a liquid at proximity of a solid surface. The physical-chemistry of liquids at solid surfaces induces striking differences between NMRD profiles of aprotic and protic (water) liquids embedded in calibrated porous disordered materials. A particular emphasis of this review concerns the extension of FFC NMR relaxation to industrial applications. For instance, it is shown that the FFC technique is sufficiently rapid for following the progressive setting of cement-based materials (plasters, cement pastes, concretes). The technique also allows studies of the dynamics of hydrocarbons in proximity of asphaltene nano-aggregates and macro-aggregates in heavy crude oils as a function of the concentration of asphaltenes. It also gives new information on the wettability of petroleum fluids (brine and oil) embedded in shale oil rocks. It is useful for understanding the relations and correlations

  18. 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)

  19. Pressure release in the containments of nuclear power stations

    International Nuclear Information System (INIS)

    Pauli, W.; Pellaud, B.

    1990-01-01

    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 attemps to reconcile these two conflicting objectives by allowing a filtered pressure relief after the 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. (orig.) [de

  20. Nuclear Quantum Effects in H(+) and OH(-) Diffusion along Confined Water Wires.

    Science.gov (United States)

    Rossi, Mariana; Ceriotti, Michele; Manolopoulos, David E

    2016-08-04

    The diffusion of protons and hydroxide ions along water wires provides an efficient mechanism for charge transport that is exploited by biological membrane channels and shows promise for technological applications such as fuel cells. However, what is lacking for a better control and design of these systems is a thorough theoretical understanding of the diffusion process at the atomic scale. Here we focus on two aspects of this process that are often disregarded because of their high computational cost: the use of first-principles potential energy surfaces and the treatment of the nuclei as quantum particles. We consider proton and hydroxide ions in finite water wires using density functional theory augmented with an apolar cylindrical confining potential. We employ machine learning techniques to identify the charged species, thus obtaining an agnostic definition that takes explicitly into account the delocalization of the charge in the Grotthus-like mechanism. We include nuclear quantum effects (NQEs) through the thermostated ring polymer molecular dynamics method and model finite system size effects by considering Langevin dynamics on the potential of mean force of the charged species, allowing us to extract the same "universal" diffusion coefficient from simulations with different wire sizes. In the classical case, diffusion coefficients depend significantly on the potential energy surface, in particular on how dispersion forces modulate water-water distances. NQEs, however, make the diffusion less sensitive to the underlying potential and geometry of the wire.

  1. 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

  2. Unexpected hazard of illegal immigration: Outbreak of viral myocarditis exacerbated by confinement and deprivation in a shipboard cargo container.

    Science.gov (United States)

    Li, Melissa K; Beck, Melinda A; Shi, Qing; Harruff, Richard C

    2004-06-01

    We present a group of 18 illegal immigrant stowaways who arrived in a shipboard cargo container suffering from gastroenteritis, dehydration, and malnutrition and showing evidence of viral myocarditis in 3 of 4 fatalities. Our investigation included an evaluation of the 2-week ocean voyage, analysis of medical records and laboratory results of the survivors, autopsies on the decedents, and viral studies on their heart tissue. Of 3 stowaways who died shipboard, 2 showed lymphocytic myocarditis and 1 could not be evaluated histologically due to decomposition. A fourth stowaway died 4 months after arrival with dilated cardiomyopathy and lymphocytic myocarditis. Reverse-transcriptase polymerase chain reaction and nucleotide sequencing of viral isolates from the decedents' heart tissues demonstrated Coxsackie virus B3 genome. We believe that these cases represent an outbreak of viral myocarditis, exacerbated by acute dehydration and malnutrition, due to confinement within the shipping container. Our evidence indicates that close confinement promoted the spread of the virus, and nutritional deprivation increased the stowaways' vulnerability. Furthermore, our observations support the conclusion, based on experimental studies, that nutritionally induced oxidative stress increased the virulence of the etiologic viral agent. In summary, these cases represent a potential infectious disease hazard of illegal immigration.

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

    Science.gov (United States)

    Powell, J.G.

    1993-11-16

    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.

  4. NF ISO 15080. Nuclear facilities. Ventilation penetrations for shielded containment

    International Nuclear Information System (INIS)

    2002-03-01

    This standard document applies to any kind of shielded containment devoted to the handling of gamma-emitting or neutron-emitting radioactive materials that require a remote handling behind a protection screen. It covers all kinds of hot cells used in the nuclear fuel cycle (reprocessing facilities, hot labs, shielded containments, radioactive waste storage facilities etc..). It can eventually apply also to the containments used in accelerators, research reactors and fusion reactors, radiography facilities and neutrons generators. It precises the general and detailed principles that must be respected during the design of ventilation penetrations inside shielded containments (conventional penetration systems and 'helicoidal cast iron screw technique' systems). (J.S.)

  5. 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.

  6. 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.)

  7. 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)

  8. 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

  9. 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.)

  10. Containment vs confinement trade study, small HTGR plant PCRV [prestressed concrete reactor vessel] concept

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-03-01

    This trade study has been conducted to evaluate the differences between four different HTGR nuclear power plants. All of the plants use a prestressed concrete reactor vessel (PCRV) to house the core and steam generation equipment. The reactor uses LEU U/Th fuel in prismatic carbon blocks. All plant concepts meet the utility/user requirements established for small HTGR plants. All plants will be evaluated with regard to their ability to produce safe, economical power to satisfy Goals 1, 2, and 3 of the HTGR program and by meeting the MUST criteria established in the concept evaluation plan. Capital costs for each plant were evaluated on a differential cost basis. These costs were developed according to the ``NUS`` code of accounts as defined in the Cost Estimating and Control Procedure, HP-20901. Accounts that were identical in scope for all four plants were not used for the comparison. Table 1-1 is a list of capital cost accounts that were evaluated for each plant.

  11. 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.

  12. 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

  13. 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

  14. 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

  15. Los Alamos National Laboratory standard nuclear material container

    Energy Technology Data Exchange (ETDEWEB)

    Stone, Timothy A [Los Alamos National Laboratory

    2009-01-01

    The shut down of United States (U.S.) nuclear-weapons production activities in the early 1990s left large quantities of nuclear materials throughout the U.S. Department of Energy (DOE) complex in forms not intended for long-term storage. In May 1994, the Defense Nuclear Facilities Safety Board (DNFSB) issued Recommendation 94-1, which called for the stabilization and disposition of 'thousands of containers of plutonium-bearing liquids and solids' in the DOE complex, including LANL in the nuclear-weapons-manufacturing pipeline when manufacturing ended. This resulted in the development of the 3013 standard with container requirements for long term storage (up to 50 years). A follow on was the Criteria For Interim Storage of Plutonium Bearing Materials, Charles B. Curtis, in 1996 to address storage other than the 3013 standard for shorter time frames. In January 2000, the DNFSB issued Recommendation 2000-1, which stated the need for LANL to repackage 'about one ton of plutonium metal and oxide,' declared excess to Defense Program (DP) needs. The DNFSB recommended that LANL 'stabilize and seal within welded containers with an inert atmosphere the plutonium oxides ... which are not yet in states conforming to the long-term storage envisaged by DOE-STD-3013,' and that they '... enclose existing and newly-generated legacy plutonium metal in sealed containers with an inert atmosphere,' and 'remediate and/or safely store the various residues.' Recommendation 2000-1, while adding to the number of items needing remediation, also reiterated the need to address remaining items from 1994-1 in a timely fashion. Since timetables slipped, the DNFSB recommended that the Complex 'prioritize and schedule tasks according to the consideration of risks.' In March 2005, the DNFSB issued Recommendation 2005-1. This recommendation addresses the need for a consistent set of criteria across the DOE complex for the interim storage of

  16. 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.

  17. 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

  18. 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.

  19. 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.

  20. 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

  1. Study of the ruthenium fission-product behavior in the containment, in the case of a nuclear reactor severe accident; Etude du comportement du produit de fission ruthenium dans l'enceinte de confinement d'un reacteur nucleaire, en cas d'accident grave

    Energy Technology Data Exchange (ETDEWEB)

    Mun, Ch

    2007-03-15

    Ruthenium tetroxide is an extremely volatile and highly radio-toxic species. During a severe accident with air ingress in the reactor vessel, ruthenium oxides may reach the reactor containment building in significant quantities. Therefore, a better understanding of the RuO{sub 4}(g) behaviour in the containment atmosphere is of primary importance for the assessment of radiological consequences, in the case of potential releases of this species into the environment. A RuO{sub 4}(g) decomposition kinetic law was determined. Steam seems to play a catalytic role, as well as the presence of ruthenium dioxide deposits. The temperature is also a key parameter. The nature of the substrate, stainless steel or paint, did not exhibit any chemical affinities with RuO{sub 4}(g). This absence of reactivity was confirmed by XPS analyses, which indicate the presence of the same species in the Ru deposits surface layer whatever the substrates considered. It has been concluded that RuO{sub 4}(g) decomposition corresponds to a bulk gas phase decomposition. The ruthenium re-volatilization phenomenon under irradiation from Ru deposits was also highlighted. An oxidation kinetic law was determined. The increase of the temperature and the steam concentration promote significantly the oxidation reaction. The establishment of Ru behavioural laws allowed making a modelling of the Ru source term. The results of the reactor calculations indicate that the values obtained for {sup 106}Ru source term are closed to the reference value considered currently by the IRSN, for 900 MWe PWR safety analysis. (author)

  2. 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.

  3. 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.

  4. 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.

  5. Zero-gravity Mean Free Surface Curvature of a Confined Liquid in a Radially-Vaned Container

    Science.gov (United States)

    Chen, Yongkang; Callahan, Michael; Weislogel, Mark

    2013-01-01

    A variety of increasingly intricate container geometries are under consideration for the passive manipulation of liquids aboard spacecraft where the impact of gravity may be neglected. In this study we examine the mean curvature of a liquid volume confined within a radial array of disconnected vanes of infinite extent. This particular geometry possesses a number of desirable characteristics relevant to waste water treatment aboard spacecraft for life support. It is observed that under certain conditions the slender shape of the free surface approaches an asymptote, which can be predicted analytically using new hybrid boundary conditions proposed herein. This contribution represents possibly the final extension of what has been referred to as the method of de Lazzer et al. (1996). The method enables the integration of the Young-Laplace equation over a domain with its boundaries, including the wetted portion of the solid boundaries, symmetry planes, and circular arcs representing free surfaces at the center plane of the liquid body. Asymptotic solutions at several limits are obtained and the analysis is confirmed with numerical computations.

  6. 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

  7. Development of waste packages for the long-term confinement of C-14 in TRU waste disposal. 2. Confinement container with titanium alloy

    International Nuclear Information System (INIS)

    Nakamura, Ario; Owada, Hitoshi; Asano, Hidekazu; Jintoku, Takashi; Nakayama, Gen

    2008-01-01

    The long-term integrity of TRU waste package, with a titanium alloy for the outer corrosion resistance layer and carbon steel for the inner structural vessel, has been evaluated. The target confinement period is settled at 60,000 years in this study (i.e., 10 times of half-life). So the outer corrosion resistance layer with titanium (Ti-Pd alloy) is developed through focus on the high corrosion resistance of Ti alloy as a technology that has long-term confinement. In investigation about integrity of its passive film, the thickness of corrosion layer of 60,000 years is calculated by building an empirical formula between temperature and corrosion current density, considering the results of constant voltage examination in the TRU waste repository assumed environment. About crevice corrosion, its occurrence conditions is investigated in the TRU waste repository assumed environment, then, Ti.Gr-17 is selected as candidate material of the corrosion resistance layer. In investigation about SCC in Ti alloy, using the models of growth of hydride-layer, the thickness of hydride-layer after 60,000 years is estimated by the results of constant currents tests. Then, the hydride-layer of this thickness is confirmed not to generate cracks, in consideration of destructive critical hydride cracking thickness and the models of crack propagation. The knowledge that the process of generation of hydrogenated layers changes with differences in acceleration conditions (i.e., current density) is obtained. So we must confirm the adequacy of this model by the examination with natural condition. (author)

  8. Retention capacity of samarium (III) in zircon for it possible use in retaining walls for confinement of nuclear residues

    International Nuclear Information System (INIS)

    Garcia G, N.

    2006-01-01

    Mexico, as country that produces part of its electric power by nuclear means, should put special emphasis in the development of technologies guided to the sure and long term confinement of the high level nuclear residuals. This work studies the capacity that has the natural zircon to retain to the samarium (III) in solution, by what due, firstly, to characterize the zircon for technical instrumental to determine the purity and characteristic of the mineral in study. The instrumental techniques that were used to carry out the physicochemical characterization were the neutron activation analysis (NAA), the infrared spectroscopy (IS), the thermal gravimetric analysis (TGA), scanning electron microscopy (SEM), transmission electron microscopy (TEM), semiquantitative analysis, dispersive energy spectroscopy (EDS), X-ray diffraction (XRD) and luminescence technique. The characterization of the surface properties carries out by means of the determination of the surface area using the BET multipoint technique, acidity constants, hydration time, the determination of the point of null charge (pH PCN ) and density of surface sites (D s ). The luminescence techniques were useful to determine the optimal point hydration of the zircon and for the quantification of the samarium, for that here intends the development of both analysis techniques. With the adjustment of the titration curves in the FITEQL 4 package the constants of surface acidity in the solid/liquid interface were determined. To the finish of this study it was corroborated that the zircon is a mineral that presents appropriate characteristics to be proposed as a contention barrier for the deep geologic confinement. With regard to the study of adsorption that one carries out the samarium retention it is superior to 90% under the described conditions. This investigation could also be applicable in the confinement of dangerous industrial residuals. (Author)

  9. Corrosion of carbon steel nuclear waste containers in marine sediment

    International Nuclear Information System (INIS)

    Marsh, G.P.; Taylor, K.J.; Harker, A.H.

    1987-08-01

    The report describes a study of the corrosion of carbon steel nuclear waste containers in deep ocean sediments, which had the objective of estimating the metal allowance needed to ensure that the containers were not breached by corrosion for 1000 years. It was concluded that under such disposal conditions carbon steel would not be subject to localised corrosion or hydrogen embrittlement, and therefore the study concentrated on evaluating the rate of general attack. This was carried out by developing a mechanistically based mathematical model which was formulated on the conservative assumption that the corrosion would be under activation control, and would not be impeded by the formation of corrosion product layers. This model predicted that an allowance of 33 mm would be required for a 1000 year life. (author)

  10. 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

  11. 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

  12. 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.

  13. 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)

  14. Hydrogen risk analysis for a generic nuclear containment ventilation system

    International Nuclear Information System (INIS)

    Xu, Z.; Jordan, T.

    2015-01-01

    Hydrogen safety issue in a ventilation system of a generic nuclear containment is studied. In accidental scenarios, a large amount of burnable gas mixture of hydrogen with certain amount of oxygen is released into the containment. In case of high containment pressure, the combustible mixture is further ventilated into the chambers and the piping of the containment ventilation system. The burnable even potentially detonable gas mixture could pose a risk to the structures of the system once being ignited unexpectedly. Therefore the main goal of the study is to apply the computational fluid dynamics (CFD) computer code - GASFLOW, to analyze the distribution of the hydrogen in the ventilation system, and to find how sensitive the mixture is to detonation in different scenarios. The CFD simulations manifest that a ventilation fan with sustained power supply can extinguish the hydrogen risk effectively. However in case of station blackout with loss of power supply to the fan, hydrogen/oxygen mixture could be accumulated in the ventilation system. A further study proves that steam injection could degrade the sensitivity of the hydrogen mixture significantly. (author)

  15. Neutron Transport and Nuclear Burnup Analysis for the Laser Inertial Confinement Fusion-Fission Energy (LIFE) Engine

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, K J; Latkowski, J F; Abbott, R P; Boyd, J K; Powers, J J; Seifried, J E

    2008-10-24

    Lawrence Livermore National Laboratory is currently developing a hybrid fusion-fission nuclear energy system, called LIFE, to generate power and burn nuclear waste. We utilize inertial confinement fusion to drive a subcritical fission blanket surrounding the fusion chamber. It is composed of TRISO-based fuel cooled by the molten salt flibe. Low-yield (37.5 MJ) targets and a repetition rate of 13.3 Hz produce a 500 MW fusion source that is coupled to the subcritical blanket, which provides an additional gain of 4-8, depending on the fuel. In the present work, we describe the neutron transport and nuclear burnup analysis. We utilize standard analysis tools including, the Monte Carlo N-Particle (MCNP) transport code, ORIGEN2 and Monteburns to perform the nuclear design. These analyses focus primarily on a fuel composed of depleted uranium not requiring chemical reprocessing or enrichment. However, other fuels such as weapons grade plutonium and highly-enriched uranium are also under consideration. In addition, we have developed a methodology using {sup 6}Li as a burnable poison to replace the tritium burned in the fusion targets and to maintain constant power over the lifetime of the engine. The results from depleted uranium analyses suggest up to 99% burnup of actinides is attainable while maintaining full power at 2GW for more than five decades.

  16. 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

  17. 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

  18. Unconventional Nuclear Warfare Defense (UNWD) containment and mitigation subtask.

    Energy Technology Data Exchange (ETDEWEB)

    Wente, William Baker

    2005-06-01

    The objective of this subtask of the Unconventional Nuclear Warfare Design project was to demonstrate mitigation technologies for radiological material dispersal and to assist planners with incorporation of the technologies into a concept of operations. The High Consequence Assessment and Technology department at Sandia National Laboratories (SNL) has studied aqueous foam's ability to mitigate the effects of an explosively disseminated radiological dispersal device (RDD). These benefits include particle capture of respirable radiological particles, attenuation of blast overpressure, and reduction of plume buoyancy. To better convey the aqueous foam attributes, SNL conducted a study using the Explosive Release Atmospheric Dispersion model, comparing the effects of a mitigated and unmitigated explosive RDD release. Results from this study compared health effects and land contamination between the two scenarios in terms of distances of effect, population exposure, and remediation costs. Incorporating aqueous foam technology, SNL created a conceptual design for a stationary containment area to be located at a facility entrance with equipment that could minimize the effects from the detonation of a vehicle transported RDD. The containment design was evaluated against several criteria, including mitigation ability (both respirable and large fragment particle capture as well as blast overpressure suppression), speed of implementation, cost, simplicity, and required space. A mock-up of the conceptual idea was constructed at SNL's 9920 explosive test site to demonstrate the containment design.

  19. 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

  20. 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)

  1. Deliberated opinion of the Environment Authority on the ECRIN basic nuclear installation 'confined warehousing of residues produced by conversion' (11)

    International Nuclear Information System (INIS)

    2013-01-01

    This report proposes an environmental review of the creation authorization request for a basic nuclear installation (INB), the ECRIN INB in Malvesi, for the confined warehousing of residues produced by uranium conversion (transformation into uranium tetrafluoride). After having formulated the main recommendations regarding the style and content of this request, this report presents and comments the context, the project and its environmental stakes, indicates the concerned administrative procedures, discusses the project content and the projected works. It proposes an analysis and a discussion of the impact study and makes recommendations to improve it. This analysis addresses the site environmental initial condition, possible alternative solutions, the dismantling plan, various temporary or permanent impacts of the project, the risk management study, and follow-up measures

  2. Fusion power by magnetic confinement: plans and the associated need for nuclear engineers

    International Nuclear Information System (INIS)

    Hirsch, R.L.; Beard, D.S.

    1975-01-01

    An essential ingredient in the fusion development plan will be the training of appropriate scientific and technical manpower. In examining the need for fusion-trained nuclear engineers, it is projected that an additional 120 to 250 engineers at the MS and PhD levels will be needed between now and 1980. To be most effective, these graduates must not only be trained in the ''classic'' physical, nuclear, mechanical, and electrical sciences, but they will need specialized training in fusion plasma physics and fusion materials science. To help develop the appropriate educational programs, close cooperation between U. S. Energy Research and Development Administration (ERDA) headquarters, ERDA laboratories, private industry, and the universities will be essential. An emerging need for a carefully structured ''fusion technology'' option in nuclear engineering departments is plainly evident and is already beginning to be developed at leading institutions

  3. 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

  4. 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.

  5. 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...

  6. 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.

  7. 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

  8. 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.

  9. Characterization of nuclear reactor containment penetrations. Final report

    International Nuclear Information System (INIS)

    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

  10. 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

  11. 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

  12. Basaltic glass alteration in confined media: analogy with nuclear glass in geological disposal conditions

    International Nuclear Information System (INIS)

    Parruzot, Benjamin

    2014-01-01

    This dissertation concerns basaltic glass alteration mechanisms and rates. Through a better understanding of the processes controlling the basaltic glass durability, this thesis attempts to establish a link between laboratory studies and volcanic glass alteration in natural environment. The methodology used here is similar to the one used for nuclear glasses. Thus, we measured for the first time the residual alteration rate of basaltic glasses. Protective effect of the alteration film is clearly established. Moreover, synthetic glass representativeness is evaluated through a study focused on the effect of iron oxidation degree on the glass structure and leaching properties. A minor effect of Fe II on the forward rate and a negligible effect on the residual rate are shown. The residual rate is extrapolated at 5 C and compared to the mean alteration rate of natural samples of ages ranging from 1900 to 10 7 years. Non-zeolitized natural glasses follow this linear tendency, suggesting a control of the long-term rate by clayey secondary phase precipitation. Natural environments are open environments: a parametric study was performed in order to quantify the water flow rate effect on chemical composition of the alteration layer. When applied to two natural samples, the obtained laws provide coherent results. It seems possible to unify the descriptive approach from the study of natural environments to the mechanistic approach developed at the laboratory. The next step will consist in developing a model to transpose these results to nuclear glasses. (author) [fr

  13. 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

  14. Structural and crystallisation study of a rare earth alumino borosilicate glass designed for nuclear waste confinement

    International Nuclear Information System (INIS)

    Quintas, A.

    2007-09-01

    This work is devoted to the study of a rare earth alumino borosilicate glass, which molar composition is 61,81 SiO 2 - 3,05 Al 2 O 3 - 8,94 B 2 O 3 - 14,41 Na 2 O - 6,33 CaO - 1,90 ZrO 2 - 3,56 Nd 2 O 3 , and envisaged for the immobilization of nuclear wastes originating from the reprocessing of high discharge burn up spent fuel. From a structural viewpoint, we investigated the role of the modifier cations on the arrangement of the glass network through different modifications of the glass composition: variation of the Na/Ca ratio and modification of the nature of the alkali and alkaline earth cations. The NMR and Raman spectroscopic techniques were useful to determine the distribution of modifier cations among the glass network and also to cast light on the competition phenomena occurring between alkali and alkaline earth cations for charge compensation of [AlO 4 ] - and [BO 4 ] - species. The neodymium local environment could be probed by optical absorption and EXAFS spectroscopies which enabled to better understand the insertion mode of Nd 3+ ions among the silicate domains of the glass network. Concerning the crystallization behavior we were interested in how the glass composition may influence the crystallization processes and especially the formation of the apatite phase of composition Ca 2 Nd 8 (SiO 4 ) 6 O 2 . In particular, this work underlined the important role of both alkaline earth and rare earth cations on the crystallization of the apatite phase. (author)

  15. Syntheses and characterizations of rare earth doped phospho-silicated apatites: application to nuclear waste confinement

    International Nuclear Information System (INIS)

    Boyer, Laurent

    1998-01-01

    Apatite matrices have been developed for the conditioning of actinides from spent fuels of PWR reactors. Silicated apatites (britholites) containing actinides and lanthanides have been discovered in the natural environment. Synthetic analogues of these britholites can be obtained by solid-solid reaction at high temperature. The compounds of the solid solution of fluorinated britholites are synthesized by the double substitution of (Ca 2+ , PO 4 3- ) by (Ln 3+ , SiO 4 4- ). Trivalent lanthanides are chemical analogues of trivalent actinides. The synthesis was performed with La, Nd and Eu. This study allows to demonstrate that the chemical immobilization comes from the fixation of rare earths at the atomic scale, thanks to their participation to the mineral structure. In part 1, the criteria for the formulation of a matrix for the conditioning of separate radionuclides are given. The structure and the different methods of apatite preparation are shown. Part 2 treats of the study of the solid solution, of the elaboration of the Ca 9 Nd 1 (SiO 4 ) 5 F 2 ceramic and of its physico chemical characterization. The last part deals with the localization of rare earths in the apatite structure, determined by europium luminescence and X-ray diffraction on monocrystal. (J.S.) [fr

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

    Science.gov (United States)

    2010-04-02

    hydroacoustic ), Earth (seismic), or atmosphere (infrasound); a fourth detects radioactive material from a nuclear test. Scientists concur that... Laboratory report LA-12942-MS, UC-700 and UC-703, June 1995, pp. 89-92, http://www.osti.gov/ bridge/servlets/purl/72900-YlaqIV/webviewable/72900.pdf...Seismic Results from DOE’s Non-Proliferation Experiment: A Comparison of Chemical and Nuclear Explosions,” Lawrence Livermore National Laboratory

  17. 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

  18. Report of a consultants' meeting on containment and confinement performance in NPPS with WWER 440/213 and 440/230 reactors

    International Nuclear Information System (INIS)

    1994-01-01

    The reactors WWER 440 model V-230 being operated in several countries of central and eastern Europe were designed in the 1960-ies assuming that the design basis accident can be limited to a break of a pipe of 100 mm diameter fitted with an orifice of 32 mm diameter. This assumption presently judged to be inadequate according to the international safety requirements, was used to determine the Emergency Core Cooling System. For this size of the break, the spray system of the reactor building was designed to prevent uncontrolled releases to the environment. For larger breaks, till full rupture of one of two pressurizer surge lines of 200 mm diameter each, large flaps were designed to open during the initial pressure peak after the accident and release to the environment large amounts of steam escaping from the broken reactor coolant system (RCS). In view of this initial release of radioactive medium and due to its other unsatisfactory features, the building surrounding the WWER 440/230 reactor is not called a containment, but is said to fulfill a confining function, in short it is termed a confinement. The shortcomings of the WWER 440/230 confinements have been long recognized and the newer model of WWER 440 units, called model V-213, has been provided with an improved building of higher leaktightness and higher strength, containing a complicated system of water trays and air traps designed to condense steam and remove a part of the air from the reactor compartments after large break loss of coolant accident (LB LOCA). For such case, the system provides fast pressure decrease in the containment to subatmospheric values and thus prevents any significant releases of radioactivity to the environment. As the main process is that of steam bubble condensation in water, the system is called bubbler condenser containment. The bubbler condenser containment was the answer to the problem of providing the old WWER 440 model 230 units with a reliable containment without excessive

  19. 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

  20. Thermal Analysis of Storage Cans Containing Special Nuclear Materials

    Energy Technology Data Exchange (ETDEWEB)

    Jerrell, J.W.

    2000-11-17

    A series of thermal analyses have been completed for ten storage can configurations representing various cases of materials stored in F-Area. The analyses determine the temperatures of the cans, the special nuclear material, and the air sealed within the cans. Analyses to aid in understanding the effect of oxide accumulation and metal aging on temperatures are also included.

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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)

  6. Process for the manufacture of nuclear fuels containing gadolinium

    International Nuclear Information System (INIS)

    Sondermann, T.

    1980-01-01

    A uranyl nitrate solution containing the required gadolinium CO 2 gas and NH 3 gas is taken to a precipitator with an aqueous condensate containing ammonium carbonate with a pH value of 9. A pump circulates the suspension. The salt is precipitated by reducing the pH value to 8.2. After filtration and drying these crystals together with AUC are changed to the oxide form, which can be sintered. (RW) [de

  7. 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)

  8. 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

  9. 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)

  10. 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

  11. INSTALLATION OF A POST-ACCIDENT CONFINEMENT HIGH-LEVEL RADIATION MONITORING SYSTEM IN THE KOLA NUCLEAR POWER STATION (UNIT 2) IN RUSSIA

    Energy Technology Data Exchange (ETDEWEB)

    GREENE,G.A.; GUPPY,J.G.

    1998-09-01

    This is the final report on the INSP project entitled, ``Post-Accident Confinement High-Level Radiation Monitoring System'' conducted by BNL under the authorization of Project Work Plan WBS 1.2.2.6 (Attachment 1). This project was initiated in February 1993 to assist the Russians in reducing risks associated with the continued operation of older Soviet-designed nuclear power plants, specifically the Kola VVER-440/230 Unit 2, through improved accident detection capability, specifically by the installation of a dual train high-level radiation detection system in the confinement of Unit 2 of the Kola NPP. The major technical objective of this project was to provide, install and make operational the necessary hardware inside the confinement of the Kola NPP Unit 2 to provide early and reliable warning of the release of radionuclides from the reactor into the confinement air space as an indication of the occurrence of a severe accident at the plant. In addition, it was intended to provide hands-on experience and training to the Russian plant workers in the installation, operation, calibration and maintenance of the equipment in order that they may use the equipment without continued US assistance as an effective measure to improve reactor safety at the plant.

  12. INSTALLATION OF A POST-ACCIDENT CONFINEMENT HIGH-LEVEL RADIATION MONITORING SYSTEM IN THE KOLA NUCLEAR POWER STATION (UNIT 2) IN RUSSIA

    International Nuclear Information System (INIS)

    GREENE, G.A.; GUPPY, J.G.

    1998-01-01

    This is the final report on the INSP project entitled, ''Post-Accident Confinement High-Level Radiation Monitoring System'' conducted by BNL under the authorization of Project Work Plan WBS 1.2.2.6 (Attachment 1). This project was initiated in February 1993 to assist the Russians in reducing risks associated with the continued operation of older Soviet-designed nuclear power plants, specifically the Kola VVER-440/230 Unit 2, through improved accident detection capability, specifically by the installation of a dual train high-level radiation detection system in the confinement of Unit 2 of the Kola NPP. The major technical objective of this project was to provide, install and make operational the necessary hardware inside the confinement of the Kola NPP Unit 2 to provide early and reliable warning of the release of radionuclides from the reactor into the confinement air space as an indication of the occurrence of a severe accident at the plant. In addition, it was intended to provide hands-on experience and training to the Russian plant workers in the installation, operation, calibration and maintenance of the equipment in order that they may use the equipment without continued US assistance as an effective measure to improve reactor safety at the plant

  13. Studies about pressure variations and their effects during a fire in a confined and forced ventilated enclosure: safety consequences in the case of a nuclear facility

    International Nuclear Information System (INIS)

    Hugues Pretrel; Laurent Bouilloux; Jerome Richard

    2005-01-01

    Full text of publication follows: In a nuclear facility, the cells are confined and forced ventilated and some of them are equipped with isolation devices designed to close in case of a fire. So, if a fire occurred, the pressure variations in the cell could be important. This contribution presents the safety concerns related to pressure variation effects (propagation of smokes and/or flames through the fire barriers, propagation of radioactive material) and the research works carried out by the french 'Institut de Radioprotection et de Surete Nucleaire' (IRSN) on this topic. These research works are composed of two different studies. The first study permits to quantify the overpressure and depression levels and to reveal the influence of the fire heat release rate (HRR), of the characteristics of the cell, of the ventilation layout (especially the airflow resistances of the ventilation branches) and of the control of the fire dampers. This study is based on three sets of experimental tests performed in three large-scale facilities of various dimensions (3600 m3, 400 m3 and 120 m3 in volume) and with several settings of the ventilation network. The analysis focuses on the conditions that lead to significant overpressure and depression peaks and identifies the level of fire HRR and airflow resistances for which pressure peaks may become a safety concern. The second study allows to characterise the behaviour of sectorisation and containment equipments subject to pressure stresses. The mechanical resistance of some equipments (doors, fire dampers) subject to pressure stresses as well as the aeraulic behaviour of this equipment (gas leak rates) are determined in order to assess the potential transfer of contamination in the ventilation networks. (authors)

  14. 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)

  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. 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.

  17. 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)

  18. 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)

  19. 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.

  20. 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

  1. 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

  2. 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.)

  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. Detecting special nuclear materials in containers using high-energy gamma rays emitted by fission products

    Science.gov (United States)

    Norman, Eric B.; Prussin, Stanley G.

    2007-10-02

    A method and a system for detecting the presence of special nuclear materials in a container. The system and its method include irradiating the container with an energetic beam, so as to induce a fission in the special nuclear materials, detecting the gamma rays that are emitted from the fission products formed by the 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.

  5. Special criteria developed for the design and analysis of floating nuclear plant containment structures

    International Nuclear Information System (INIS)

    Tsai, J.C.; Orr, R.S.

    1977-01-01

    Offshore Power Systems (OPS) is designing and manufacturing standardized Floating Nuclear Plants (FNP). The plant utilizes an ice-condenser and dual containment system with a free-standing steel shell welded to the upper deck of the steel platform. The pressure retaining portion of the platform upper deck in the containment area is designated as the containment baseplate. Certain unique loadings and configurations dictate the development of special criteria for the design and analysis of containment and other safety-related structures. Three topics of these special criteria are presented herein: Containment Shell Buckling Criteria; Containment Baseplate Design Criteria; Criteria for Missile Impact on Steel Panels

  6. 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)

  7. 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

  8. 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

  9. 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

  10. 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)

  11. 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

  12. Quark confinement

    International Nuclear Information System (INIS)

    Joos, H.

    1976-07-01

    The main topics of these lectures are: phenomenological approach to quark confinement, standard Lagrangian of hadrondynamics, Lagrangian field theory and quark confinement, classical soliton solutions in a simple model, quantization of extended systems, colour charge screening and quantization on a lattice and remarks on applications. A survey of the scientific publications listed according to the topics until 26 March 1976 is supplemented. (BJ) [de

  13. 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)

  14. 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

  15. 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

  16. Localization of a sound source in in a guided medium and reverberating field. Contribution to a study on leak localization in the internal wall of containment of a nuclear reactor in the case of a severe reactor accident; Localisation d`une source acoustique en milieu guide et champ reverberant. Contribution a l`etude sur la localisation de fuite de l`enceinte de confinement d`une centrale nucleaire en situation accidentelle grave

    Energy Technology Data Exchange (ETDEWEB)

    Thomann, F

    1996-11-28

    Basic data necessary for the localization of a leak in the internal wall of the containment are presented by studying the sound generated by gas jets coming out of (leaking fissures) as well as propagation in a guided medium. The results acquired have led us to choose the simple intercorrelation method and the matched filed processing method, both of which are likely to adequately handle our problems. Whereas the intercorrelation method appears to be limited in scope when dealing in the guided medium, the matched field processing is suited to leak localization over a surface of approximately 1000 m{sup 2} (for a total surface of 10 000 m{sup 2}). Preliminary studies on the leak signal and on replicated signals have led us to limit the frequency band to 2600 - 3000 Hz. We have succeeded in locating a leak situated in an ordinary position with a minimum amount of replicated signals and basic data. We have improved on the estimation of Bartlett and MVDE (minimum variance distortion less filter) rendering them even more effective. Afterwards, we considered the severe accident situation and showed that the system can be installed in situ. (author) 88 refs.

  17. 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.

  18. 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)

  19. Special criteria developed for the design and analysis of floating nuclear plant containment structures

    International Nuclear Information System (INIS)

    Tsai, J.C.; Orr, R.S.

    1977-01-01

    Offshore Power Systems (OPS) is designing and manufacturing standardized Floating Nuclear Plants (FNP). The plant utilizes an ice-condenser and dual containment system with a free-standing steel shell welded to the upper deck of the steel platform. The pressure retaining portion of the platform upper deck in the containment area is designated as the containment baseplate. Certain unique loadings and configurations dictate the development of special criteria for the design and analysis of containment and other safety related structures. Three topics of these special criteria are presented herein. (Auth.)

  20. Fusion, magnetic confinement

    Energy Technology Data Exchange (ETDEWEB)

    Berk, H.L.

    1992-08-06

    An overview is presented of the principles of magnetic confinement of plasmas for the purpose of achieving controlled fusion conditions. Sec. 1 discusses the different nuclear fusion reactions which can be exploited in prospective fusion reactors and explains why special technologies need to be developed for the supply of tritium or {sup 3}He, the probable fuels. In Sec. 2 the Lawson condition, a criterion that is a measure of the quality of confinement relative to achieving fusion conditions, is explained. In Sec. 3 fluid equations are used to describe plasma confinement. Specific confinement configurations are considered. In Sec. 4 the orbits of particle sin magneti and electric fields are discussed. In Sec. 5 stability considerations are discussed. It is noted that confinement systems usually need to satisfy stability constraints imposed by ideal magnetohydrodynamic (MHD) theory. The paper culminates with a summary of experimental progress in magnetic confinement. Present experiments in tokamaks have reached the point that the conditions necessary to achieve fusion are being satisfied.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. Plasma confinement

    CERN Document Server

    Hazeltine, R D

    2003-01-01

    Detailed and authoritative, this volume examines the essential physics underlying international research in magnetic confinement fusion. It offers readable, thorough accounts of the fundamental concepts behind methods of confining plasma at or near thermonuclear conditions. Designed for a one- or two-semester graduate-level course in plasma physics, it also represents a valuable reference for professional physicists in controlled fusion and related disciplines.

  3. Analysis of temperature and pressure distribution of containers for nuclear waste material disposal in space

    Science.gov (United States)

    Vanbibber, L. E.; Parker, W. G.

    1973-01-01

    A computer program was adapted from a previous generation program to analyze the temperature and internal pressure response of a radioactive nuclear waste material disposal container following impact on the earth. This program considers component melting, LiH dissociation, temperature dependent properties and pressure and container stress response. Analyses were performed for 21 cases with variations in radioactive power level, container geometry, degree of deformation of the container, degree of burial and soil properties. Results indicated that the integrity of SS-316 containers could be maintained with partial burials of either underformed or deformed containers. Results indicated that completely buried waste containers, with power levels above 5 kW, experienced creep stress rupture failures in 4 to 12 days.

  4. 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

  5. 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.

  6. 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

  7. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Floyd, D.; Edstrom, C. [Manufacturing Sciences Corp. (United States); Biddle, K.; Orlowski, R. [BNFL, Inc. (United States); Geinitz, R. [Safe Sites of Colorado, Golden, CO (United States); Keenan, K. [USDOE-RFFO (United States); Rivera, M. [Science Applications International Corp./LATA (United States)

    1997-03-01

    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.

  9. 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

  10. 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

  11. 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)

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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)

  17. 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

  18. Effects of strain and quantum confinement in optically pumped nuclear magnetic resonance in GaAs: Interpretation guided by spin-dependent band structure calculations

    Science.gov (United States)

    Wood, R. M.; Saha, D.; McCarthy, L. A.; Tokarski, J. T.; Sanders, G. D.; Kuhns, P. L.; McGill, S. A.; Reyes, A. P.; Reno, J. L.; Stanton, C. J.; Bowers, C. R.

    2014-10-01

    A combined experimental-theoretical study of optically pumped nuclear magnetic resonance (OPNMR) has been performed in a GaAs /A l0.1G a0.9As quantum well film epoxy bonded to a Si substrate with thermally induced biaxial strain. The photon energy dependence of the Ga OPNMR signal was recorded at magnetic fields of 4.9 and 9.4 T at a temperature of 4.8-5.4 K. The data were compared to the nuclear spin polarization calculated from the electronic structure and differential absorption to spin-up and spin-down states of the electron conduction band using a modified k .p model based on the Pidgeon-Brown model. Comparison of theory with experiment facilitated the assignment of features in the OPNMR energy dependence to specific interband Landau level transitions. The results provide insight into how effects of strain and quantum confinement are manifested in optical nuclear polarization in semiconductors.

  19. 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

  20. Report by the AERES on the unit: Unit of researches on neutron transport and radioactivity confinement in nuclear installations under the supervision of the establishments and bodies: IRSN

    International Nuclear Information System (INIS)

    2010-07-01

    This report is a kind of audit report on a research laboratory whose activity is organized in three departments: neutron transport and criticality (themes: numerical methods, maths and statistics related to the simulation of neutral particle propagation, nuclear data, uncertainty propagation and bias estimation, code qualification and associated experimental programs, neutron transport in reactors and fuel cycle, criticality accidents), radionuclide transfer in radioactive waste disposals (site identification strategy, hydro-mechanical phenomena affecting storage performance, physical-chemical evolution factors, storage modelling), and metrology and confinement of radioactive gases and aerosols. The authors discuss an assessment of the unit activities in terms of strengths and opportunities, aspects to be improved and recommendations, productions and publications. A more detailed assessment is presented for each department in terms of scientific quality, influence and attractiveness (awards, recruitment capacity, capacity to obtain financing and to tender, participation to international programs), strategy and governance, and project

  1. 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)

  2. Thermal and stress analysis of hot isostatically pressed, alumina ceramic, nuclear waste containers

    International Nuclear Information System (INIS)

    Chang, Yun; Hoenig, C.L.

    1990-01-01

    The Yucca Mountain Project is studying design and fabrication options for a safe durable container in which to store nuclear waste underground at Yucca Mountain, Nevada. The ceramic container discussed here is an alternative to using a metal container. This ceramic alternative would be selected if site conditions prove too corrosive to use metals for nuclear waste storage. Some of the engineering problems addressed in this study were: the stress generated in the alumina container by compressive loads when 4000 to 40,000 psi of external pressure is applied; the thermal stress in the container during the heating and cooling processes; the temperature histories of the container in various production scenarios and the power required for typical heaters; the fastest possible turnaround time to heat, seal, and cool the container commensurate with preserving the structural integrity of the ceramic and the closure; the testing of some commercial heating elements to determine the maximum available heat output; and the trade-offs between the minimization in thermal stress and cycle time for closure. 2 refs., 23 figs., 2 tabs

  3. 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)

  4. 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.

  5. 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.

  6. 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

  7. 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

  8. 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

  9. Retention capacity of samarium (III) in zircon for it possible use in retaining walls for confinement of nuclear residues; Capacidad de retencion de samario (III) en circon para su posible uso en barreras de contencion para confinamiento de residuos nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Garcia G, N

    2006-07-01

    Mexico, as country that produces part of its electric power by nuclear means, should put special emphasis in the development of technologies guided to the sure and long term confinement of the high level nuclear residuals. This work studies the capacity that has the natural zircon to retain to the samarium (III) in solution, by what due, firstly, to characterize the zircon for technical instrumental to determine the purity and characteristic of the mineral in study. The instrumental techniques that were used to carry out the physicochemical characterization were the neutron activation analysis (NAA), the infrared spectroscopy (IS), the thermal gravimetric analysis (TGA), scanning electron microscopy (SEM), transmission electron microscopy (TEM), semiquantitative analysis, dispersive energy spectroscopy (EDS), X-ray diffraction (XRD) and luminescence technique. The characterization of the surface properties carries out by means of the determination of the surface area using the BET multipoint technique, acidity constants, hydration time, the determination of the point of null charge (pH{sub PCN}) and density of surface sites (D{sub s}). The luminescence techniques were useful to determine the optimal point hydration of the zircon and for the quantification of the samarium, for that here intends the development of both analysis techniques. With the adjustment of the titration curves in the FITEQL 4 package the constants of surface acidity in the solid/liquid interface were determined. To the finish of this study it was corroborated that the zircon is a mineral that presents appropriate characteristics to be proposed as a contention barrier for the deep geologic confinement. With regard to the study of adsorption that one carries out the samarium retention it is superior to 90% under the described conditions. This investigation could also be applicable in the confinement of dangerous industrial residuals. (Author)

  10. 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.

  11. 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)

  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. 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

  14. Container materials for high-level nuclear waste at the proposed Yucca Mountain site

    International Nuclear Information System (INIS)

    McCright, R.D.

    1991-05-01

    Candidate container materials for high-level nuclear waste packages to be emplaced at the proposed Yucca Mountain repository site are being considered for their long-term resistance to corrosion, oxidation, embrittlement and other kinds of degradation. An extensive compilation of the degradation phenomena for six candidate materials is complete, and further studies have begun on the degradation modes affecting additional candidate materials. Phenomenological models for predicting container degradation rates are being advanced for environmental conditions applicable to Yucca Mountain. An experimental program is underway to evaluate the susceptibility of container materials is localized corrosion, stress corrosion cracking, and enhancement of corrosion and oxidation attack by gamma radiation. Initial evaluations of container fabrication and welding processes have identified some processes that appear to alleviate some long-term susceptibility concerns. 10 refs., 2 figs

  15. 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

  16. 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

  17. 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.

  18. Structural integrity analysis of the degraded drywell containment at the Oyster Creek Nuclear generating station.

    Energy Technology Data Exchange (ETDEWEB)

    Petti, Jason P.

    2007-01-01

    This study examines the effects of the degradation experienced in the steel drywell containment at the Oyster Creek Nuclear Generating Station. Specifically, the structural integrity of the containment shell is examined in terms of the stress limits using the ASME Boiler and Pressure Vessel (B&PV) Code, Section III, Division I, Subsection NE, and examined in terms of buckling (stability) using the ASME B&PV Code Case N-284. Degradation of the steel containment shell (drywell) at Oyster Creek was first observed during an outage in the mid-1980s. Subsequent inspections discovered reductions in the shell thickness due to corrosion throughout the containment. Specifically, significant corrosion occurred in the sandbed region of the lower sphere. Since the presence of the wet sand provided an environment which supported corrosion, a series of analyses were conducted by GE Nuclear Energy in the early 1990s. These analyses examined the effects of the degradation on the structural integrity. The current study adopts many of the same assumptions and data used in the previous GE study. However, the additional computational recourses available today enable the construction of a larger and more sophisticated structural model.

  19. 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)

  20. 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?

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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.

  6. 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

  7. Crevice corrosion and the lifetime performance of titanium nuclear waste containers

    International Nuclear Information System (INIS)

    Ikeda, B.M.; Noel, J.J.; Shoesmith, D.W.

    1997-12-01

    A multibarrier system is being proposed for the safe disposal of Canada's nuclear fuel waste. The container in which the fuel wastes are emplaced is an essential barrier and one that is expected to remain intact for at least 500 a, and possibly 10 5 a. Titanium is one candidate material that has been evaluated as a potential container material. The possible localized corrosion processes which could occur on titanium have been extensively reviewed and only crevice corrosion and hydrogen induced cracking (HIC) merit consideration under Canadian waste vault conditions. HIC is dealt with in a separate report. In this report we discuss the ability to predict titanium-container lifetimes based on current crevice corrosion modeling capabilities. An analysis of available data on the initiation, propagation and repassivation stages of crevice corrosion shows a variety of ways that titanium-container lifetime can be extended beyond the 1000-a lifetime predicted in our previously established container failure model. To justify the use of any material for construction of a nuclear fuel waste container requires a computational model to predict the time to failure. The previously reported container failure function consisted of three components: a statistical evaluation of early container failure; a low temperature hydrogen induced cracking failure mode; and a temperature dependent crevice corrosion failure mode. This model predicted lifetimes >1000 a for 99.9% of the containers in the disposal vault. An independent assessment of corrosion damage based on the empirical relationship between maximum penetration depth due to crevice corrosion and oxygen consumption agreed with the predictions of the container failure function. The data requirement for establishing the crevice corrosion limits of Grades-2, -12 and -16 titanium and for improving the arguments for the long lifetime prediction are presented. An extensive database including critical oxygen concentration for

  8. 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)

  9. Method of constructing primary containment vessel for large-capacity nuclear power plant

    International Nuclear Information System (INIS)

    Fukunaga, Teruaki; Nakano, Kazuo; Kyoda, Shigeru

    1982-01-01

    The latest trends of large-capacity nuclear power plants such as is rated 1,100 MWe are the increase in materials handled and terms of construction work accompanying the growth of rated capacity. From the viewpoint of shortening the term of construction, Toshiba has made research and investigation for improving various methods of construction. The Unit No. 3 in the Fukushima Daini Nuclear Power Plant of Tokyo Electric Power Co., which is now under construction, is to be curtailed of its term of construction work through the application of the new method. Description is made here of the new method of construction for the primary containment vessel which has completed its initial pressure-leak test in last April. (author)

  10. 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.)

  11. 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

  12. 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

  13. 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.

  14. Water/rock interactions and mass transport within a thermal gradient Application to the confinement of high level nuclear waste

    International Nuclear Information System (INIS)

    Poinssot, Ch.; Ecole Normale Superieure, 92 - Fontenay-aux-Roses

    1998-01-01

    The initial stage of a high level nuclear waste disposal will be characterised by a large heat release within the near-field environment of the canisters. This heat flux caused by radioactive decay will lead to an increase of temperature and a subsequent thermal gradient between the 'hot' canisters and the 'cold'geological medium. In addition, this thermal gradient will decrease with time due to the heat decay although it could last hundred years. What will be the consequences of such a thermal field varying both on space and time for the alteration of the different constituents of the near field environment. In particular, what could be the effects on the radionuclides migration in the accidental case of an early breach of a canister during the thermal stage? This study brings significant answers to these questions in the light of a performance assessment study. This work is supported by a triple methodological approach involving experimental studies, modelling calculations and a natural analogues study. This complete work demonstrates that a thermal gradient leads to a large re-distribution of elements within the system: some elements are incorporated in the solid phases of the hot end (Si, Zr, Ca) whereas some others are in those of the cold end (Fe, Al, Zn). The confrontation of the results of very simple experiments with the results of a model built on equilibrium thermodynamics allow us to evidence the probable mechanisms causing this mass transport: out-of-equilibrium thermodiffusion processes coupled to irreversible precipitation. Moreover, the effects of the variation of temperatures with time is studied by the way of a natural system which underwent a similar temperature evolution as a disposal and which was initially rich in uranium: the Jurassic Alpine bauxites. In addition, part of the initial bauxite escaped this temperature transformations due to their incorporation in outer thrusting nappes. They are used as a reference. (author)

  15. 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.

  16. 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)

  17. 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.

  18. 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.)

  19. 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.

  20. 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.

  1. 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.

  2. Nuclear Magnetic Resonance and Elastic Wave Velocity of Chalk Saturated with Brines Containing Divalent Ions

    DEFF Research Database (Denmark)

    Katika, Konstantina; Alam, Mohammad Monzurul; Fabricius, Ida Lykke

    Nuclear magnetic resonance (NMR) has proven a good technique for measuring pore size distribution in reservoir rocks. The use of low field NMR together with sonic and electrical resistivity measurements, can contribute to illustrate the effect of adsorbing ions on chalk elasticity. NMR is useful...... divided into groups of three and each group was saturated either with deionized water, calcite equilibrated water, or sodium chloride, magnesium chloride and calcium chloride solutions of the same ionic strength. Saturation with solutions that contain divalent ions caused major shifts in the distribution...

  3. 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

  4. 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

  5. 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)

  6. Topological confinement and superconductivity

    Energy Technology Data Exchange (ETDEWEB)

    Al-hassanieh, Dhaled A [Los Alamos National Laboratory; Batista, Cristian D [Los Alamos National Laboratory

    2008-01-01

    We derive a Kondo Lattice model with a correlated conduction band from a two-band Hubbard Hamiltonian. This mapping allows us to describe the emergence of a robust pairing mechanism in a model that only contains repulsive interactions. The mechanism is due to topological confinement and results from the interplay between antiferromagnetism and delocalization. By using Density-Matrix-Renormalization-Group (DMRG) we demonstrate that this mechanism leads to dominant superconducting correlations in aID-system.

  7. 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

  8. 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

  9. Special section containing papers presented at the 13th IAEA Technical Meeting on Energetic Particles in Magnetic Confinement Systems (Beijing, China, 17-20 September 2013) Special section containing papers presented at the 13th IAEA Technical Meeting on Energetic Particles in Magnetic Confinement Systems (Beijing, China, 17-20 September 2013)

    Science.gov (United States)

    Lin, Z.

    2014-10-01

    In magnetic fusion plasmas, a significant fraction of the kinetic pressure is contributed by superthermal charged particles produced by auxiliary heating (fast ions and electrons) and fusion reactions (a-particles). Since these energetic particles are often far away from thermal equilibrium due to their non-Maxwellian distribution and steep pressure gradients, the free energy can excite electromagnetic instabilities to intensity levels well above the thermal fluctuations. The resultant electromagnetic turbulence could induce large transport of energetic particles, which could reduce heating efficiency, degrade overall plasma confinement, and damage fusion devices. Therefore, understanding and predicting energetic particle confinement properties are critical to the success of burning plasma experiments such as ITER since the ignition relies on plasma self-heating by a-particles. To promote international exchanges and collaborations on energetic particle physics, the biannual conference series under the auspices of the International Atomic Energy Agency (IAEA) were help in Kyiv (1989), Aspenas (1991), Trieste (1993), Princeton (1995), JET/Abingdon (1997), Naka (1999), Gothenburg (2001), San Diego (2003), Takayama (2005), Kloster Seeon (2007), Kyiv (2009), and Austin (2011). The papers in this special section were presented at the most recent meeting, the 13th IAEA Technical Meeting on Energetic Particles in Magnetic Confinement Systems, which was hosted by the Fusion Simulation Center, Peking University, Beijing, China (17-20 September 2013). The program of the meeting consisted of 71 presentations, including 13 invited talks, 26 oral contributed talks, 30 posters, and 2 summary talks, which were selected by the International Advisory Committee (IAC). The IAC members include H. Berk, L.G. Eriksson, A. Fasoli, W. Heidbrink, Ya. Kolesnichenko, Ph. Lauber, Z. Lin, R. Nazikian, S. Pinches, S. Sharapov, K. Shinohara, K. Toi, G. Vlad, and X.T. Ding. The conference program

  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. 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

  12. 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.

  13. Dynamics and reactivity of confined water

    International Nuclear Information System (INIS)

    Musat, R.

    2008-01-01

    In the context of new sustainable energy sources quest, the nuclear energy remains a key solution. However, with the development of nuclear technology, problems relating to nuclear waste disposal arise; thus, the radiolysis of water in confined media is extremely important with respect to matters related to long time storage of nuclear waste. Studies in model porous media would allow the projection of a confined water radiolysis simulator. A first step in this direction was made by studying the radiolysis of water confined in Vycor and CPG glasses; this study continues the trend set and investigates the effects of confinement in metal materials upon the water radiolysis allowing the understanding of metal - water radiation induced corrosion. A further/complete understanding of the radiolytic process under confinement requires knowledge of the effect of confinement upon the dynamics of confined molecules and on the evolution of the species produced upon ionizing radiation. In this respect, we have used the OH vibrator as a probe of the hydrogen bond network properties and thus investigated the dynamics of confined water using IR time resolved spectroscopy. The evolution of the hydrated electron under confinement was studied on a nano and picosecond time scale using UV pump - visible probe technique and single shot spectroscopy. (author) [fr

  14. 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

  15. 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)

  16. 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.

  17. 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

  18. 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

  19. 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.

  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. 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

  2. Magnetic and electrostatic confinement of plasma with tuning of electrostatic field

    Science.gov (United States)

    Rostoker, Norman [Irvine, CA; Binderbauer, Michl [Irvine, CA; Qerushi, Artan [Irvine, CA; Tahsiri, Hooshang [Irvine, CA

    2008-10-21

    A system and method for containing plasma and forming a Field Reversed Configuration (FRC) magnetic topology are described in which plasma ions are contained magnetically in stable, non-adiabatic orbits in the FRC. Further, the electrons are contained electrostatically in a deep energy well, created by tuning an externally applied magnetic field. The simultaneous electrostatic confinement of electrons and magnetic confinement of ions avoids anomalous transport and facilitates classical containment of both electrons and ions. In this configuration, ions and electrons may have adequate density and temperature so that upon collisions they are fused together by nuclear force, thus releasing fusion energy. Moreover, the fusion fuel plasmas that can be used with the present confinement system and method are not limited to neutronic fuels only, but also advantageously include advanced fuels.

  3. 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)

  4. Magnetic confinement

    Energy Technology Data Exchange (ETDEWEB)

    Batistoni, Paola; De Marco, Francesco; Pieroni, Leonardo (ed.)

    2005-07-01

    The Frascati Tokamak Upgrade (FTU) is a compact, high-magnetic-field tokamak capable of operating at density and magnetic field values similar to, or even encompassing, those of International Thermonuclear Experimental Reactor (ITER) and therefore provides a unique opportunity to explore physics issues that are directly relevant to ITER. During 2004 the experimental activities were focussed on fully exploiting the lower hybrid system (for generating and controlling the plasma current) and the electron cyclotron heating system (joint experiment with the Institute of Plasma Physics of the National Research Council, Milan). With all four gyrotrons in operation, full electron cyclotron power was achieved up to a record level of 1.5 MW. By simultaneously injecting lower hybrid waves, to tailor the plasma current radial profile, and electron cyclotron waves, to heat the plasma centre, good confinement regimes with internal transport barriers were obtained at the highest plasma density values ever achieved for this operation regime (n {approx}1.5X10{sup 20}m{sup -3}). Specific studies were devoted to optimising the coupling of lower hybrid waves to the plasma (by real-time control of the plasma position) and to generating current by electron cyclotron current drive. The new scanning CO{sub 2} interferometer (developed by the Reversed Field Experiment Consortium) for high spatial and time resolution (1 cm/50 {mu}s) density profile measurements was extensively used. The Thomson scattering diagnostic was upgraded and enabled observation of scattered signals associated with the Confinement background plasma dynamics. As for theoretical studies on the dynamics of turbulence in plasmas, the transition from Bohm-like scaling to gyro-Bohm scaling of the local plasma diffusivity was demonstrated on the basis of a generalised four wave model (joint collaboration with Princeton Plasma Physics Laboratory and the University of California at Irvine). The transition from weak to strong

  5. Magnetic confinement

    International Nuclear Information System (INIS)

    Batistoni, Paola; De Marco, Francesco; Pieroni, Leonardo

    2005-01-01

    The Frascati Tokamak Upgrade (FTU) is a compact, high-magnetic-field tokamak capable of operating at density and magnetic field values similar to, or even encompassing, those of International Thermonuclear Experimental Reactor (ITER) and therefore provides a unique opportunity to explore physics issues that are directly relevant to ITER. During 2004 the experimental activities were focussed on fully exploiting the lower hybrid system (for generating and controlling the plasma current) and the electron cyclotron heating system (joint experiment with the Institute of Plasma Physics of the National Research Council, Milan). With all four gyrotrons in operation, full electron cyclotron power was achieved up to a record level of 1.5 MW. By simultaneously injecting lower hybrid waves, to tailor the plasma current radial profile, and electron cyclotron waves, to heat the plasma centre, good confinement regimes with internal transport barriers were obtained at the highest plasma density values ever achieved for this operation regime (n ∼1.5X10 20 m -3 ). Specific studies were devoted to optimising the coupling of lower hybrid waves to the plasma (by real-time control of the plasma position) and to generating current by electron cyclotron current drive. The new scanning CO 2 interferometer (developed by the Reversed Field Experiment Consortium) for high spatial and time resolution (1 cm/50 μs) density profile measurements was extensively used. The Thomson scattering diagnostic was upgraded and enabled observation of scattered signals associated with the Confinement background plasma dynamics. As for theoretical studies on the dynamics of turbulence in plasmas, the transition from Bohm-like scaling to gyro-Bohm scaling of the local plasma diffusivity was demonstrated on the basis of a generalised four wave model (joint collaboration with Princeton Plasma Physics Laboratory and the University of California at Irvine). The transition from weak to strong energetic particle

  6. 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

  7. 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

  8. 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

  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. 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.)

  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. 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)

  13. 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 2 O 3 - Al 2 O 3 - Na 2 O - Li 2 O - CaO - La 2 O 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

  14. 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.

  15. 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

  16. 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

  17. 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

  18. 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

  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. 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)

  1. 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.)

  2. 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

  3. Sizing of the overpacks containing high-level radioactive waste packages from French nuclear facilities

    International Nuclear Information System (INIS)

    Codau, Claudia; Fernandes, Roméo

    2017-01-01

    This work is a contribution to the project Cigeo (Geological Storage Industrial Center: www.cigeo.com) which aims to store in a deep geological layer, intermediate, high-level and long-lived radioactive waste from all French nuclear facilities. The project is led by ANDRA (French National Radioactive Waste Management Agency). EDF (Electricity of France), as the nuclear electricity producer, will be a beneficiary of the facility and follows it for reasons of consistency with the rest of the nuclear cycle. Packages of the most radioactive vitrified waste from the processing of spent fuel will be placed in steel overpacks (storage containers) whose main function will be to protect the waste from water during the so-called thermal phase of the glass (t° glass < 50°C - 70°C). In addition to the waterproof criterion, the challenges associated with the design of the overpacks are: To limit the production of hydrogen due to corrosion (∼50.000 overpacks). To increase glass life: a decrease in steel thickness reduces the amount of corrosion products and improves the longevity of the glass. To reduce production costs. The objective of this work is to implement the radiological dimensioning methodology based on neutron modeling. As an illustration of what this method allows to do, a sizing of the steel overpack is proposed in order to achieve these technical and economic specifications. The overpack is mainly designed by taking into account the mechanical load that will be applied and degradation due to corrosion, particularly the corrosion by radiolysis. According to ANDRA, the most important criterion is to maintain the dose rate below 10 Gy/h at a distance of 5 cm from the overpack interface. The dose rates were calculated with the TRIPOLI4 code using as input data the temporal radiological evolution of waste obtained with DARWIN/PEPIN2. This work has shown that, in the most penalizing case among the studied configurations, a steel thickness of 4.6 cm for the overpack

  4. 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

  5. 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)

  6. Comments on confinement criteria

    International Nuclear Information System (INIS)

    Kurak, V.; Schroer, B.; Swieca, J.A.

    1977-01-01

    For a QED 2 model with SU(n) flavour, the nature of the physical states space is more subtle than one expects on the basis of the loop criterion for confinement. One may have colour confinement without confinement of the fundamental flavour representation. Attempts to formulate confinement criteria in which the quark fields play a more fundamental role are discussed [pt

  7. Development of 3D models of buildings for containment of the nuclear power plant of Almaraz and of the Trillo Nuclear with the GOTHIC 8.0 code

    International Nuclear Information System (INIS)

    Jimenez, G.; Bocanegra Melian, R.; Fernandez Cosils, K.; Barreira Pereira, P.; Rey Peinado, L.; Posada Barral, J. M.

    2014-01-01

    The objective of the first phase of the research of CNAT and the UPM project is the construction of several three-dimensional models detailed GOTHIC 8.0 code of containment of a buildings plant type PWR-W and KWU, corresponding to the Central Nuclear de Almaraz (CNA) and Trillo (CNT) respectively. (Author)

  8. Turbulent jet erosion of a stably stratified gas layer in a nuclear reactor test containment

    Energy Technology Data Exchange (ETDEWEB)

    Ishay, Liel [Department of Mechanical Engineering, Ben-Gurion University of the Negev, Beer-Sheva 84105 (Israel); Bieder, Ulrich [Commissariat à l’énergie atomique et aux énergies alternatives, Centre de SACLAY DEN/SAC/DANS/DM2S/STMF/LMSF, F-91191 Gif-sur-Yvette (France); Ziskind, Gennady [Department of Mechanical Engineering, Ben-Gurion University of the Negev, Beer-Sheva 84105 (Israel); Rashkovan, Alex, E-mail: rashbgu@gmail.com [Physics Department, Nuclear Research Center Negev (NRCN), PO Box 9001, Beer-Sheva 84190 (Israel)

    2015-10-15

    Highlights: • We model stably stratified layer erosion by vertical turbulent round jet. • Separate effect studies are performed as a platform for choosing modeling approach. • A test performed in MISTRA facility, CEA, Saclay is modeled using Fluent and Trio-U codes. • The proposed modeling approach showed good agreement with the MISTRA facility LOWMA-3 test. - Abstract: A number of integral and separate effect experiments were performed in the last two decades for validation of containment computational tools. The main goal of these benchmark experiments was to assess the ability of turbulence models and computational fluid dynamics codes to predict hydrogen concentration distribution and steam condensation rate in a nuclear reactor containment in the course of severe accidents. It appears from the published literature that the predictive capability of the existing computational tools still needs to be improved. This work examines numerically the temporal evolution of helium concentration in the experiment called LOWMA-3, performed in the MISTRA facility of CEA-Saclay, France. In the experiment, helium is used to mimic hydrogen of a real-case accident. The aim of this separate effect experiment, where steam condensation was not involved, is to predict helium concentration field. The conditions of the experiment are such that both the momentum transport and molecular diffusion contributions to the mixing process are of the same order of magnitude (Fr ∼ 1). A commercial CFD code, Fluent, and a CEA in-house code, Trio-U, are used for flow and helium concentration fields temporal evolution prediction in the present study. The preliminary separate effect studies provide guidance to an optimal modeling approach for the LOWMA-3 experiment. Temporal evolution of helium concentration in the stratification layer is shown, and a comparison to the experiment is discussed. It is shown that correct modeling of the round jet flowfield is essential for a reliable

  9. Turbulent jet erosion of a stably stratified gas layer in a nuclear reactor test containment

    International Nuclear Information System (INIS)

    Ishay, Liel; Bieder, Ulrich; Ziskind, Gennady; Rashkovan, Alex

    2015-01-01

    Highlights: • We model stably stratified layer erosion by vertical turbulent round jet. • Separate effect studies are performed as a platform for choosing modeling approach. • A test performed in MISTRA facility, CEA, Saclay is modeled using Fluent and Trio-U codes. • The proposed modeling approach showed good agreement with the MISTRA facility LOWMA-3 test. - Abstract: A number of integral and separate effect experiments were performed in the last two decades for validation of containment computational tools. The main goal of these benchmark experiments was to assess the ability of turbulence models and computational fluid dynamics codes to predict hydrogen concentration distribution and steam condensation rate in a nuclear reactor containment in the course of severe accidents. It appears from the published literature that the predictive capability of the existing computational tools still needs to be improved. This work examines numerically the temporal evolution of helium concentration in the experiment called LOWMA-3, performed in the MISTRA facility of CEA-Saclay, France. In the experiment, helium is used to mimic hydrogen of a real-case accident. The aim of this separate effect experiment, where steam condensation was not involved, is to predict helium concentration field. The conditions of the experiment are such that both the momentum transport and molecular diffusion contributions to the mixing process are of the same order of magnitude (Fr ∼ 1). A commercial CFD code, Fluent, and a CEA in-house code, Trio-U, are used for flow and helium concentration fields temporal evolution prediction in the present study. The preliminary separate effect studies provide guidance to an optimal modeling approach for the LOWMA-3 experiment. Temporal evolution of helium concentration in the stratification layer is shown, and a comparison to the experiment is discussed. It is shown that correct modeling of the round jet flowfield is essential for a reliable

  10. Inertial confinement fusion

    International Nuclear Information System (INIS)

    Nuckolls, J.H.; Wood, L.L.

    1988-01-01

    Edward Teller has been a strong proponent of harnessing nuclear explosions for peaceful purposes. There are two approaches: Plowshare, which utilizes macro- explosions, and inertial confinement fusion, which utilizes microexplosions. The development of practical fusion power plants is a principal goal of the inertial program. It is remarkable that Teller's original thermonuclear problem, how to make super high yield nuclear explosions, and the opposite problem, how to make ultra low yield nuclear explosions, may both be solved by Teller's radiation implosion scheme. This paper reports on the essential physics of these two thermonuclear domains, which are separated by nine orders of magnitude in yield, provided by Teller's similarity theorem and its exceptions. Higher density makes possible thermonuclear burn of smaller masses of fuel. The leverage is high: the scale of the explosion diminishes with the square of the increase in density. The extraordinary compressibility of matter, first noticed by Teller during the Los Alamos atomic bomb program, provides an almost incredible opportunity to harness fusion. The energy density of thermonuclear fuels isentropically compressed to super high-- -densities---even to ten thousand times solid density---is small compared to the energy density at thermonuclear ignition temperatures. In small masses of fuel imploded to these super high matter densities, the energy required to achieve ignition may be greatly reduced by exploiting thermonuclear propagation from a relatively small hot spot

  11. Measurement and modelling of reactive transport in geological barriers for nuclear waste containment.

    Science.gov (United States)

    Xiong, Qingrong; Joseph, Claudia; Schmeide, Katja; Jivkov, Andrey P

    2015-11-11

    Compacted clays are considered as excellent candidates for barriers to radionuclide transport in future repositories for nuclear waste due to their very low hydraulic permeability. Diffusion is the dominant transport mechanism, controlled by a nano-scale pore system. Assessment of the clays' long-term containment function requires adequate modelling of such pore systems and their evolution. Existing characterisation techniques do not provide complete pore space information for effective modelling, such as pore and throat size distributions and connectivity. Special network models for reactive transport are proposed here using the complimentary character of the pore space and the solid phase. This balances the insufficient characterisation information and provides the means for future mechanical-physical-chemical coupling. The anisotropy and heterogeneity of clays is represented using different length parameters and percentage of pores in different directions. Resulting networks are described as mathematical graphs with efficient discrete calculus formulation of transport. Opalinus Clay (OPA) is chosen as an example. Experimental data for the tritiated water (HTO) and U(vi) diffusion through OPA are presented. Calculated diffusion coefficients of HTO and uranium species are within the ranges of the experimentally determined data in different clay directions. This verifies the proposed pore network model and validates that uranium complexes are diffusing as neutral species in OPA. In the case of U(vi) diffusion the method is extended to account for sorption and convection. Rather than changing pore radii by coarse grained mathematical formula, physical sorption is simulated in each pore, which is more accurate and realistic.

  12. Evaluation of seismic behavior of soils under nuclear containment structures via dynamic centrifuge test

    International Nuclear Information System (INIS)

    Ha, Jeong Gon; Kim, Dong-Soo

    2014-01-01

    Highlights: • A series of dynamic centrifuge tests were performed for NPP structure to investigate the soil–foundation-structure interaction with various soil conditions from loose sand to weathered rock. • SFSI phenomena for NPP structure were observed directly using experimental method. • Effect of the soil stiffness and nonlinear characteristics on SFSI was estimated. • There are comparisons of the control motions for seismic design of a NPP structure. • Subsoil condition, earthquake intensity and control motion affected to seismic load. - Abstract: To evaluate the earthquake loads for the seismic design of a nuclear containment structure, it is necessary to consider the soil–foundation-structure interaction (SFSI) due to their interdependent behavior. Especially, understanding the effects of soil stiffness under the structure and the location of control motion to SFSI are very important. Motivated by these requirements, a series of dynamic centrifuge tests were performed with various soil conditions from loose sand to weathered rock (WR), as well as different seismic intensities for the bedrock motion. The different amplification characteristics in peak-accelerations profile and effects of soil-nonlinearity in response spectrum were observed. The dynamic behaviors were compared between surface of free-field and foundation of the structure for the evaluation of the control motion for seismic design. It was found that dynamic centrifuge test has potentials to estimate the seismic load considering SFSI

  13. Nuclear waste shipping container response to severe accident conditions, A brief critique of the modal study

    International Nuclear Information System (INIS)

    Audin, L.

    1990-12-01

    The Modal Study (NUREG/CR-4829) attempts to upgrade the analysis of spent nuclear fuel transportation accidents, and to verify the validity of the present regulatory scheme of cask performance standards as a means to minimize risk. While an improvement over many prior efforts in this area (such as NUREG-0170), it unfortunately fails to create a realistic simulation either of a shipping cask, the severe conditions to which it could be subjected, or the potential damage to the spent fuel cargo during an accident. There are too many deficiencies in its analysis to allow acceptance of its results for the presumed cask design, and many pending changes in new containers, cargoes and shipping patterns will limit applicability of the Modal Study to future shipments. In essence, the Modal Study is a good start, but is too simplistic, incomplete, outdated and open to serious question to be used as the basis for any present-day environmental or risk assessment of spent fuel transportation. It needs to be redone, with peer review during its production and experimental verification of its assumptions, before it has any relevance to the shipments planned to Yucca Mountain. Finally, it must be expanded into a full risk assessment by inputing its radiological release fractions and probabilities into a valid dispersal simulation to properly determine the impact of its results. 51 refs

  14. Dynamic localization of tripartite motif-containing 22 in nuclear and nucleolar bodies

    Energy Technology Data Exchange (ETDEWEB)

    Sivaramakrishnan, Gayathri; Sun, Yang; Tan, Si Kee [School of Biological Sciences, Nanyang Technological University, 60 Nanyang Drive, Singapore 637551 (Singapore); Lin, Valerie C.L., E-mail: cllin@ntu.edu.sg [School of Biological Sciences, Nanyang Technological University, 60 Nanyang Drive, Singapore 637551 (Singapore)

    2009-05-01

    Tripartite motif-containing 22 (TRIM22) exhibits antiviral and growth inhibitory properties, but there has been no study on the localization and dynamics of the endogenous TRIM22 protein. We report here that TRIM22 is dramatically induced by progesterone in MDA-MB-231-derived ABC28 cells and T47D cells. This induction was associated with an increase in TRIM22 nuclear bodies (NB), and an even more prominent increase in nucleolar TRIM22 bodies. Distinct endogenous TRIM22 NB were also demonstrated in several other cell lines including MCF7 and HeLa cells. These TRIM22 NB resemble Cajal bodies, co-localized with these structures and co-immunoprecipitated with p80-coilin. However, IFN{gamma}-induced TRIM22 in HeLa and MCF7 cells did not form NB, implying the forms and distribution of TRIM22 are regulated by specific cellular signals. This notion is also supported by the observation that TRIM22 NB undergoes dynamic cell-cycle dependent changes in distribution such that TRIM22 NB started to form in early G0/G1 but became dispersed in the S-phase. In light of its potential antiviral and antitumor properties, the findings here provide an interesting gateway to study the relationship between the different forms and functions of TRIM22.

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

    International Nuclear Information System (INIS)

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

    1989-08-01

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

  16. System of large transport containers for waste from dismantling light water and gas-cooled nuclear reactors. Volume 2

    International Nuclear Information System (INIS)

    Price, M.S.T.

    1986-09-01

    This report deals with the operational, radiological and economic aspects of transport as well as conceptual designs of large containers for the transport of radioactive decommissioning wastes from nuclear power plants within the member states of the European Economic Community. The means of transport, the costs and radiological detriment are considered, and conceptual designs of containers are described. Recommendations are made for further studies. (U.K.)

  17. Delayed behaviour of concrete in nuclear power plant containment: analysis and modelling

    International Nuclear Information System (INIS)

    Granger, L.

    1995-02-01

    The containment of French nuclear power plant of the 1300 and 1400 MWe PWR type are made of prestressed concrete and their delayed behaviour is systematically monitored by a very complete instrumentation. In an accidental phase, the tightness of the 1.2 m thick structure, dimensioned to withstand an internal absolute pressure of 0.5 MPa depends mainly on the residual prestress of concrete. But surveillance devices reveal substantial differences from one site to another, from which the regulation calculation models cannot make satisfactory allowance. For the purpose of improving the management of the population of power stations, EDF in 1992 initiated a large study aimed at predicting the true creep behaviour of the containments already built. This study, more material oriented, includes numerous shrinkage and creep tests on reconstructed concrete in laboratory as well as on cement paste and aggregate. The main results are presented in part one. In the second part, we consider the different delayed strains of concrete one by one. A precise analysis of the physico-chemical phenomena at the origin of the delayed behaviours, leads us to propose a practical modelling of concrete in an overall equivalent continuous material approach. Secondly, the few parameters of the model are determined on the experimental results. In order to do so, two particular finite element programs in CESAR-LCPC have been developed. The first one permits to take into account the non linear diffusion of humidity in concrete as a function of temperature. The diffusion coefficient D(C) (C = water content) is fitted on the loss of weight tests as a function of time. The second step is a creep calculation; first, the program reads back the temperature and humidity results of the previous computations and then calculates the different delayed strains in time. For basic creep, we have chosen a viscoelastic model function of temperature and humidity. The numerical scheme uses the principle of

  18. Investigation of fission product retention by the HTR containment building according to the vented confinement concept. Technical report 1.5

    International Nuclear Information System (INIS)

    Kreidler, R.

    1988-10-01

    The physical behaviour of aerosols during the transient primary gas blowdown phase and their long-time behavior through 10 hours after the beginning of a potential accident were investigated. Three scenarios were considered: pressure relief by way of an open failing safety valve; pressure relief by way of a safety valve after water ingress; severance type break of the pressure balancing pipe in the reactor cell. The deposition of aerosols was found to be decisively influenced by the composition of the atmosphere and the atmosphere mixture (pure air, He atmosphere), resp. in the containment building. (DG) [de

  19. The long-term behavior of glasses for wastes containment purposes; Le comportement a long terme des verres pour le confinement des dechets

    Energy Technology Data Exchange (ETDEWEB)

    Gin, S. [Division de l' energie nucleaire, CEA Centre de Valrho-Marcoule (France)

    2010-07-01

    In the presence of water, nuclear glasses undergo reactions that may be attributed, in part, to the nature of the chemical bonds set up within the glass structure, and - as regards other reactions - owing to the properties of the solute species. The main reactions involved include ion exchanges, these chiefly involving alkali metals, weakly bonded as these are to the glass network, silicon hydrolysis-condensation reactions, resulting in the formation of a porous hydrated layer, gradually taking on a passivating role, along with the precipitation of crystallized secondary phases. At the temperatures of interest in the disposal context (25 - 90 C degrees), such secondary phases mainly involve clay minerals. Fundamental research studies, conducted to gain an understanding of the mechanisms involved, have highlighted the existence of a strong coupling, at the mesoscopic scale, between the aforementioned chemical reactions, and solute transport, the hydrated layer having the ability to take on a passivating role, as its porosity closes. A new model named GRAAL (for Glass Reactivity with Allowance for the Alteration Layer), includes an explicit description of the four chief alteration mechanisms acting on glass: formation of the passivating layer by glass hydration, diffusion of water across that layer, dissolution of that layer over its outside surface, and precipitation of crystallized secondary phases. The equations may either be solved analytically, for simple cases, or be integrated into a geochemical code, to cater for chemistry-transport couplings, and simulate complex systems

  20. Reduction of 68Ge activity containing liquid waste from 68Ga PET chemistry in nuclear medicine and radiopharmacy by solidification

    NARCIS (Netherlands)

    E. de Blois (Erik); H.S. Chan (Ho Sze); K. Roy (Kamalika); E.P. Krenning (Eric); W.A.P. Breeman (Wouter)

    2011-01-01

    textabstractPET with68Ga from the TiO2- or SnO2- based68Ge/68Ga generators is of increasing interest for PET imaging in nuclear medicine. In general, radionuclidic purity (68Ge vs.68Ga activity) of the eluate of these generators varies between 0.01 and 0.001%. Liquid waste containing low amounts

  1. System of large transport containers for waste from dismantling light water and gas-cooled nuclear reactors. Volume 1

    International Nuclear Information System (INIS)

    Price, M.S.T.; Lafontaine, I.

    1985-01-01

    The purpose of this volume is to introduce the main types of nuclear reactor in the European Community (EC), select reference plants for further study, estimate the waste streams from the reference reactors, survey the transport regulations and assess existing containers

  2. Quality of water from the pool, original containers and aluminum drums used for storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Idjakovic, Z.; Milonjic, S.; Cupic, S.

    2001-01-01

    Results of chemical analyses of water from the pool, including original containers and aluminium drums, for storage of spent nuclear fuel of the research reactor RA at the VINCA Institute and a short survey of the water properties from similar pools of other countries are presented in the paper. (author)

  3. Investigation of active interrogation techniques to detect special nuclear material in maritime environments: Boarded search of a cargo container ship

    Energy Technology Data Exchange (ETDEWEB)

    Grogan, Brandon R., E-mail: groganbr@ornl.gov; Henkel, James J.; Johnson, Jeffrey O.; Mihalczo, John T.; Miller, Thomas M.; Patton, Bruce W.

    2013-12-01

    The detonation of a terrorist nuclear weapon in the United States would result in the massive loss of life and grave economic damage. Even if a device was not detonated, its known or suspected presence aboard a cargo container ship in a U.S. port would have major economic and political consequences. One possible means to prevent this threat would be to board a ship at sea and search for the device before it reaches port. The scenario considered here involves a small Coast Guard team with strong intelligence boarding a container ship to search for a nuclear device. Using active interrogation, the team would nonintrusively search a block of shipping containers to locate the fissile material. Potential interrogation source and detector technologies for the team are discussed. The methodology of the scan is presented along with a technique for calculating the required interrogation source strength using computer simulations. MCNPX was used to construct a computer model of a container ship, and several search scenarios were simulated. The results of the simulations are presented in terms of the source strength required for each interrogation scenario. Validation measurements were performed in order to scale these simulation results to expected performance. Interrogations through the short (2.4 m) axis of a standardized shipping container appear to be feasible given the entire range of container loadings tested. Interrogations through several containers at once or a single container through its long (12.2 m) axis do not appear to be viable with a portable interrogation system.

  4. Investigation of active interrogation techniques to detect special nuclear material in maritime environments: Boarded search of a cargo container ship

    International Nuclear Information System (INIS)

    Grogan, Brandon R.; Henkel, James J.; Johnson, Jeffrey O.; Mihalczo, John T.; Miller, Thomas M.; Patton, Bruce W.

    2013-01-01

    The detonation of a terrorist nuclear weapon in the United States would result in the massive loss of life and grave economic damage. Even if a device was not detonated, its known or suspected presence aboard a cargo container ship in a U.S. port would have major economic and political consequences. One possible means to prevent this threat would be to board a ship at sea and search for the device before it reaches port. The scenario considered here involves a small Coast Guard team with strong intelligence boarding a container ship to search for a nuclear device. Using active interrogation, the team would nonintrusively search a block of shipping containers to locate the fissile material. Potential interrogation source and detector technologies for the team are discussed. The methodology of the scan is presented along with a technique for calculating the required interrogation source strength using computer simulations. MCNPX was used to construct a computer model of a container ship, and several search scenarios were simulated. The results of the simulations are presented in terms of the source strength required for each interrogation scenario. Validation measurements were performed in order to scale these simulation results to expected performance. Interrogations through the short (2.4 m) axis of a standardized shipping container appear to be feasible given the entire range of container loadings tested. Interrogations through several containers at once or a single container through its long (12.2 m) axis do not appear to be viable with a portable interrogation system

  5. Multi-Quanta Spin-Locking Nuclear Magnetic Resonance Relaxation Measurements: An Analysis of the Long-Time Dynamical Properties of Ions and Water Molecules Confined within Dense Clay Sediments

    Directory of Open Access Journals (Sweden)

    Patrice Porion

    2017-11-01

    Full Text Available Solid/liquid interfaces are exploited in various industrial applications because confinement strongly modifies the physico-chemical properties of bulk fluids. In that context, investigating the dynamical properties of confined fluids is crucial to identify and better understand the key factors responsible for their behavior and to optimize their structural and dynamical properties. For that purpose, we have developed multi-quanta spin-locking nuclear magnetic resonance relaxometry of quadrupolar nuclei in order to fill the gap between the time-scales accessible by classical procedures (like dielectric relaxation, inelastic and quasi-elastic neutron scattering and obtain otherwise unattainable dynamical information. This work focuses on the use of quadrupolar nuclei (like 2H, 7Li and 133Cs, because quadrupolar isotopes are the most abundant NMR probes in the periodic table. Clay sediments are the confining media selected for this study because they are ubiquitous materials implied in numerous industrial applications (ionic exchange, pollutant absorption, drilling, waste storing, cracking and heterogeneous catalysis.

  6. Experiences in the emptying of waste silos containing solid nuclear waste from graphite- moderated reactors

    International Nuclear Information System (INIS)

    Wall, S.; Schwarz, T.

    2003-01-01

    difference in the two streams lays in the chosen retrieval technology, reflecting the different problems involved. The most significant differences in the retrieval solutions are; - MAC recovery is achieved with a hydraulic Artisan manipulator arm deployed through a series of new access points in the vault roof; - FED recovery utilizes two retrieval units which deploy petal grabs through tubes originally connected to the waste discharge conveyors. The waste is then transported for further treatment including monitoring, packing, lid and grouting fitting, box filling and swabbing of the boxes to eliminate external contamination. This poster session discusses the experiences made with the retrieval and emptying of graphite containing nuclear waste silos

  7. The aftermath of the Fukushima nuclear accident: Measures to contain groundwater contamination

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo, Adrian H., E-mail: adgallardo@geowater.com.au [CONICET (Argentina National Scientific and Technical Research Council), San Luis National University, FCFMyN, Department of Geology, San Luis 5700 (Argentina); Marui, Atsunao [AIST Geological Survey of Japan, Geo-Resources and Environment Institute, Groundwater Research Group, Ibaraki-ken, Tsukuba 305-8567 (Japan)

    2016-03-15

    Several measures are being implemented to control groundwater contamination at the Fukushima Daiichi Nuclear Plant. This paper presents an overview of work undertaken to contain the spread of radionuclides, and to mitigate releases to the ocean via hydrological pathways. As a first response, contaminated water is being held in tanks while awaiting treatment. Limited storage capacity and the risk of leakage make the measure unsustainable in the long term. Thus, an impervious barrier has been combined with a drain system to minimize the discharge of groundwater offshore. Caesium in seawater at the plant port has largely dropped, although some elevated concentrations are occasionally recorded. Moreover, a dissimilar decline of the radioactivity in fish could indicate additional sources of radionuclides intake. An underground frozen shield is also being constructed around the reactors. This structure would reduce inflows to the reactors and limit the interaction between fresh and contaminated waters. Additional strategies include groundwater abstraction and paving of surfaces to lower water levels and further restrict the mobilisation of radionuclides. Technical difficulties and public distrust pose an unprecedented challenge to the site remediation. Nevertheless, the knowledge acquired during the initial work offers opportunities for better planning and more rigorous decisions in the future. - Highlights: • Measures are being undertaken to manage groundwater contamination in Fukushima. • Methods focus on isolating the source and controlling the radionuclides migration. • Wastewater is being temporarily held in tanks for treatment. • Impervious walls inhibit the transport of contaminants toward the ocean. • Paving and pumping further mitigate the dispersion of pollutants by water.

  8. 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

  9. 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

  10. Plasma physics and controlled nuclear fusion research 1994. V. 3. Proceedings of the fifteenth international conference

    International Nuclear Information System (INIS)

    1996-01-01

    This is the third volume of the proceedings of the 15th International Atomic Energy Agency Conference on Plasma Physics and Controlled Nuclear Fusion Research held in Seville, Spain, from 26 September - 1 October 1994. Contained in it are 29 papers on inertial confinement and 46 papers on magnetic confinement. Refs, figs, tabs

  11. Greater-confinement disposal

    International Nuclear Information System (INIS)

    Trevorrow, L.E.; Schubert, J.P.

    1989-01-01

    Greater-confinement disposal (GCD) is a general term for low-level waste (LLW) disposal technologies that employ natural and/or engineered barriers and provide a degree of confinement greater than that of shallow-land burial (SLB) but possibly less than that of a geologic repository. Thus GCD is associated with lower risk/hazard ratios than SLB. Although any number of disposal technologies might satisfy the definition of GCD, eight have been selected for consideration in this discussion. These technologies include: (1) earth-covered tumuli, (2) concrete structures, both above and below grade, (3) deep trenches, (4) augered shafts, (5) rock cavities, (6) abandoned mines, (7) high-integrity containers, and (8) hydrofracture. Each of these technologies employ several operations that are mature,however, some are at more advanced stages of development and demonstration than others. Each is defined and further described by information on design, advantages and disadvantages, special equipment requirements, and characteristic operations such as construction, waste emplacement, and closure

  12. Design analysis of the molten core confinement within the reactor vessel in the case of severe accidents at nuclear power plants equipped with a reactor of the VVER type

    Science.gov (United States)

    Zvonaryov, Yu. A.; Budaev, M. A.; Volchek, A. M.; Gorbaev, V. A.; Zagryazkin, V. N.; Kiselyov, N. P.; Kobzar', V. L.; Konobeev, A. V.; Tsurikov, D. F.

    2013-12-01

    The present paper reports the results of the preliminary design estimate of the behavior of the core melt in vessels of reactors of the VVER-600 and VVER-1300 types (a standard optimized and informative nuclear power unit based on VVER technology—VVER TOI) in the case of beyond-design-basis severe accidents. The basic processes determining the state of the core melt in the reactor vessel are analyzed. The concept of molten core confinement within the vessel based on the idea of outside cooling is discussed. Basic assumptions and models, as well as the results of calculation of the interaction between molten materials of the core and the wall of the reactor vessel performed by means of the SOCRAT severe accident code, are presented and discussed. On the basis of the data obtained, the requirements on the operation of the safety systems are determined, upon the fulfillment of which there will appear potential prerequisites for implementing the concept of the confinement of the core melt within the reactor in cases of severe accidents at nuclear power plants equipped with VVER reactors.

  13. Nuclear fuel element containing strips of an alloyed Zr, Ti, and Ni getter material

    International Nuclear Information System (INIS)

    Grossman, L.N.; Packard, D.R.

    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 strips and preferably the strips are positioned inside a helical member in the plenum. The position of the alloy strips permits gases and liquids entering the plenum to contact and react with the alloy strips. (U.S.)

  14. Advanced containment research for the Canadian Nuclear Fuel Waste Management Program

    International Nuclear Information System (INIS)

    Onofrei, M.; Mathew, P.M.; McKay, P.; Hosaluk, L.J.; Oscarson, D.W.

    1986-09-01

    This document outlines the program on the development of advanced containment systems for the disposal of used fuel in a vault deep in plutonic rock. Possible advanced containment concepts, the strategy adopted in selecting potential container materials, and experimental programs currently underway or planned are presented. Most effort is currently directed toward developing long-term containment systems based on non-metallic materials and massive metal containers. The use of additional independent barriers to extend the lifetime of simple containment systems is also being evaluated. 58 refs

  15. 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)

  16. PREFACE: Water in confined geometries

    Science.gov (United States)

    Rovere, Mauro

    2004-11-01

    particularly important to understand whether the glass transition temperature could be experimentally accessible for confined water. In this respect the modifications induced by the confinement on the dynamics of water on supercooling are of extreme interest and a number of experimental and computer simulation studies have been devoted in recent years to this topic. This special section contains papers from different groups which have contributed with various experimental and computer simulation techniques to the progress made in the study of water in confined geometry. I thank all of the authors for their stimulating contributions. I am very pleased in particular that Sow-Hsin Chen agreed to contribute since he has done pioneering experimental work on the dynamical properties of confined water upon supercooling, and he is still very active in the field. The work presented by the group of J Swenson concerns also the glass transition of confined water. The Messina group (Crupi et al) is very active in the study of dynamical properties of confined water and they present their results on water in zeolites. From the experimental side there is also a contribution from J Dore's group, one of the first to perform neutron scattering studies on confined water. The work of J Klein looks at the mobility of water molecules confined in subnanometre films. Important contributions on the computer simulation side come from the Geiger group (Brovchenko et al). They performed very accurate simulations of water in nanopores, exploring a large portion of the phase space. Puibasset et al were able to build a very realistic model to simulate water inside Vycor. Zangi et al review the extensive work performed on confined water. Jedlovszky is an expert on the model potential for water and studied how the hydrogen bond network of water can be modified by the presence of an interface. The special issue is intended to stimulate interest and future work on this important subject.

  17. A process and container for transport and storage of irradiated internal equipment from a nuclear reactor

    International Nuclear Information System (INIS)

    Dubourg, M.

    1996-01-01

    A transfer container is assembled (from at least two components) around the highly irradiated internal equipment, in the reactor pool; the entire internal equipment is then transported, as a whole, in the container to a storage zone. The storage container walls are fitted with biological protection and cooling means

  18. Development, production technology and assembly of full-pressure steel containment for WWER 1000 nuclear power plant

    International Nuclear Information System (INIS)

    Pales, I.; Csekey, L.

    1984-01-01

    Described and represented are two alternatives of the project of a full-pressure steel containment and its use in a double full-pressure containment of a WWER-1000 nuclear power plant. The first alternative assumes a containment in the shape of a cylindrical vessel with 40 mm wall thickness, with a semispherical cupola and elliptic bottom with wall thickness of 30 mm. On the basis of the results of an expertise of strength calculation and specialized analyses of the technological aspects of the design the second alternative was made of a containment of the same shape without reinforcement with wall thickness of 40 mm and a 20 mm semispherical cupola. Also discussed is the problem of its manufacture, namely the welding and forming of parts, strength and stability calculations and assembly problems. (B.S.)

  19. A Prototype Scintillating Fibre Tracker for the Cosmic-Ray Muon Tomography of Legacy Nuclear Waste Containers

    OpenAIRE

    Kaiser, R.; Clarkson, A.; Hamilton, D. J.; Hoek, M.; Ireland, D.G.; Johnstone, J.R.; Keri, T.; Lumsden, S.; Mahon, D. F.; McKinnon, B.; Murray, M.; Nutbeam-Tuffs, S.; Shearer, C.; Staines, C.; Yang, G.

    2014-01-01

    Cosmic-ray muons are highly-penetrative charged particles observed at sea level with a flux of approximately 1 cm−2 min−1. They interact with matter primarily through Coulomb scattering which can be exploited in muon tomography to image objects within industrial nuclear waste containers. This paper presents the prototype scintillating-fibre detector developed for this application at the University of Glasgow. Experimental results taken with test objects are shown in comparison to results from...

  20. 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

  1. Commissioning of the steel containment and its related components of the Loviisa II. nuclear power plant

    International Nuclear Information System (INIS)

    Tuominen, J.; Pietikaeinen, L.; Kutramoinen, H.

    1982-01-01

    The outer concrete wall of the containment building serves as a protective system for the components in side. It contains the hermetically sealed steel pressure vessel for retaining the release of radioactive contamination in an accident situation. During a loss-of-coolant accident the pressure is reduced in two steps. The various testing procedures of the containment locks, their main-tenance and repair, the pressure and tightness tests of the steel containment and the preliminary operational tests of the other components of the containment system has been presented. (R.P.)

  2. Report by the AERES on the unit: Unit of researches on neutron transport and radioactivity confinement in nuclear installations under the supervision of the establishments and bodies: IRSN; Rapport de l'AERES sur l'unite: Unite de recherches en neutronique et confinement de la radioactivite dans les installations nucleaires sous tutelle des etablissements et organismes: IRSN

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-07-15

    This report is a kind of audit report on a research laboratory whose activity is organized in three departments: neutron transport and criticality (themes: numerical methods, maths and statistics related to the simulation of neutral particle propagation, nuclear data, uncertainty propagation and bias estimation, code qualification and associated experimental programs, neutron transport in reactors and fuel cycle, criticality accidents), radionuclide transfer in radioactive waste disposals (site identification strategy, hydro-mechanical phenomena affecting storage performance, physical-chemical evolution factors, storage modelling), and metrology and confinement of radioactive gases and aerosols. The authors discuss an assessment of the unit activities in terms of strengths and opportunities, aspects to be improved and recommendations, productions and publications. A more detailed assessment is presented for each department in terms of scientific quality, influence and attractiveness (awards, recruitment capacity, capacity to obtain financing and to tender, participation to international programs), strategy and governance, and project

  3. Design response spectra-compliant real and synthetic GMS for seismic analysis of seismically isolated nuclear reactor containment building

    Energy Technology Data Exchange (ETDEWEB)

    Ali, Ahmer [ENVICO Consultants Co. Ltd., Seoul (Korea, Republic of); Abu-Hayah, Nadin; Kim, Doo Kie [Civil and Environmental Engineering, Kunsan National University, Gunsan (Korea, Republic of); Cho, Sung Gook [Innose Tech Co., Ltd., Incheon (Korea, Republic of)

    2017-06-15

    Due to the severe impacts of recent earthquakes, the use of seismic isolation is paramount for the safety of nuclear structures. The diversity observed in seismic events demands ongoing research to analyze the devastating attributes involved, and hence to enhance the sustainability of base-isolated nuclear power plants. This study reports the seismic performance of a seismically-isolated nuclear reactor containment building (NRCB) under strong short-period ground motions (SPGMs) and long-period ground motions (LPGMs). The United States Nuclear Regulatory Commission-based design response spectrum for the seismic design of nuclear power plants is stipulated as the reference spectrum for ground motion selection. Within the period range(s) of interest, the spectral matching of selected records with the target spectrum is ensured using the spectral-compatibility approach. NRC-compliant SPGMs and LPGMs from the mega-thrust Tohoku earthquake are used to obtain the structural response of the base-isolated NRCB. To account for the lack of earthquakes in low-to-moderate seismicity zones and the gap in the artificial synthesis of long-period records, wavelet-decomposition based autoregressive moving average modeling for artificial generation of real ground motions is performed. Based on analysis results from real and simulated SPGMs versus LPGMs, the performance of NRCBs is discussed with suggestions for future research and seismic provisions.

  4. Multipurpose containers for the transport of nuclear material: The example of transport flask CF6

    International Nuclear Information System (INIS)

    Gualdrini, G.F.; Borgia, M.G.

    1989-03-01

    The present paper summarizes the design and licensing activity carried out in the frame work of an ENEA working group which was set up with the aim of developing transport flasks for radioactive and non radioactive dangerous materials. In particular the nuclear design of the multipurpose transport flask CF6 is described. The paper was presented at the seminar on 'Nuclear wastes and transport of radioactive materials' held in Bologna on June 4th and 5th 1987 under the aegis of the Department of Physics of the University of Bologna. (author)

  5. Proceedings of the seventh symposium on containment of underground nuclear explosions. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Olsen, C.W. [ed.

    1993-12-31

    This is Volume 1 of two unclassified volumes of a meeting of workers at all levels in the science and technology of containment. Papers on containment and related geological, geophysical, engineering, chemical, and computational topics were included. Particular topics included in this volume are: General containment,tunnel and LOS topics, cavity conditions, and LYNER and chemical kiloton. Individual papers are indexed separately on the data base.

  6. Surveillance Report on SAVY-4000 and Hagan Nuclear Material Storage Containers for FY 2017

    Energy Technology Data Exchange (ETDEWEB)

    Reeves, Kirk Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Karns, Tristan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Weis, Eric [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Oka, Jude M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Smith, Paul Herrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stone, Timothy Amos [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Narlesky, Joshua Edward [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-12-14

    In accordance with the SAVY-4000 Surveillance Plan [1] and DOE M441.1-1 requirements, storage container surveillance continued through fiscal year 2017 at Los Alamos National Laboratory. Surveillance items for the year consisted of 8 SAVY-4000 storage containers, 8 Hagan containers, and 39 SAVY-4000 transfer containers. The SAVY-4000 surveillance items ranged in age from 1 year to 5.6 years and the Hagan containers ranged in age from 6.3 years to 17.6 years. The surveillance containers for this year were selected primarily to better understand the extent of corrosion of the stainless steel components of the containers. Accelerated aging studies indicate that the O-ring and filter components of the SAVY-4000 will last at least 40 years under LANL storage conditions. However, the observation of corrosion on the inside of SAVY-4000 and Hagan surveillance containers has shifted the emphasis to understanding both the nature and the extent of corrosion on the stainless steel body. The restriction on handling soluble residues greater than 500 grams continued this year, delaying the surveillance of some items that was scheduled in earlier surveillance plans.

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

    International Nuclear Information System (INIS)

    1993-07-01

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

  8. Preliminary technique assessment for nondestructive evaluation certification of the NNWSI [Nevada Nuclear Waste Storage Investigations] disposal container closure

    International Nuclear Information System (INIS)

    Day, R.A.

    1988-01-01

    Under the direction of the Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) program, the Nevada Nuclear Waste Storage Investigations (NNWSI) project is evaluating a candidate repository site at Yucca Mountain, Nevada, for permanent disposal of high-level nuclear waste. The Lawrence Livermore National Laboratory (LLNL), a participant in the NNWSI project, is developing waste package designs to meet the NRC requirements. One aspect of this waste package is the nondestructive testing of the final closure of the waste container. The container closure weld can best be nondestructively examined (NDE) by a combination of ultrasonics and liquid penetrants. This combination can be applied remotely and can meet stringent quality control requirements common to nuclear applications. Further development in remote systems and inspection will be required to meet anticipated requirements for flaw detection reliability and sensitivity. New research is not required but might reduce cost or inspection time. Ultrasonic and liquid penetrant methods can examine all closure methods currently being considered, which include fusion welding and inertial welding, among others. These NDE methods also have a history of application in high radiation environments and a well developed technology base for remote operation that can be used to reduce development and design costs. 43 refs., 23 figs., 3 tabs

  9. Short-term pressure and temperature MSLB response analyses for large dry containment of the Maanshan nuclear power station

    Energy Technology Data Exchange (ETDEWEB)

    Dai, Liang-Che, E-mail: lcdai@iner.gov.tw; Chen, Yen-Shu; Yuann, Yng-Ruey

    2014-12-15

    Highlights: • The GOTHIC code is used for the PWR dry containment pressure and temperature analysis. • Boundary conditions are hot standby and 102% power main steam line break accidents. • Containment pressure and temperature responses of GOTHIC are similar with FSAR. • The capability of the developed model to perform licensing calculation is assessed. - Abstract: Units 1 and 2 of the Maanshan nuclear power station are the typical Westinghouse three-loop PWR (pressurized water reactor) with large dry containments. In this study, the containment analysis program GOTHIC is adopted for the dry containment pressure and temperature analysis. Free air space and sump of the PWR dry containment are individually modeled as control volumes. The containment spray system and fan cooler unit are also considered in the GOTHIC model. The blowdown mass and energy data of the main steam line break (hot standby condition and various reactor thermal power levels) are tabulated in the Maanshan Final Safety Analysis Report (FSAR) 6.2 which could be used as the boundary conditions for the containment model. The calculated containment pressure and temperature behaviors of the selected cases are in good agreement with the FSAR results. In this study, hot standby and 102% reactor thermal power main steam line break accidents are selected. The calculated peak containment pressure is 323.50 kPag (46.92 psig) for hot standby MSLB, which is a little higher than the FSAR value of 311.92 kPag (45.24 psig). But it is still below the design value of 413.69 kPag (60 psig). The calculated peak vapor temperature inside the containment is 187.0 °C (368.59 F) for 102% reactor thermal power MSLB, which is lower than the FSAR result of 194.42 °C (381.95 F). The effects of the containment spray system and fan cooler units could be clearly observed in the GOTHIC analysis. The calculated containment pressure and temperature behaviors of the selected cases are in good agreement with the FSAR

  10. Short-term pressure and temperature MSLB response analyses for large dry containment of the Maanshan nuclear power station

    International Nuclear Information System (INIS)

    Dai, Liang-Che; Chen, Yen-Shu; Yuann, Yng-Ruey

    2014-01-01

    Highlights: • The GOTHIC code is used for the PWR dry containment pressure and temperature analysis. • Boundary conditions are hot standby and 102% power main steam line break accidents. • Containment pressure and temperature responses of GOTHIC are similar with FSAR. • The capability of the developed model to perform licensing calculation is assessed. - Abstract: Units 1 and 2 of the Maanshan nuclear power station are the typical Westinghouse three-loop PWR (pressurized water reactor) with large dry containments. In this study, the containment analysis program GOTHIC is adopted for the dry containment pressure and temperature analysis. Free air space and sump of the PWR dry containment are individually modeled as control volumes. The containment spray system and fan cooler unit are also considered in the GOTHIC model. The blowdown mass and energy data of the main steam line break (hot standby condition and various reactor thermal power levels) are tabulated in the Maanshan Final Safety Analysis Report (FSAR) 6.2 which could be used as the boundary conditions for the containment model. The calculated containment pressure and temperature behaviors of the selected cases are in good agreement with the FSAR results. In this study, hot standby and 102% reactor thermal power main steam line break accidents are selected. The calculated peak containment pressure is 323.50 kPag (46.92 psig) for hot standby MSLB, which is a little higher than the FSAR value of 311.92 kPag (45.24 psig). But it is still below the design value of 413.69 kPag (60 psig). The calculated peak vapor temperature inside the containment is 187.0 °C (368.59 F) for 102% reactor thermal power MSLB, which is lower than the FSAR result of 194.42 °C (381.95 F). The effects of the containment spray system and fan cooler units could be clearly observed in the GOTHIC analysis. The calculated containment pressure and temperature behaviors of the selected cases are in good agreement with the FSAR

  11. Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tan, Lizhen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yang, Ying [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sridharan, K. [Univ. of Wisconsin, Madison, WI (United States)

    2015-12-01

    The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe2+) irradiation.

  12. On the implications of confinement

    International Nuclear Information System (INIS)

    Roberts, C.D.

    1992-01-01

    In this paper, the authors consider some implications of confinement starting from the basic observation that cross-sections for the production of colored asymptotic states, such as free quarks and gluons, from color singlet initial states must be zero if QCD is to be confining. The authors discuss two pictures of confinement: the failure of the cluster decomposition property and the absence of a pole at timelike momenta in the propagator of a confined particle. The authors use QCD-based models as a framework to relate the failure of the cluster decomposition property to other ideas, such as the role of a nonzero gluon condensate. The authors' primary interest is to address the question of the absence of a mass pole through a study of model Schwinger-Dyson equations. These equations contain some of the dynamical information that is present in the study of the cluster decomposition property. The authors discuss the problems within this idea and its study using the Schwinger-Dyson equations

  13. Failure modes of a concrete nuclear-containment building subjected to hydrogen detonation

    International Nuclear Information System (INIS)

    Fugelso, L.E.; Butler, T.A.

    1983-01-01

    Calculated response for the Indian Point reactor containment building to static internal pressure and one case of a dynamic pressure representing hydrogen combustion and detonation are presented. Comparison of the potential failure modes is made. 9 figures

  14. Tokamak confinement scaling laws

    International Nuclear Information System (INIS)

    Connor, J.

    1998-01-01

    The scaling of energy confinement with engineering parameters, such as plasma current and major radius, is important for establishing the size of an ignited fusion device. Tokamaks exhibit a variety of modes of operation with different confinement properties. At present there is no adequate first principles theory to predict tokamak energy confinement and the empirical scaling method is the preferred approach to designing next step tokamaks. This paper reviews a number of robust theoretical concepts, such as dimensional analysis and stability boundaries, which provide a framework for characterising and understanding tokamak confinement and, therefore, generate more confidence in using empirical laws for extrapolation to future devices. (author)

  15. Proceedings of the seventh symposium on containment of underground nuclear explosions. Volume 2

    International Nuclear Information System (INIS)

    Olsen, C.W.

    1993-01-01

    This is Volume 2 of two unclassified volumes of a meeting of workers at all levels in the science and technology of containment. Papers on containment and related geological, geophysical, engineering, chemical, and computational topics were included. Particular topics in this volume include: Low-yield test beds, modeling and residual stress, material properties, collapse phenomena and shock diagnostics, stemming practices and performance, geophysics, and geosciences and weapons destruction. Individual papers are indexed separately on the data base

  16. Beyond design considerations for bonded and un-bonded tendons in nuclear containments

    International Nuclear Information System (INIS)

    Varpasuo, P.

    2015-01-01

    The simplest method to assess the ultimate pressure capacity of the post-tensioned concrete containment vessel (PCCV) is to use analytical formulas. This approach results in pressure-displacement, which gives the mid-height radial displacement in the cylindrical part of the containment plotted against the internal pressure. The curve exhibits five significant corner points, namely: pressure to overcome pre-stress, concrete cracking, deformation of the lining, plastic deformation of the reinforcing bar, tendon yield (ultimate limit state) [ru

  17. Proceedings of the seventh symposium on containment of underground nuclear explosions. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Olsen, C.W. [ed.

    1993-12-31

    This is Volume 2 of two unclassified volumes of a meeting of workers at all levels in the science and technology of containment. Papers on containment and related geological, geophysical, engineering, chemical, and computational topics were included. Particular topics in this volume include: Low-yield test beds, modeling and residual stress, material properties, collapse phenomena and shock diagnostics, stemming practices and performance, geophysics, and geosciences and weapons destruction. Individual papers are indexed separately on the data base.

  18. A methodology to analyze the creep behaviour of nuclear fuel waste containers

    International Nuclear Information System (INIS)

    Dutton, R.

    1995-12-01

    The concept for the disposal of used-fuel waste from CANDU reeactors operating in Canada comprises a system of natural and engineered barriers surrounding the waste in a mined vault situated at a depth of 500 - 1000 m in plutonic rock of the Canadian Shield. The fuel would be packaged in a highly durable metal container, within a matrix of compacted particulate. The design of the container takes into account that it would be subjected to an external hydrostatic pressure. Consideration of the rate of radioactive decay of the radionuclides contained in the fuel, suggests that the lifetime of the container should be at least 500 years. Consequently, the role of creep deformation, and the possibility of creep rupture of the container shell, must be included in the assessment of time-dependent mechanical integrity. This report describes an analytical approach that can be used to quantify the long-term creep properties of the container material and facilitate the engineering design. The overall objective is to formulate a constitutive creep equation that provides the required input for a finite element computer model being developed to analyze the elastic-plastic behaviour of the container. Alternative forms of such equations are reviewed. It is shown that the capability of many of these equations to extrapolate over long time scales is limited by their empirical nature. Thus, the recommended equation is based on current mechanistic understanding of creep deformation and creep rupture. A criterion for determining the onset of material failure by creep rupture, that could be used in the design of containers with extended structural integrity, is proposed. Interpretation and extrapolation will be supported by the complementary Deformation and Fracture Mechanism Maps. (author) 103 refs., 2 tabs., 54 figs

  19. Multilayer Protective Coatings for High-Level Nuclear Waste Storage Containers

    Science.gov (United States)

    Fusco, Michael

    Corrosion-based failures of high-level nuclear waste (HLW) storage containers are potentially hazardous due to a possible release of radionuclides through cracks in the canister due to corrosion, especially for above-ground storage (i.e. dry casks). Protective coatings have been proposed to combat these premature failures, which include stress-corrosion cracking and hydrogen-diffusion cracking, among others. The coatings are to be deposited in multiple thin layers as thin films on the outer surface of the stainless steel waste basket canister. Coating materials include: TiN, ZrO2, TiO2, Al 2O3, and MoS2, which together may provide increased resistances to corrosion and mechanical wear, as well as act as a barrier to hydrogen diffusion. The focus of this research is on the corrosion resistance and characterization of single layer coatings to determine the possible benefit from the use of the proposed coating materials. Experimental methods involve electrochemical polarization, both DC and AC techniques, and corrosion in circulating salt brines of varying pH. DC polarization allows for estimation of corrosion rates, passivation behavior, and a qualitative survey of localized corrosion, whereas AC electrochemistry has the benefit of revealing information about kinetics and interfacial reactions that is not obtainable using DC techniques. Circulation in salt brines for nearly 150 days revealed sustained adhesion of the coatings and minimal weight change of the steel samples. One-inch diameter steel coupons composed of stainless steel types 304 and 316 and A36 low alloy carbon steel were coated with single layers using magnetron sputtering with compound targets in an inert argon atmosphere. This resulted in very thin films for the metal-oxides based on low sputter rates. DC polarization showed that corrosion rates were very similar between bare and coated stainless steel samples, whereas a statistically significant decrease in uniform corrosion was measured on coated

  20. Polarographic determination of Iodide and Iodate, in Solutions Coming from Aerosols in Fission Products Containment Studies in Nuclear Power Stations

    International Nuclear Information System (INIS)

    Sanchez, M.; Ballesteros, O.; Fernandez, M.; Clavero, M.A.; Gonzalez, A.M.

    2000-01-01

    A polarographic method is described for the iodine species determination, iodide and iodate in water solutions. the iodate can be determined by differential pulse polarography. Calibration curves and the detection and determination limits have been obtained. Iodides is oxidized to iodate with sodium hypochlorite and the excess of oxidizing agent is destroyed with sodium sulphide. The concentration of iodide is calculated as the difference between the concentration of iodate in the sample before and after the oxidation. As an application, species of iodine in samples coming from the experimental plants GIRS (Gaseous Iodine Removal by Sprays) of Nuclear Fission Department of the CIEMAT, dedicated to fission products containment studies in nuclear power station, were determined. (Author) 10 refs

  1. The potential for stress corrosion cracking of copper containers in a Canadian nuclear fuel waste disposal vault

    International Nuclear Information System (INIS)

    King, F.

    1996-09-01

    The potential for stress corrosion cracking (SCC) of copper nuclear fuel waste containers in a conceptual Canadian disposal vault has been assessed through a review of the literature and comparison of those environmental factors that cause SCC with the expected disposal environment. Stress-corrosion cracking appears to be an unlikely failure mode for Cu containers in a Canadian disposal vault because of a combination of environmental factors. Most importantly, there is only a relatively short period during which the containers will be undergoing strain when cracking should be possible at all, and then cracking is not expected because of the absence of known SCC agents, such as NH 3 , NO 2 - or organic acids. In addition, other environmental factors will mitigate SCC, namely, the presence of C1 - and its effect on film properties and the limited supply of oxidants. These arguments, to greater or lesser extent, apply to the three major mechanisms proposed for SCC of Cu alloys in aqueous solutions: film-rupture/anodic dissolution, tarnish rupture and film-induced cleavage. Detailed reviews of the SCC literature are presented as Appendices. The literature on the SCC of Cu (>99 wt.% Cu) is reviewed, including studies carried out in a number of countries under nuclear waste disposal conditions. Because of similarities with the behaviour of Cu, the more extensive literature on the SCC of α-brass in ammonia solutions is also reviewed. (author). 140 refs., 3 tabs., 25 figs

  2. High polymer composites for containers for the long-term storage of spent nuclear fuel and high level radioactive waste

    International Nuclear Information System (INIS)

    Bonin, H.W.; Vui, V.T.; Legault, J.-F.

    1997-01-01

    The feasibility of using polymeric composite materials as an alternative to metals in the design of a nuclear waste disposal container was examined. The disposal containers would be stored in deep underground vaults in plutonic rock formations within the Canadian Shield for several thousands of years. The conditions of disposal considered in the evaluation of the polymeric composite materials were based on the long-term disposal concept proposed by Atomic Energy of Canada Limited. Four different composites were considered for this work, all based on boron fibre as reinforcing material, imbedded in polymeric matrices made of polystyrene (PS), polymethyl methacrylate (PMMA), Devcon 10210 epoxy, and polyetheretherketone (PEEK). Both PS and PMMA were determined as unsuitable for use in the fabrication of the storage container because of thermal failure. This was determined following thermal analysis of the materials in which heat transfer calculations yielded the temperature of the container wall and of the surroundings resulting from the heat generated by the spent nuclear fuel stored inside the container. In the case of the PS, the temperature of the container, the buffer and the backfill would exceed the 100 degrees C imposed in the AECL's proposal as the maximum allowable. In the case of the PMMA, the 100 degrees temperature is too close to the glass transition temperature of this material (105 degrees C) and would cause structural degradation of the container wall. The other two materials present acceptable thermal characteristics for this application. An important concern for polymeric materials in such use is their resistance to radiations. The Devcon 10210 epoxy has been the object of research at the Royal Military College in the past years and fair, but limited, resistance to both neutrons and gamma radiation has been demonstrated, with the evidence of increased mechanical strength when subjected to moderate doses. Provided that the container wall could be

  3. Pressure behavior in nuclear reactor containment following a loss of coolant accident

    International Nuclear Information System (INIS)

    Khattab, M.; Ibrahim, N.A.; Bedrose, C.D.

    1995-01-01

    The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The seniors of small, medium and large LOCA at 2%, 15%, and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs

  4. Integrity assessment of nuclear containment vessel steels due to missile impact load

    International Nuclear Information System (INIS)

    Miyamoto, H.

    1984-01-01

    The impact behavior of the steel plate of BWR containment vessels against missiles, that are caused by the postulated catastrophic failure of components with a high kinematic energy has been investigated. The research started in 1977 and was completed in 1979 by a project team of the Technical Research Association for Integrity of Structures at Elevated Service Temperatures, sponsored by three Electric Power Companies (Tokyo, Chubu, and Chugoku). Although the probability of the occurrence of missiles inside and outside of containment vessels is extremely low, the following items are required to maintain the integrity of containment vessels. (1) The probability of the occurrence of missiles. (2) The weight and energy of the missiles. (3) The impact behavior of containment vessel steel plate against postulated missiles. In connection with the above item (3), an actual-scale missile test was conducted. In addition, a computation analysis was performed to confirm the impact behavior against the missiles, in order to search for wide applicability to the various kinds of postulated missiles. This research has finally aimed at obtaining a new empirical formula which carries out the assessment of the integrity of containment vessels. (orig.)

  5. A four-scale homogenization analysis of creep of a nuclear containment structure

    Energy Technology Data Exchange (ETDEWEB)

    Tran, A.B. [Université Paris-Est, Laboratoire Modélisation et Simulation Multi Échelle, MSME UMR 8208 CNRS, 5 bd Descartes, F-77454 Marne-la-Vallée (France); EDF R and D – Département MMC Site des Renardières – Avenue des Renardières - Ecuelles, 77818 Moret sur Loing Cedex (France); Department of Applied Informatics in Construction, National University of Civil Engineering, 55 Giai Phong Road, Hai Ba Trung District, Hanoi (Viet Nam); Yvonnet, J., E-mail: julien.yvonnet@univ-paris-est.fr [Université Paris-Est, Laboratoire Modélisation et Simulation Multi Échelle, MSME UMR 8208 CNRS, 5 bd Descartes, F-77454 Marne-la-Vallée (France); He, Q.-C. [Université Paris-Est, Laboratoire Modélisation et Simulation Multi Échelle, MSME UMR 8208 CNRS, 5 bd Descartes, F-77454 Marne-la-Vallée (France); Toulemonde, C.; Sanahuja, J. [EDF R and D – Département MMC Site des Renardières – Avenue des Renardières - Ecuelles, 77818 Moret sur Loing Cedex (France)

    2013-12-15

    A four-scale approach is proposed to predict the creep behavior of a concrete structure. The behavior of concrete is modeled through a numerical multiscale methodology, by successively homogenizing the viscoelastic behavior at different scales, starting from the cement paste. The homogenization is carried out by numerically constructing an effective relaxation tensor at each scale. In this framework, the impact of modifying the microstructural parameters can be directly observed on the structure response, like the interaction of the creep of concrete with the prestressing tendons network, and the effects of an internal pressure which might occur during a nuclear accident.

  6. Earthquake-proof and anti-vibration device for nuclear containers

    International Nuclear Information System (INIS)

    Hayano, Mutsuhiko; Imayoshi, Sho; Hirota, Koichi.

    1985-01-01

    Purpose: To surely support the reactor vessel by a lower support structure upon earthquakes. Constitution: Low melting alloys are filled to a gap between the lower ring of a reactor container and a retainer and to a gap at the vertical portion between the protrusion of a guard vessel and the lower support structure. The liquid level of the low melting metal in the guard vessel upon melting is set such that it forms a liquid level filling the gap between the lower ring of the reactor container and the retainer even if the reactor vessel is axially shrinked or expanded. By filling the low melting metal in this way into the gap, the temperature-dependency of the defined members is improved and the support for the vessel or container upon earthquake can be made reliable due to the structural material. (Yoshino, Y.)

  7. In situ testing of titanium and mild steel nuclear waste containers at the WIPP site

    International Nuclear Information System (INIS)

    Molecke, M.A.

    1990-01-01

    An overview of the Waste Isolation Pilot Plant (WIPP) in situ tests on the corrosion of titanium and mild steel for high level waste containers is presented. The tests at Sandia have moved out of the laboratory into a test underground facility in order to evaluate the performance of the waste package material. The tests are being performed under both near-reference and accelerated salt repository conditions. Some containers are filled with high level waste glass (non-radioactive); others contain electric heaters. Backfill material is either bentonite/sand or crushed salt. In other tests metals and glasses are exposed directly to brine. The tests are designed to study the corrosion and metallurgy of the canister and overpack materials; the feasibility and performance of backfill materials; and near-field effects such as brine migration

  8. Thermal and mechanical behaviour of an experimental mock-up of a nuclear containment

    International Nuclear Information System (INIS)

    Chauvel, D.; Barre, F.

    2007-01-01

    In order to better understand the behaviour of a reactor containment submitted to combined pressure and temperature loads by means of studies of the concrete permeability and the state of cracking evolution, EDF and its French partners have built a prestressed concrete test model which represents a PWR containment typical section. The monitoring system was designed to follow the evolution of strains, temperature and state of cracking of the concrete wall from construction stage to air and steam tests. The measurements results as well as their comparison with theoretical laws or data and calculated values, allow to determine the main thermal and mechanical characteristics of the concrete, to analyse the thermo-mechanical behaviour of the structure and also to check the design criteria of prestressed concrete containments. (authors)

  9. A Prototype Scintillating-Fibre Tracker for the Cosmic-ray Muon Tomography of Legacy Nuclear Waste Containers

    Directory of Open Access Journals (Sweden)

    Kaiser R.

    2014-03-01

    Full Text Available Cosmic-ray muons are highly-penetrative charged particles observed at sea level with a flux of approximately 1 cm−2 min−1. They interact with matter primarily through Coulomb scattering which can be exploited in muon tomography to image objects within industrial nuclear waste containers. This paper presents the prototype scintillating-fibre detector developed for this application at the University of Glasgow. Experimental results taken with test objects are shown in comparison to results from GEANT4 simulations. These results verify the simulation and show discrimination between the low, medium and high-Z materials imaged.

  10. A Prototype Scintillating-Fibre Tracker for the Cosmic-ray Muon Tomography of Legacy Nuclear Waste Containers

    Science.gov (United States)

    Kaiser, R.; Clarkson, A.; Hamilton, D. J.; Hoek, M.; Ireland, D. G.; Johnston, J. R.; Keri, T.; Lumsden, S.; Mahon, D. F.; McKinnon, B.; Murray, M.; Nutbeam-Tuffs, S.; Shearer, C.; Staines, C.; Yang, G.; Zimmerman, C.

    2014-03-01

    Cosmic-ray muons are highly-penetrative charged particles observed at sea level with a flux of approximately 1 cm-2 min-1. They interact with matter primarily through Coulomb scattering which can be exploited in muon tomography to image objects within industrial nuclear waste containers. This paper presents the prototype scintillating-fibre detector developed for this application at the University of Glasgow. Experimental results taken with test objects are shown in comparison to results from GEANT4 simulations. These results verify the simulation and show discrimination between the low, medium and high-Z materials imaged.

  11. Compatibility of Space Nuclear Power Plant Materials in an Inert He/Xe Working Gas Containing Reactive Impurities

    International Nuclear Information System (INIS)

    MM Hall

    2006-01-01

    A major materials selection and qualification issue identified in the Space Materials Plan is the potential for creating materials compatibility problems by combining dissimilar reactor core, Brayton Unit and other power conversion plant materials in a recirculating, inert He/Xe gas loop containing reactive impurity gases. Reported here are results of equilibrium thermochemical analyses that address the compatibility of space nuclear power plant (SNPP) materials in high temperature impure He gas environments. These studies provide early information regarding the constraints that exist for SNPP materials selection and provide guidance for establishing test objectives and environments for SNPP materials qualification testing

  12. Compatibility of Space Nuclear Power Plant Materials in an Inert He/Xe Working Gas Containing Reactive Impurities

    Energy Technology Data Exchange (ETDEWEB)

    MM Hall

    2006-01-31

    A major materials selection and qualification issue identified in the Space Materials Plan is the potential for creating materials compatibility problems by combining dissimilar reactor core, Brayton Unit and other power conversion plant materials in a recirculating, inert He/Xe gas loop containing reactive impurity gases. Reported here are results of equilibrium thermochemical analyses that address the compatibility of space nuclear power plant (SNPP) materials in high temperature impure He gas environments. These studies provide early information regarding the constraints that exist for SNPP materials selection and provide guidance for establishing test objectives and environments for SNPP materials qualification testing.

  13. Containment And Siting Of Nuclear Power Plants. Proceedings Of The Symposium

    International Nuclear Information System (INIS)

    1967-01-01

    Proceedings of a Symposium organized by the IAEA and held in Vienna, 3-7 April 1967. The meeting was attended by 246 participants from 29 countries and 6 international organizations. Contents: National practices in reactor siting (14 papers); Reactor siting considerations (12 papers); Containment for particular reactor types (6 papers); Containment for reactor types - general considerations (6 papers); Release and transport of pollutants (10 papers); Panel discussion. Each paper is in its original language (35 English, 9 French and 4 Spanish) and is preceded by an abstract in English and one in the original language if this is not English. Discussions are in English. (author)

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

    Science.gov (United States)

    Travelli, A.

    1985-10-25

    A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

  15. Dynamics in geometrical confinement

    CERN Document Server

    Kremer, Friedrich

    2014-01-01

    This book describes the dynamics of low molecular weight and polymeric molecules when they are constrained under conditions of geometrical confinement. It covers geometrical confinement in different dimensionalities: (i) in nanometer thin layers or self supporting films (1-dimensional confinement) (ii) in pores or tubes with nanometric diameters (2-dimensional confinement) (iii) as micelles embedded in matrices (3-dimensional) or as nanodroplets.The dynamics under such conditions have been a much discussed and central topic in the focus of intense worldwide research activities within the last two decades. The present book discusses how the resulting molecular mobility is influenced by the subtle counterbalance between surface effects (typically slowing down molecular dynamics through attractive guest/host interactions) and confinement effects (typically increasing the mobility). It also explains how these influences can be modified and tuned, e.g. through appropriate surface coatings, film thicknesses or pore...

  16. Elastic membranes in confinement.

    Science.gov (United States)

    Bostwick, J B; Miksis, M J; Davis, S H

    2016-07-01

    An elastic membrane stretched between two walls takes a shape defined by its length and the volume of fluid it encloses. Many biological structures, such as cells, mitochondria and coiled DNA, have fine internal structure in which a membrane (or elastic member) is geometrically 'confined' by another object. Here, the two-dimensional shape of an elastic membrane in a 'confining' box is studied by introducing a repulsive confinement pressure that prevents the membrane from intersecting the wall. The stage is set by contrasting confined and unconfined solutions. Continuation methods are then used to compute response diagrams, from which we identify the particular membrane mechanics that generate mitochondria-like shapes. Large confinement pressures yield complex response diagrams with secondary bifurcations and multiple turning points where modal identities may change. Regions in parameter space where such behaviour occurs are then mapped. © 2016 The Author(s).

  17. The dynamic response of the containment of the Qinshan nuclear power plant to the aircraft impact loading

    International Nuclear Information System (INIS)

    Zuo Jiahong; Han Liangbi; Xia Zufeng

    1991-08-01

    The structural response of the containment of the Qinshan Nuclear Power Plant under the standard-load-function to aircraft impact has been analyzed by using the ADINA code considering an axisymmetric continuum model, which is assumed a mixed-model for the steel-concrete mixture. It consists of 179 four-node isoparametric concrete elements and 118 steel elements. In order to obtain optimum results, the nonlinear behavior of materials and structures, dynamic modes of failure and damage have been considered in the numerical solution. The coordinate system is based on the total Lagrangian formulation. The F.E. system has been solved using an incremental interactions (BFGS method) with 600 steps totally. A discussion of the overall behavior of the containment for the aircraft impact loading, especially the nonlinear behavior of the local impacted area is presented

  18. Gravitationally confined relativistic neutrinos

    Science.gov (United States)

    Vayenas, C. G.; Fokas, A. S.; Grigoriou, D.

    2017-09-01

    Combining special relativity, the equivalence principle, and Newton’s universal gravitational law with gravitational rather than rest masses, one finds that gravitational interactions between relativistic neutrinos with kinetic energies above 50 MeV are very strong and can lead to the formation of gravitationally confined composite structures with the mass and other properties of hadrons. One may model such structures by considering three neutrinos moving symmetrically on a circular orbit under the influence of their gravitational attraction, and by assuming quantization of their angular momentum, as in the Bohr model of the H atom. The model contains no adjustable parameters and its solution, using a neutrino rest mass of 0.05 eV/c2, leads to composite state radii close to 1 fm and composite state masses close to 1 GeV/c2. Similar models of relativistic rotating electron - neutrino pairs give a mass of 81 GeV/c2, close to that of W bosons. This novel mechanism of generating mass suggests that the Higgs mass generation mechanism can be modeled as a latent gravitational field which gets activated by relativistic neutrinos.

  19. Advances in HYDRA and its applications to simulations of inertial confinement fusion targets

    Directory of Open Access Journals (Sweden)

    Marinak M.M.

    2013-11-01

    Full Text Available A new set of capabilities has been implemented in the HYDRA 2D/3D multiphysics inertial confinement fusion simulation code. These include a Monte Carlo particle transport library. It models transport of neutrons, gamma rays and light ions, as well as products they generate from nuclear and coulomb collisions. It allows accurate simulations of nuclear diagnostic signatures from capsule implosions. We apply it to here in a 3D simulation of a National Ignition Facility (NIF ignition capsule which models the full capsule solid angle. This simulation contains a severely rough ablator perturbation and provides diagnostics signatures of capsule failure due to excessive instability growth.

  20. Evaluation of nitrogen containing reducing agents for the corrosion control of materials relevant to nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Padma S. [Water and Steam Chemistry Division, BARC Facilities, Kalpakkam, Tamilnadu (India); Mohan, D. [Department of Chemistry, Anna University, Chennai, Tamilnadu (India); Chandran, Sinu; Rajesh, Puspalata; Rangarajan, S. [Water and Steam Chemistry Division, BARC Facilities, Kalpakkam, Tamilnadu (India); Velmurugan, S., E-mail: svelu@igcar.gov.in [Water and Steam Chemistry Division, BARC Facilities, Kalpakkam, Tamilnadu (India)

    2017-02-01

    Materials undergo enhanced corrosion in the presence of oxidants in aqueous media. Usually, hydrogen gas or water soluble reducing agents are used for inhibiting corrosion. In the present study, the feasibility of using alternate reducing agents such as hydrazine, aqueous ammonia, and hydroxylamine that can stay in the liquid phase was investigated. A comparative study of corrosion behavior of the structural materials of the nuclear reactor viz. carbon steel (CS), stainless steel (SS-304 LN), monel-400 and incoloy-800 in the oxidizing and reducing conditions was also made. In nuclear industry, the presence of radiation field adds to the corrosion problems. The radiolysis products of water such as oxygen and hydrogen peroxide create an oxidizing environment that enhances the corrosion. Electrochemical studies at 90 °C showed that the reducing agents investigated were efficient in controlling corrosion processes in the presence of oxygen and hydrogen peroxide. Evaluation of thermal stability of hydrazine and its effect on corrosion potential of SS-304 LN were also investigated in the temperature range of 200–280 °C. The results showed that the thermal decomposition of hydrazine followed a first order kinetics. Besides, a change in electrochemical corrosion potential (ECP) was observed from −0.4 V (Vs SHE) to −0.67 V (Vs SHE) on addition of 5 ppm of hydrazine at 240 °C. Investigations were also made to understand the distribution behavior of hydrogen peroxide and hydrazine in water-steam phases and it was found that both the phases showed identical behavior. - Highlights: • Hydrazine was found to be a promising reducing agent for oxidant control. • In presence of hydrazine corrosion potential of SS304 LN was well below −230 mV. • SS304LN could be protected from IGSCC by hydrazine addition. • Thermal and radiation stability of hydrazine at 285 °C was found satisfactory.

  1. 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

  2. Analysis of the radioisotopes contained in nuclear waste drums by a tomography assay system

    International Nuclear Information System (INIS)

    Eifler, P.; Kroth, K.; Odoj, R.

    1995-01-01

    The disposal of radioactive wastes into a repository requires a detailed specification of the radioisotopic inventory of the waste containers. This paper describes an advanced tomography assay system capable of identifying and localizing all detectable radioactive nuclides contained in Low-Level-Waste-drums (LLW) and quantitatively determining their activities. The assay system uses a combined transmission and emission tomography technique. Whereas transmission computerized tomography is applied to examine the contents of the drums in respect of inhomogeneous fillings (density distribution), emission computerized tomography is used to inspect the radionuclide inventory (activity distribution). The paper describes in detail the tomography assay system including its major components. Measurements on LLW-drums as well as reconstructed images of the density and activity distribution of the drums' contents are a special subject of the paper

  3. The multi-canister overpack -- Hanford`s N Reactor spent nuclear fuel container

    Energy Technology Data Exchange (ETDEWEB)

    Goldmann, L.H.

    1998-05-03

    The Hanford Site has developed an integrated process strategy for the irradiated fuel, including sorting and cleaning SNF in the K basins, loading the SNF into multi-canister overpacks, drying the fuel at the Cold Vacuum Drying Facility, and transporting the dried fuel to the Canister Storage Building for validation, testing and finally interim storage. This presentation provides a description of the MCO and an overview of the proposed use of the MCO as a container for spent fuel.

  4. Nuclear criticality safety for warehousing of 55-gal drums containing highly enriched uranium metal

    International Nuclear Information System (INIS)

    Robinson, R.C.; Dodds, H.L.

    1989-01-01

    Subcritical data for six separate arrays of 55-gal 17H steel shipping/storage containers is calculated and used to determine safe storage limits for U(97.5)-metal in a warehouse environment at the Oak Ridge Y-12 Plant. Results are presented for three different forms of fissile material: 97.5 wt% 235 U enriched metal (ρ 0 = 18.76 g/cm 3 ) cylinders, metal spheres, and low density (ρ = 0.25 ρ 0 ) metal cylinders

  5. Nuclear Receptor-Mediated Alleviation of Alcoholic Fatty Liver by Polyphenols Contained in Alcoholic Beverages

    OpenAIRE

    Yao, Ruiqing; Yasuoka, Akihito; Kamei, Asuka; Ushiama, Shota; Kitagawa, Yoshinori; Rogi, Tomohiro; Shibata, Hiroshi; Abe, Keiko; Misaka, Takumi

    2014-01-01

    To elucidate the effect of the polyphenols contained in alcoholic beverages on the metabolic stress induced by ethanol consumption, four groups of mice were fed for five weeks on Lieber's diet with or without ethanol, with ethanol plus ellagic acid, and with ethanol plus trans-resveratrol. Alcoholic fatty liver was observed in the group fed the ethanol diet but not in those fed the ethanol plus polyphenol diets. Liver transcriptome analysis revealed that the addition of the polyphenols suppre...

  6. Design of the containment structure in prestressed concrete for the Embalse-Cordoba Nuclear Power Plant

    International Nuclear Information System (INIS)

    Godoy, A.R.; Marinelli, C.A.; Gruenbaum, C.E.

    1978-01-01

    The design of a typical prestressed concrete containment structure for a 600 MW Candu - PHW Reactor, presently under construction at Embalse - Cordoba, Argentina is briefly described. The structural behaviour , adcpted prestressing system and tendon pattern are described. Afterwards the evaluation of the prestressing forces as well as the losses assessment and the prestressing sequence are discussed. Finally, some conclusions are drawn in the light of the experience gained at different stages of the construction. (Author)

  7. Prediction of hydrogen concentration in nuclear power plant containment under severe accidents using cascaded fuzzy neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Geon Pil; Kim, Dong Yeong; Yoo, Kwae Hwan; Na, Man Gyun, E-mail: magyna@chosun.ac.kr

    2016-04-15

    Highlights: • We present a hydrogen-concentration prediction method in an NPP containment. • The cascaded fuzzy neural network (CFNN) is used in this prediction model. • The CFNN model is much better than the existing FNN model. • This prediction can help prevent severe accidents in NPP due to hydrogen explosion. - Abstract: Recently, severe accidents in nuclear power plants (NPPs) have attracted worldwide interest since the Fukushima accident. If the hydrogen concentration in an NPP containment is increased above 4% in atmospheric pressure, hydrogen combustion will likely occur. Therefore, the hydrogen concentration must be kept below 4%. This study presents the prediction of hydrogen concentration using cascaded fuzzy neural network (CFNN). The CFNN model repeatedly applies FNN modules that are serially connected. The CFNN model was developed using data on severe accidents in NPPs. The data were obtained by numerically simulating the accident scenarios using the MAAP4 code for optimized power reactor 1000 (OPR1000) because real severe accident data cannot be obtained from actual NPP accidents. The root-mean-square error level predicted by the CFNN model is below approximately 5%. It was confirmed that the CFNN model could accurately predict the hydrogen concentration in the containment. If NPP operators can predict the hydrogen concentration in the containment using the CFNN model, this prediction can assist them in preventing a hydrogen explosion.

  8. An approach for evaluating the general and localised corrosion of carbon steel containers for nuclear waste disposal

    International Nuclear Information System (INIS)

    Marsh, G.P.; Taylor, K.J.; Sharland, S.M.; Tasker, P.W.

    1987-06-01

    The paper considers the long term corrosion of carbon steel containers for heat generating nuclear waste in a granitic repository. Under such conditions carbon steel may exhibit general, localised or passive corrosion behaviour depending on the exact composition and redox potential of the groundwater contacting the containers; localised corrosion being of most concern because it has the fastest propagation rate. It is well established, however, that such localised corrosion is only possible when the environment is sufficiently oxidising to maintain a positive potential gradient between the cathodic surface and the corrosion sites, which requires that species which oxidising potentials greater than water need to be present. This fact provides a basis for estimating the periods during which containers may be subject to localised and subsequently to general corrosion, and hence for making an overall assessment of the metal allowance required for a specified container life. A model for the diffusion transport of oxygen has been developed, and a sensitivity analysis has shown that the period of possible attack is strongly dependent on the passive film leakage current, the radiation dose rate and the oxygen diffusion coefficient. (orig.)

  9. Approach for evaluating the general and localized corrosion of carbon-steel containers for nuclear waste disposal

    International Nuclear Information System (INIS)

    Marsh, G.P.; Taylor, K.J.; Sharland, S.M.; Tasker, P.W.

    1987-01-01

    The paper considers the long term corrosion of carbon-steel containers for heat generating nuclear waste in a granitic repository. Under such conditions carbon steel may exhibit general, localized or passive corrosion behavior depending on the exact composition and redox potential of the groundwater contacting the containers; localized corrosion being of most concern because it has the fastest propagation rate. It is well established, however, that such localized corrosion is only possible when the environment is sufficiently oxidizing to maintain a positive potential gradient between the cathodic surface and the corrosion sites, which requires that species with oxidizing potentials greater than water need to be present. This fact provides a basis for estimating the periods during which containers may be subject to localized and subsequently to general corrosion, and hence for making an overall assessment of the metal allowance required for a specified container life. A model for the diffusion transport of oxygen has been developed, and a sensitivity analysis has shown that the period of possible localized attack is strongly dependent on the passive film leakage current, the radiation dose rate and the oxygen diffusion coefficient. 20 references, 5 figures

  10. Nuclear Magnetic Resonance and Elastic Wave Velocity of Chalk Saturated with Brines Containing Divalent Ions

    DEFF Research Database (Denmark)

    Katika, Konstantina; Alam, Mohammad Monzurul; Fabricius, Ida Lykke

    divided into groups of three and each group was saturated either with deionized water, calcite equilibrated water, or sodium chloride, magnesium chloride and calcium chloride solutions of the same ionic strength. Saturation with solutions that contain divalent ions caused major shifts in the distribution...... of the relaxation time. Core samples saturated with calcium chloride solution relaxed slower and those saturated with magnesium chloride solution relaxed faster than the rest of the samples. Along with the changes in relaxation the samples experienced smaller velocities of elastic waves when saturated with MgCl2...

  11. Assessment of the Biodegradability of Containers for Low and Intermediate Level Nuclear Waste

    International Nuclear Information System (INIS)

    Zlobenko, B.P.

    2013-01-01

    Concrete and reinforced concrete are widely used as engineered barriers (containers) for radioactive waste disposal facilities due to their isolating ability, mechanical stability and low cost. Several types of protective reinforced concrete containers for low and intermediate level waste have been designed in Ukraine. Evaluation of these containers for microbial stability is required according to NRC of Ukraine Regulation No.306.608-96. The research was therefore aimed at studying the degradation of the cement material due to microbiological interaction and the possibility of biodegraded cement as an ideal environment for the growth of other microorganisms under waste disposal conditions to satisfy the regulatory requirements. Results from this study indicated that Aspergillus niger induced gluconic and oxalic acids that dissolve portlandite (with a low leaching of calcium) after one year of contact time. This resulted in an increase in porosity, loss in tensile strength biomechanically deteriorated and cracking. XRD analysis identified crystalline precipitates within the biomass on the concrete surface as calcium oxalate dehydrate (weddellite) and calcium oxalate monohydrate (whewellite). The mechanism regarding of the microbiological interaction on the concrete surface can be summarized as follows: Phase 1: Fungi accumulate on the surface of the concrete, thereby degrading the concrete surface by biochemical and biomechanical interactions. When this effect is in the presence of air with available carbon dioxide, the micro fungi reduces the pH of the concrete from >13 to 8.5. During this phase no accumulation were observed in sections where granite aggregates are present. Phase 2: After reducing the pH of the concrete paste during phase 1, and provided that sufficient nutrients, moisture and oxygen are present sulphur oxidizing bacteria start to accumulate on the concrete surface. The result form this study therefore concluded that fungal biogeochemical activity

  12. The confinement problem

    International Nuclear Information System (INIS)

    Seiler, E.

    1985-01-01

    Confinement of quarks is sometimes taken as some kind of dogma in the contemporary theory of strong interactions - quantum chromo-dynamics (QCD). Scientists should not be content with that. What is meant by ''permanent confinement'' should be formulated more precisely to see whether the theory has this property or not. The author looks at some possible interpretations of ''confinement'' and their shortcomings and then turns to the most widely used rather pragmatic definition based on the somewhat unphysical notion of infinitely heavy external sources. He describes what is known about the problem and tries to bring into focus some aspects that are insufficiently understood in his opinion

  13. Nuclear Criticality Safety of the DOT 9975 Container for237NpO2Storage, Handling, and Transport

    Energy Technology Data Exchange (ETDEWEB)

    Reed, D.A.

    2003-08-29

    Nuclear criticality safety considerations are presented to address use of the DOT 9975 shipping container for {sup 237}NpO{sub 2}. The DOT 9975 container will be used by multiple DOE sites and contractors. Various of site- and activity-specific NCS and facility safety documents are yet to be developed. For these reasons, an overall assessment of criticality safety of {sup 237}NpO{sub 2}-loaded DOT 9975 containers is considered useful to personnel involved in generating, reviewing, or approving these various documents. It is concluded that inherent container features, the loading per container (maximum of 6 kg {sup 237}Np), and the nuclear physics properties of {sup 237}NpO{sub 2} combine to preclude the potential for a nuclear criticality accident. This conclusion applies to storage, handling, and transport operations involving closed DOT 9975 packages, including credible off-normal conditions that may result in damage to packages during those operations.

  14. Crevice Corrosion Behavior of Candidate Nuclear Waste Container Materials in Repository Environment Paper Number 02529

    International Nuclear Information System (INIS)

    Hua, F.; Sarver, J.; Mohn, W.

    2001-01-01

    Alloy 22 (UNS N06022) and Ti Grade 7 (UNS R52400) have been proposed as the corrosion resistant materials for fabricating the waste package outer barrier and the drip shield, respectively for the proposed nuclear waste repository Yucca Mountain Project. In this work, the susceptibility of welded and annealed Alloy 22 (N06022) and Ti Grade 7 (UNS R52400) to crevice corrosion was studied by the Multiple Crevice Assembly (ASTM G78) method combined with surface morphological observation after four and eight weeks of exposure to the Basic Saturated Water (BSW-12) in a temperature range from 60 to 105 C. The susceptibility of the materials to crevice corrosion was evaluated based on the appearance of crevice attack underneath the crevice formers and the weight loss data. The results showed that, after exposed to BSW-12 for four and eight weeks, no obvious crevice attack was observed on these materials. The descaled weight loss increased with the increase in temperature for all materials. The weight loss, however, is believed to be caused by general corrosion, rather than crevice corrosion. There was no significant difference between the annealed and welded materials either. On the other hand, to conclude that these materials are immune to crevice corrosion in BSW-12 will require longer term testing

  15. Crevice Corrosion Behavior of Candidate Nuclear Waste Container Materials in Repository Environment

    Energy Technology Data Exchange (ETDEWEB)

    F. Hua; J. Sarver; W. Mohn

    2001-11-08

    Alloy 22 (UNS N06022) and Ti Grade 7 (UNS R52400) have been proposed as the corrosion resistant materials for fabricating the waste package outer barrier and the drip shield, respectively for the proposed nuclear waste repository Yucca Mountain Project. In this work, the susceptibility of welded and annealed Alloy 22 (N06022) and Ti Grade 7 (UNS R52400) to crevice corrosion was studied by the Multiple Crevice Assembly (ASTM G78) method combined with surface morphological observation after four and eight weeks of exposure to the Basic Saturated Water (BSW-12) in a temperature range from 60 to 105 C. The susceptibility of the materials to crevice corrosion was evaluated based on the appearance of crevice attack underneath the crevice formers and the weight loss data. The results showed that, after exposed to BSW-12 for four and eight weeks, no obvious crevice attack was observed on these materials. The descaled weight loss increased with the increase in temperature for all materials. The weight loss, however, is believed to be caused by general corrosion, rather than crevice corrosion. There was no significant difference between the annealed and welded materials either. On the other hand, to conclude that these materials are immune to crevice corrosion in BSW-12 will require longer term testing.

  16. Nuclear magnetic resonance studies on bacterial dihydrofolate reductase containing (methyl-/sup 13/C)methionine

    Energy Technology Data Exchange (ETDEWEB)

    Blakley, R.L.; Cocco, L.; London, R.E.; Walker, T.E.; Matwiyoff, N.A.

    1978-06-13

    (methyl-/sup 13/C)Methionine has been incorporated with high efficiency by Streptococcus faecium var. Durans strain A into dihydrofolate reductase isoenzyme 2. In the /sup 13/C NMR spectrum of the purified enzyme the resonances corresponding to the seven methionine residues are partially resolved into three composite peaks. Denaturation with urea collapses these into a single peak centered at 15.32 ppm, whereas the resonance of free methionine is at 15.04 ppm. Spectra of the free enzyme, its complex with methotrexate, and its complex with methotrexate and reduced nicotinamide adenine dinucleotide phosphate (NADPH) have been simulated, permitting more accurate estimates of line widths and nuclear Overhauser enhancement (NOE) values. These, together with the T/sub 1/ values, cannot be explained solely by the effects of macromolecular tumbling and very rapid rotation of the methionine methyl group about its axis. A model assuming, in addition, the occurrence of free rotation about the methionine CH/sub 2/-S bond is also unsatisfactory, and it is concluded that internal rotation about the CH/sub 2/-S bond is highly restricted so that the methyl group oscillates through a relatively narrow angular range. Complex formation with NADPH produced rather small changes in the spectrum of the native enzyme, probably due to conformational transitions in the enzyme. However, NADP/sup +/ produced several changes,including movement of one resonance downfield by at least 1.7 ppM.

  17. Closure development for high-level nuclear waste containers for the tuff repository

    International Nuclear Information System (INIS)

    Robitz, E.S. Jr.; McAninch, M.D. Jr.; Edmonds, D.P.; Babcock and Wilcox Co., Alliance, OH

    1990-09-01

    This report summarizes Phase 1 activities for closure development of the high-level nuclear waste package task for the tuff repository. Work was conducted under U.S. Department of Energy (DOE) Contract 9172105, administered through the Lawrence Livermore National Laboratory (LLNL), as part of the Yucca Mountain Project (YMP), funded through the DOE Office of Civilian Radioactive Waste Management (OCRWM). The goal of this phase was to select five closure processes for further evaluation in later phases of the program. A decision tree methodology was utilized to perform an objective evaluation of 15 potential closure processes. Information was gathered via a literature survey, industrial contacts, and discussions with project team members, other experts in the field, and the LLNL waste package task staff. The five processes selected were friction welding, electron beam welding, laser beam welding, gas tungsten arc welding, and plasma arc welding. These are felt to represent the best combination of weldment material properties and process performance in a remote, radioactive environment. Conceptual designs have been generated for these processes to illustrate how they would be implemented in practice. Homopolar resistance welding was included in the Phase 1 analysis, and developments in this process will be monitored via literature in Phases 2 and 3. Work was conducted in accordance with the YMP Quality Assurance Program. 223 refs., 20 figs., 9 tabs

  18. Platinoids and molybdenum in nuclear waste containment glasses: a structural study

    International Nuclear Information System (INIS)

    Le Grand, M.

    2000-01-01

    This work deals with the structure of borosilicate nuclear glasses and with some relationships between structure and macroscopic properties. Two types of elements which may disturb the industrial process - platinoids (Ru and Pd) and molybdenum - are central to this work. Platinoids induce weak modifications on the structure of the glass, causing a depolymerization of the glassy network, an increase of the [3] B/ [4] B ratio and a modification of the medium range order around Si between 3.3 and 4.5 angstrom. The modifications of viscosity and density induced by platinoids in the glass are not due to the structural effect of the platinoids. The increase of viscosity is attributed to needle shaped RuO 2 . It can be moderated by imposing reducing conditions during the elaboration of the glass. The slight difference between experimental and calculated densities is due to the increase of the volume percentage of bubbles in the glass with increasing platinoid content. Mo is either present in the glass as molybdic groupings, or mobilized in chemically complex molybdic crystalline phases. The chemical composition and mineralogy of these phases has been obtained using electronic microprobe data and XRD with Rietveld analysis. The distribution of the different elements between the crystalline phases and the glass is strongly influenced by the structural role of the various cations in the glass. The Mo present in the glass appears as MoO 4 tetrahedra, independent of the borosilicate network. The formation of the crystalline phases can be explained by the existence of a precursor in which the MoO 4 tetrahedra are concentrated in rich alkali and earth-alkali bearing areas of the glass. (author)

  19. Synthesis and characterization of novel lanthanide- and actinide-containing titanates and zircono-titanates; relevance to nuclear waste disposal

    International Nuclear Information System (INIS)

    Shoup, S.L.S.

    1995-08-01

    Before experiments using actinide elements are performed, synthetic routes are tested using lanthanides of comparable ionic radii as surrogates. Compound and solid solution formation in several lanthanide-containing titanate and zircono-titanate systems have been established using X-ray diffraction (XRD) analysis, which helped to define interesting and novel experiments, some of which have been performed and are discussed, for selected actinide elements. The aqueous solubilities of several lanthanide- and actinide-containing compounds, representative of the systems studied, were tested in several leachants, including the WIPP open-quotes Aclose quotes brine, following modified Materials Characterization Center procedures (MCC-3). The WIPP open-quotes Aclose quotes brine is a synthetic substitute for that found in nature at the Waste Isolation Pilot Plant (WIPP) in New Mexico. The concentrations of cerium, used as a surrogate for plutonium, leached by the WIPP open-quotes Aclose quotes brine from all the cerium-containing compounds and solid solutions tested were below the Inductively Coupled Plasma (ICP) atomic emission spectrometry limit of detection (10 ppm) established for cerium in this brine. The concentrations of plutonium leached from the two plutonium-containing solid solutions were less than 1 ppm as determined by gross alpha counting and alpha pulse height analysis. Concentrations of strontium leached by the WIPP brine from stable strontium containing titanate compounds, studied as possible immobilizers of both 90 Sr and actinide elements, were also quite low. These compound and solid solution formation investigations and the aqueous solubility studies suggest that the types of titanate and zircono-titanate compounds and solid solutions studied in this work appear to be useful as host matrices for nuclear waste immobilization

  20. Tools for the simulation of cathodic protection systems for diesel containers of a nuclear power plant

    International Nuclear Information System (INIS)

    Bucio V, F. J.; Sainz M, E.; Ramirez S, R.; Celis del Angel, L.; Palacios H, J.

    2008-01-01

    The cathodic protection is the technique by which the harmful effects of the corrosion for metallic storage tanks can be resisted that they are on the ground. This method causes that all the metallic surface behaves like the cathode of an electrochemical cell. For systems of external cathodic protection, sacrifice anodes can be used or printed current, when sacrifice anodes are used, buried galvanic anodes in the ground are used and connected to the tank, and they are used generally in small tanks. With this technique the anodes are degraded leaving the metallic plate intact. The system proposed for the mathematical simulation of external cathodic protection systems, is an application dedicated to the simulation of a system of anodes for the corrosion protection of diesel containers installed on the ground, that allows to analyze the effects of the resulting current given to the motherboard of the container, that is to say, allows to observe the results before any physical installation and with this of reducing the expenses in the acquisition of unnecessary equipment or its defect, to realize an optimization in the equipment to install to obtain better results. The proposed system was divided in units of sequential processing and not like a whole, the previous thing to allow processing the problem in stages, because it consumes great amount of processing resources and storage during the execution of the same. The aplication was programmed in the programming language of called high level Python. Due to the auxiliary characteristics and units on which it counts, having like basic characteristic the being of open source, reason why it does not require of expensive payments to obtain a license. In the development of the system I am used and unit that allows to visualize controls style Windows, although it is counted on the same version that can be used in operating systems based on Linux. (Author)

  1. The confining trailing string

    CERN Document Server

    Kiritsis, E; Nitti, F

    2014-01-01

    We extend the holographic trailing string picture of a heavy quark to the case of a bulk geometry dual to a confining gauge theory. We compute the classical trailing confining string solution for a static as well as a uniformly moving quark. The trailing string is infinitely extended and approaches a confining horizon, situated at a critical value of the radial coordinate, along one of the space-time directions, breaking boundary rotational invariance. We compute the equations for the fluctuations around the classical solutions, which are used to obtain boundary force correlators controlling the Langevin dynamics of the quark. The imaginary part of the correlators has a non-trivial low-frequency limit, which gives rise to a viscous friction coefficient induced by the confining vacuum. The vacuum correlators are used to define finite-temperature dressed Langevin correlators with an appropriate high-frequency behavior.

  2. The confining trailing string

    Energy Technology Data Exchange (ETDEWEB)

    Kiritsis, Elias [APC, Université Paris 7, CNRS/IN2P3, CEA/IRFU, Obs. de Paris, Sorbonne Paris Cité,Bâtiment Condorcet, F-75205, Paris Cedex 13 (UMR du CNRS 7164) (France); Theory Group, Physics Department, CERN,CH-1211, Geneva 23 (Switzerland); Crete Center for Theoretical Physics, Department of Physics, University of Crete,71003 Heraklion (Greece); Mazzanti, Liuba [Institute for Theoretical Physics and Spinoza Institute, Utrecht University,3508 TD Utrecht (Netherlands); Nitti, Francesco [APC, Université Paris 7, CNRS/IN2P3, CEA/IRFU, Obs. de Paris, Sorbonne Paris Cité,Bâtiment Condorcet, F-75205, Paris Cedex 13 (UMR du CNRS 7164) (France)

    2014-02-19

    We extend the holographic trailing string picture of a heavy quark to the case of a bulk geometry dual to a confining gauge theory. We compute the classical trailing confining string solution for a static as well as a uniformly moving quark. The trailing string is infinitely extended and approaches a confining horizon, situated at a critical value of the radial coordinate, along one of the space-time directions, breaking boundary rotational invariance. We compute the equations for the fluctuations around the classical solutions, which are used to obtain boundary force correlators controlling the Langevin dynamics of the quark. The imaginary part of the correlators has a non-trivial low-frequency limit, which gives rise to a viscous friction coefficient induced by the confining vacuum. The vacuum correlators are used to define finite-temperature dressed Langevin correlators with an appropriate high-frequency behavior.

  3. Solitons and confinement

    International Nuclear Information System (INIS)

    Swieca, J.A.

    1976-01-01

    Some aspects of two recent developments in quantum field theory are discussed. First, related with 'extended particles' such as soliton, kink and the 't Hooft monopole. Second, with confinement of particles which are realized in the Schwinger model [pt

  4. Project for qualification of the Kozloduy confinement system

    International Nuclear Information System (INIS)

    Montes Rodriguez, J.L.

    1993-01-01

    One of the projects awarded to Empresarios Agrupados within the Six-Month WANO Programme for Kozloduy NPP, financed through the European Community's PHARE Programme, relates to the first tasks of plant confinement system qualification. Development of this project - the results of which will serve as a reference for other power plants with VVER-440/230 models - aroused considerable interest in the Bulgarian nuclear community as well as in international entities which render assistance to Eastern power plants. In fact, this is one of the few projects in the programme which takes into account hardware-oriented activities to be carried out urgently in the plant. The VVER-440/230 confinement system performs functions parallel to the containment system of Western PWR reactors. However, it differs significantly in its criteria and operation details. These important differences form the basis for a rational and prudent application of the essence of Western codes and standards relating to the qualification of these systems. The criteria which are developed to make this application viable constitute the most challenging and, at the same time, most risky part of this job. The project will be in its final stages when the 18th Annual Meeting of the Spanish Nuclear Society is held; it is therefore likely that this paper will advance the results and conclusions expected. (author)

  5. Ignition and burn in inertially confined magnetized fuel

    International Nuclear Information System (INIS)

    Kirkpatrick, R.C.; Lindemuth, I.R.

    1991-01-01

    At the third International Conference on Emerging Nuclear Energy Systems, we presented computational results which suggested that ''breakeven'' experiments in inertial confinement fusion (ICF) may be possible with existing driver technology. We recently used the ICF simulation code LASNEX to calculate the performance of an idealized magnetized fuel target. The parameter space in which magnetized fuel operates is remote from that of both ''conventional'' ICF and magnetic confinement fusion devices. In particular, the plasma has a very high β and is wall confined, not magnetically confined. The role of the field is to reduce the electron thermal conductivity and to partially trap the DT alphas. The plasma is contained in a pusher which is imploded to compress and adiabatically heat the plasma from an initial condition of preheat and pre-magnetization to the conditions necessary for fusion ignition. The initial density must be quite low by ICF standards in order to insure that the electron thermal conductivity is suppressed and to minimize the generation of radiation from the plasma. Because the energy loss terms are effectively suppressed, the implosion may proceed at a relatively slow rate of about 1 to 3 cm/μs. Also, the need for low density fuel dictates a much larger target, so that magnetized fuel can use drivers with much lower power and power density. Therefore, magnetized fuel allows the use of efficient drivers that are not suitable for laser or particle beam fusion due to insufficient focus or too long pulse length. The ignition and burn of magnetized fuel involves very different dominant physical processes than does ''conventional'' ICF. The fusion time scale becomes comparable to the hydrodynamic time scale, but other processes that limit the burn in unmagnetized fuel are of no consequence. The idealized low gain magnetized fuel target presented here is large and requires a very low implosion velocity. 11 refs

  6. Noble metal catalyzed hydrogen generation from formic acid in nitrite-containing simulated nuclear waste media

    International Nuclear Information System (INIS)

    King, R.B.; Bhattacharyya, N.K.; Wiemers, K.D.

    1994-08-01

    Simulants for the Hanford Waste Vitrification Plant (HWVP) feed containing the major non-radioactive components Al, Cd, Fe, Mn, Nd, Ni, Si, Zr, Na, CO 3 2- , NO 3 -, and NO 2 - were used as media to evaluate the stability of formic acid towards hydrogen evolution by the reaction HCO 2 H → H 2 + CO 2 catalyzed by the noble metals Ru, Rh, and/or Pd found in significant quantities in uranium fission products. Small scale experiments using 40-50 mL of feed simulant in closed glass reactors (250-550 mL total volume) at 80-100 degree C were used to study the effect of nitrite and nitrate ion on the catalytic activities of the noble metals for formic acid decomposition. Reactions were monitored using gas chromatography to analyze the CO 2 , H 2 , NO, and N 2 O in the gas phase as a function of time. Rhodium, which was introduced as soluble RhCl 3 ·3H 2 O, was found to be the most active catalyst for hydrogen generation from formic acid above ∼80 degree C in the presence of nitrite ion in accord with earlier observations. The inherent homogeneous nature of the nitrite-promoted Rh-catalyzed formic acid decomposition is suggested by the approximate pseudo first-order dependence of the hydrogen production rate on Rh concentration. Titration of the typical feed simulants containing carbonate and nitrite with formic acid in the presence of rhodium at the reaction temperature (∼90 degree C) indicates that the nitrite-promoted Rh-catalyzed decomposition of formic acid occurs only after formic acid has reacted with all of the carbonate and nitrite present to form CO 2 and NO/N 2 O, respectively. The catalytic activities of Ru and Pd towards hydrogen generation from formic acid are quite different than those of Rh in that they are inhibited rather than promoted by the presence of nitrite ion

  7. 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 is being constructed at the Savannah River Plant at a cost of $870 million to immobilize the defense high-level radioactive waste. This radioactive waste is being added to borosilicate glass for later disposal in a federal repository. The borosilicate glass is poured into stainless steel canisters for storage. These canisters must be handled remotely because of their high radioactivity, up to 5000 R/h. After the glass has been poured into the canister which will be temporarily sealed, it is transferred to a decontamination cell and decontaminated. The canister is then transferred to the weld cell where a permanent cap is welded into place. The canisters must then be transported from the processing building to a storage vault on the plant until the federal repository is available. A shielded canister transporter (SCT) has been designed and constructed for this purpose. The design of the SCT vehicle allows the safe transport of a highly radioactive canister containing borosilicate glass weighing 2300 kg with a radiation level up to 5000 R/h from one building to another. The design provides shielding for the operator in the cab of the vehicle to be below 0.5 rem/h. The SCT may also be used to load the final shipping cask when the federal repository is ready to receive the canisters

  8. Nuclear Receptor-Mediated Alleviation of Alcoholic Fatty Liver by Polyphenols Contained in Alcoholic Beverages

    Science.gov (United States)

    Yao, Ruiqing; Yasuoka, Akihito; Kamei, Asuka; Ushiama, Shota; Kitagawa, Yoshinori; Rogi, Tomohiro; Shibata, Hiroshi; Abe, Keiko; Misaka, Takumi

    2014-01-01

    To elucidate the effect of the polyphenols contained in alcoholic beverages on the metabolic stress induced by ethanol consumption, four groups of mice were fed for five weeks on Lieber's diet with or without ethanol, with ethanol plus ellagic acid, and with ethanol plus trans-resveratrol. Alcoholic fatty liver was observed in the group fed the ethanol diet but not in those fed the ethanol plus polyphenol diets. Liver transcriptome analysis revealed that the addition of the polyphenols suppressed the expression of the genes related to cell stress that were up-regulated by ethanol alone. Conversely, the polyphenols up-regulated the genes involved in bile acid synthesis, unsaturated fatty acid elongation, and tetrahydrofolate synthesis that were down-regulated by ethanol alone. Because parts of these genes were known to be regulated by the constitutive androstane receptor (CAR), we performed the same experiment in the CAR-deficient mice. As a result, fatty liver was observed not only in the ethanol group but also with the ethanol plus polyphenol groups. In addition, there was no segregation of the gene expression profiles among these groups. These results provide a molecular basis for the prevention of alcohol-induced stress by the polyphenols in alcoholic beverages. PMID:24498295

  9. Nuclear receptor-mediated alleviation of alcoholic fatty liver by polyphenols contained in alcoholic beverages.

    Directory of Open Access Journals (Sweden)

    Ruiqing Yao

    Full Text Available To elucidate the effect of the polyphenols contained in alcoholic beverages on the metabolic stress induced by ethanol consumption, four groups of mice were fed for five weeks on Lieber's diet with or without ethanol, with ethanol plus ellagic acid, and with ethanol plus trans-resveratrol. Alcoholic fatty liver was observed in the group fed the ethanol diet but not in those fed the ethanol plus polyphenol diets. Liver transcriptome analysis revealed that the addition of the polyphenols suppressed the expression of the genes related to cell stress that were up-regulated by ethanol alone. Conversely, the polyphenols up-regulated the genes involved in bile acid synthesis, unsaturated fatty acid elongation, and tetrahydrofolate synthesis that were down-regulated by ethanol alone. Because parts of these genes were known to be regulated by the constitutive androstane receptor (CAR, we performed the same experiment in the CAR-deficient mice. As a result, fatty liver was observed not only in the ethanol group but also with the ethanol plus polyphenol groups. In addition, there was no segregation of the gene expression profiles among these groups. These results provide a molecular basis for the prevention of alcohol-induced stress by the polyphenols in alcoholic beverages.

  10. Genomic origin and nuclear localization of TERRA telomeric repeat-containing RNA: from Darkness to Dawn.

    Science.gov (United States)

    Diman, Aurélie; Decottignies, Anabelle

    2017-12-14

    Long noncoding RNAs, produced from distinct regions of the chromosomes, are emerging as new key players in several important biological processes. The long noncoding RNAs add a new layer of complexity to cellular regulatory pathways, from transcription to cellular trafficking or chromatin remodeling. More than 25 years ago, the discovery of a transcriptional activity at telomeres of protozoa ended the long-lasting belief that telomeres were transcriptionally silent. Since then, progressively accumulating evidences established that production of TElomeric Repeat-containing RNA (TERRA) was a general feature of eukaryotic cells. Whether TERRA molecules always originate from the telomeres or whether they can be transcribed from internal telomeric repeats as well is however still a matter of debate. Whether TERRA transcripts always localize to telomeres and play similar roles in all eukaryotic cells is also unclear. We review the studies on TERRA localization in the cell, its composition and some aspects of its transcriptional regulation to summarize the current knowledge and controversies about the genomic origin of TERRA, with a focus on human and mouse TERRA. © 2017 Federation of European Biochemical Societies.

  11. INLAY method using the cylindrical container for reactor vessel outlet nozzle of PWR for Ohi Nuclear Power Station

    International Nuclear Information System (INIS)

    Ichiki, Shigeki; Iwahashi, Hiroki; Terada, Norio; Nakamura, Yasuo; Yamamoto, Kazuhide

    2010-01-01

    Preventive maintenance techniques against PWSCC found in Alloy 600 welds have been applied progressively, however, in February 2008, a defect of PWSCC had been found in an outlet nozzle to safe-end weld of Reactor vessel of PWR in Ohi Nuclear Power Station Unit 3 during an outage. Then the defect was removed and after its safety was confirmed, the unit was placed in operation. In the next outage, INLAY method using the cylindrical container was applied and repair welding for the removed part was performed. In addition, circumferential grooving to remove Alloy 600 of the inner surface of the outlet nozzle and circumferential welding with PWSCC resistant Alloy 690 were applied. This circumferential grooving and welding techniques were also applied for all outlet nozzles of Ohi Unit 4 as a preventive maintenance. (author)

  12. The future supply of and demand for candidate materials for the fabrication of nuclear fuel waste disposal containers

    International Nuclear Information System (INIS)

    Grover, L.K.

    1990-01-01

    This report summarizes the findings of a literature survey carried out to assess the future world supply of and demand for titanium, copper and lead. These metals are candidate materials for the fabrication of containers for the immobilization and disposal of Canada's nuclear used-fuel waste for a reference Used-fuel Disposal Centre. Such a facility may begin operation by approximately 2020, and continue for about 40 years. The survey shows that the world has abundant supplies of titanium minerals (mostly in the form of ilmenite), which are expected to last up to at least 2110. However, for copper and lead the balance between supply and demand may warrant increased monitoring beyond the year 2000. A number of factors that can influence future supply and demand are discussed in the report

  13. Water hammer phenomena occurring in nuclear power installations while filling horizontal pipe containing saturated steam with liquid

    International Nuclear Information System (INIS)

    Selivanov, Y.F.; Kirillov, P.L.; Yefanov, A.D.

    1995-01-01

    The potentiality of the water hammer occurrence in nuclear reactor loop components has been considered under the conditions of filling a steam-containing pipeline leg involving horizontal and vertical sections with liquid subcooled to the saturation temperature. As a result of free discharging from the tank, the liquid enters the horizontal pipeline. When the liquid slug formation in the pipeline is fulfilled. The pressure drop being occurred in steam flowing along the pipelines causes the liquid slug to move to the pipeline inlet. When the liquid slug decelerates, a water hammer occurs. This mechanism of water hammer occurrence is tested by experiments. The regimes of the occurrence of multiple considerable water hammers were identified

  14. Technical assistance to AECL: electron beam welding of thick-walled copper containers for nuclear fuel waste disposal

    International Nuclear Information System (INIS)

    Maak, P.Y.Y.

    1984-01-01

    This report describes the results of Phase Two of the copper electron beam welding project for the final closure of copper containers for nuclear fuel waste disposal. It has been demonstrated that single pass, electron beam square butt welds (depth of weld penetration > 25 mm) can be made without preheat in both electrolytic tough-pitch copper and oxygen-free copper plates. The present results show that oxygen-free copper exhibits better weldability than the electrolytic tough-pitch copper in terms of weld penetration and vulnerability to weld defects such as gas porosity, erratic metal overflow and blow holes. The results of ultrasonic inspection studies of the welds are also discussed

  15. Water hammer phenomena occurring in nuclear power installations while filling horizontal pipe containing saturated steam with liquid

    Energy Technology Data Exchange (ETDEWEB)

    Selivanov, Y.F.; Kirillov, P.L.; Yefanov, A.D. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    1995-09-01

    The potentiality of the water hammer occurrence in nuclear reactor loop components has been considered under the conditions of filling a steam-containing pipeline leg involving horizontal and vertical sections with liquid subcooled to the saturation temperature. As a result of free discharging from the tank, the liquid enters the horizontal pipeline. When the liquid slug formation in the pipeline is fulfilled. The pressure drop being occurred in steam flowing along the pipelines causes the liquid slug to move to the pipeline inlet. When the liquid slug decelerates, a water hammer occurs. This mechanism of water hammer occurrence is tested by experiments. The regimes of the occurrence of multiple considerable water hammers were identified.

  16. Simulation of KAEVER experiments on aerosol behavior in a nuclear power plant containment at accident conditions with the ASTEC code

    International Nuclear Information System (INIS)

    Kljenak, I.; Mavko, B.

    2006-01-01

    Experiments on aerosol behaviour in saturated and non-saturated atmosphere, which were performed in the KAEVER experimental facility, were simulated with the severe accident computer code ASTEC CPA V1.2. The specific purpose of the work was to assess the capability of the code to model aerosol condensation and deposition in the containment of a light-water-reactor nuclear power plant at severe accident conditions, if the atmosphere saturation conditions are simulated adequately. Five different tests were first simulated with boundary conditions, obtained from the experiments. In all five tests, a non-saturated atmosphere was simulated, although, in four tests, the atmosphere was allegedly saturated. The simulations were repeated with modified boundary conditions, to obtain a saturated atmosphere in all tests. Results of dry and wet aerosol concentrations in the test vessel atmosphere for both sets of simulations are compared to experimental results. (author)

  17. Response analysis of a nuclear containment structure with nonlinear soil-structure interaction under bi-directional ground motion

    Science.gov (United States)

    Kumar, Santosh; Raychowdhury, Prishati; Gundlapalli, Prabhakar

    2015-06-01

    Design of critical facilities such as nuclear power plant requires an accurate and precise evaluation of seismic demands, as any failure of these facilities poses immense threat to the community. Design complexity of these structures reinforces the necessity of a robust 3D modeling and analysis of the structure and the soil-foundation interface. Moreover, it is important to consider the multiple components of ground motion during time history analysis for a realistic simulation. Present study is focused on investigating the seismic response of a nuclear containment structure considering nonlinear Winkler-based approach to model the soil-foundation interface using a distributed array of inelastic springs, dashpots and gap elements. It is observed from this study that the natural period of the structure increases about 10 %, whereas the force demands decreases up to 24 % by considering the soil-structure interaction. Further, it is observed that foundation deformations, such as rotation and sliding are affected by the embedment ratio, indicating an increase of up to 56 % in these responses for a reduction of embedment from 0.5 to 0.05× the width of the footing.

  18. Nuclear analysis of the Chornobyl fuel containing masses with heterogeneous fuel distribution

    International Nuclear Information System (INIS)

    Turski, R. B.

    1998-01-01

    Although significant data has been obtained on the condition and composition of the fuel containing masses (FCM) located in the concrete chambers under the Chernobyl Unit 4 reactor cavity, there is still uncertainty regarding the possible recriticality of this material. The high radiation levels make access extremely difficult, and most of the samples are from the FCM surface regions. There is little information on the interior regions of the FCM, and one cannot assume with confidence that the surface measurements are representative of the interior regions. Therefore, reasonable assumptions on the key parameters such as fuel concentration, the concentrations of impurities and neutron poisons (especially boron), the void fraction of the FCM due to its known porosity, and the degrees of fuel heterogeneity, are necessary to evaluate the possibility of recriticality. The void fraction is important since it introduces the possibility of water moderator being distributed throughout the FCM. Calculations indicate that the addition of 10 to 30 volume percent (v/o) water to the FCM has a significant impact on the calculated reactivity of the FCM. Therefore, water addition must be considered carefully. The other possible moderators are graphite and silicone dioxide. As discussed later in this paper, silicone dioxide moderation does not represent a criticality threat. For graphite, both heterogeneous fuel arrangements and very large volume fractions of graphite are necessary for a graphite moderated system to go critical. Based on the observations and measurements of the FCM compositions, these conditions do not appear creditable for the Chernobyl FCM. Therefore, the focus of the analysis reported in this paper will be on reasonable heterogeneous fuel arrangements and water moderation. The analysis will evaluate a range of fuel and diluent compositions

  19. The immobilisation of nuclear waste materials containing different alkali elements into single-phase NZP based ceramics

    International Nuclear Information System (INIS)

    Pet'kov, V.I.; Orlova, A.I.; Trubach, I.G.; Demarin, T.; Kurazhkovskaya, V.S.

    2002-01-01

    The NZP matrix, which is based on NaZr 2 (PO 4 ) 3 , is a viable candidate for nuclear waste immobilisation. We examined the possibilities of incorporating of alkali elements into the NZP host structure, investigated the conditions of the crystalline solution formation, and determined the regions of the NZP structure compositional stability for a series of complex orthophosphates of titanium or zirconium and alkali elements A m-x A' x M 2-(m-1)/4 (PO 4 ) 3 with m = 1, 3, or 5 and 0 ≤ x ≤ m, where A and A' are mutually different alkali elements and M is Ti or Zr. The phosphates containing Li-Na, Li-K, Li-Rb, Li-Cs, Na-K, Na-Rb, Na-Cs, K-Rb, K-Cs and Rb-Cs pairs were prepared and studied by X-ray powder analysis, IR spectroscopy, simultaneous DTA-TG measurements and electron microprobe analyses. In the systems studied, wide ranges of crystalline phosphate solutions with tailored alkali metal substitutions were formed owing to the large number of sites available for substitution and high degree of flexibility in the NZP structure. It was found that introduction of the less expensive and lighter Ti in the host phase in place of the commonly used Zr permits cheaper ceramics, having in some cases larger alkali element contents, to be obtained. The phases containing alkali metals can be formed, for instance, during phosphate solidification of molten alkali chlorides with radioactive nuclides from the pyroelectrochemical technologies of nuclear fuel recycling

  20. 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

  1. Vitrification of high level nuclear waste inside ambient temperature disposal containers using inductive heating: The SMILE system

    Energy Technology Data Exchange (ETDEWEB)

    Powell, J.; Reich, M.; Barletta, R.

    1996-03-01

    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 ({approximately}1 m{sup 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., {approximately}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.

  2. POTENTIAL FOR STRESS CORROSION CRACKING OF A537 CARBON STEEL NUCLEAR WASTE TANKS CONTAINING HIGHLY CAUSTIC SOLUTIONS

    Energy Technology Data Exchange (ETDEWEB)

    Lam, P.; Stripling, C.; Fisher, D.; Elder, J.

    2010-04-26

    The evaporator recycle streams of nuclear waste tanks may contain waste in a chemistry and temperature regime that exceeds the current corrosion control program, which imposes temperature limits to mitigate caustic stress corrosion cracking (CSCC). A review of the recent service history found that two of these A537 carbon steel tanks were operated in highly concentrated hydroxide solution at high temperature. Visual inspections, experimental testing, and a review of the tank service history have shown that CSCC has occurred in uncooled/un-stress relieved tanks of similar construction. Therefore, it appears that the efficacy of stress relief of welding residual stress is the primary corrosion-limiting mechanism. The objective of this experimental program is to test A537 carbon steel small scale welded U-bend specimens and large welded plates (30.48 x 30.38 x 2.54 cm) in a caustic solution with upper bound chemistry (12 M hydroxide and 1 M each of nitrate, nitrite, and aluminate) and temperature (125 C). These conditions simulate worst-case situations in these nuclear waste tanks. Both as-welded and stress-relieved specimens have been tested. No evidence of stress corrosion cracking was found in the U-bend specimens after 21 days of testing. The large plate test was completed after 12 weeks of immersion in a similar solution at 125 C except that the aluminate concentration was reduced to 0.3 M. Visual inspection of the plate revealed that stress corrosion cracking had not initiated from the machined crack tips in the weld or in the heat affected zone. NDE ultrasonic testing also confirmed subsurface cracking did not occur. Based on these results, it can be concluded that the environmental condition of these tests was unable to develop stress corrosion cracking within the test periods for the small welded U-bends and for the large plates, which were welded with an identical procedure as used in the construction of the actual nuclear waste tanks in the 1960s. The

  3. ICENES '91:Sixth international conference on emerging nuclear energy systems

    International Nuclear Information System (INIS)

    1991-01-01

    This document contains the program and abstracts of the sessions at the Sixth International Conference on Emerging Nuclear Energy Systems held June 16--21, 1991 at Monterey, California. These sessions included: The plenary session, fission session, fission and nonelectric session, poster session 1P; (space propulsion, space nuclear power, electrostatic confined fusion, fusion miscellaneous, inertial confinement fusion, μ-catalyzed fusion, and cold fusion); Advanced fusion session, space nuclear session, poster session 2P, (nuclear reactions/data, isotope separation, direct energy conversion and exotic concepts, fusion-fission hybrids, nuclear desalting, accelerator waste-transmutation, and fusion-based chemical recycling); energy policy session, poster session 3P (energy policy, magnetic fusion reactors, fission reactors, magnetically insulated inertial fusion, and nuclear explosives for power generation); exotic energy storage and conversion session; and exotic energy storage and conversion; review and closing session

  4. Plasma physics and controlled nuclear fusion research 1988. V.2

    International Nuclear Information System (INIS)

    1989-01-01

    Volume 2 of the Proceedings of the Twelfth International Conference on Plasma Physics and Controlled Nuclear Fusion contains papers given in two of the sessions: D and C. Session D contains papers on magnetic confinement theory and modelling tokamaks, and session C the papers on non-tokamak confinement system. Each of these papers and their authors is listed in the Contents; in turn, each paper contains an abstract for more information as to the contents of a specific paper. Refs, figs and tabs

  5. Japanese magnetic confinement fusion research

    International Nuclear Information System (INIS)

    McKenney, B.L.; McGrain, M.; Horton, C.W.; Rutherford, P.H.

    1990-01-01

    This report is the work of six US scientists who surveyed and assessed Japanese research and development in magnetic fusion. All of the panelists are very familiar with Japanese fusion research through their knowledge of the published scientific literature and through personal contacts with Japanese colleagues and with US colleagues who have visited Japanese research facilities. This report concentrates on the period from the early 1980s through June 1989. The technical accomplishments during this period are reviewed, and the Japanese capabilities and outlook for future contributions are assessed. Detailed evaluations are provided in the areas of basic and applied plasma physics, tokamak confinement, alternate confinement approaches, plasma technology, and fusion nuclear technology and materials. With a sustained national commitment, Japan will surpass US and West European capabilities in the early to middle 1990s in several important areas of fusion research and development. For example, it is expected that the planned upgrade of the Japanese JT-60 tokamak will surpass both the US Tokamak Fusion Test Reactor (TFTR) and the Joint European tours (JET) in the usual measures of plasma performance in the 1992 to 1993 timeframe, and will take a clear international lead in large-tokamak research by 1994 to 1995. The Japanese fusion program has the human and technological resources required to build and operate a fusion engineering test reactor without external participation. By the same measure, Japan would be a highly desirable partner in the bilateral undertaking of such a project

  6. Reduction of 68Ge activity containing liquid waste from 68Ga PET chemistry in nuclear medicine and radiopharmacy by solidification.

    Science.gov (United States)

    de Blois, Erik; Chan, Ho Sze; Roy, Kamalika; Krenning, Eric P; Breeman, Wouter A P

    PET with 68 Ga from the TiO 2 - or SnO 2 - based 68 Ge/ 68 Ga generators is of increasing interest for PET imaging in nuclear medicine. In general, radionuclidic purity ( 68 Ge vs. 68 Ga activity) of the eluate of these generators varies between 0.01 and 0.001%. Liquid waste containing low amounts of 68 Ge activity is produced by eluting the 68 Ge/ 68 Ga generators and residues from PET chemistry. Since clearance level of 68 Ge activity in waste may not exceed 10 Bq/g, as stated by European Directive 96/29/EURATOM, our purpose was to reduce 68 Ge activity in solution from >10 kBq/g to waste. Most efficient method to reduce the 68 Ge activity is by sorption of TiO 2 or Fe 2 O 3 and subsequent centrifugation. The required 10 Bq per mL level of 68 Ge activity in waste was reached by Fe 2 O 3 logarithmically, whereas with TiO 2 asymptotically. The procedure with Fe 2 O 3 eliminates ≥90% of the 68 Ge activity per treatment. Eventually, to simplify the processing a recirculation system was used to investigate 68 Ge activity sorption on TiO 2 , Fe 2 O 3 or Zeolite. Zeolite was introduced for its high sorption at low pH, therefore 68 Ge activity containing waste could directly be used without further interventions. 68 Ge activity containing liquid waste at different HCl concentrations (0.05-1.0 M HCl), was recirculated at 1 mL/min. With Zeolite in the recirculation system, 68 Ge activity showed highest sorption.

  7. Formation of Nup98-containing nuclear bodies in HeLa sublines is linked to genomic rearrangements affecting chromosome 11.

    Science.gov (United States)

    Romana, Serge; Radford-Weiss, Isabelle; Lapierre, Jean-Michel; Doye, Valérie; Geoffroy, Marie-Claude

    2016-09-01

    Nup98 is an important component of the nuclear pore complex (NPC) and also a rare but recurrent target for chromosomal translocation in leukaemogenesis. Nup98 contains multiple cohesive Gly-Leu-Phe-Gly (GLFG) repeats that are critical notably for the formation of intranuclear GLFG bodies. Previous studies have reported the existence of GLFG bodies in cells overexpressing exogenous Nup98 or in a HeLa subline (HeLa-C) expressing an unusual elevated amount of endogenous Nup98. Here, we have analysed the presence of Nup98-containing bodies in several human cell lines. We found that HEp-2, another HeLa subline, contains GLFG bodies that are distinct from those identified in HeLa-C. Rapid amplification of cDNA ends (RACE) revealed that HEp-2 cells express additional truncated forms of Nup98 fused to a non-coding region of chromosome 11q22.1. Cytogenetic analyses using FISH and array-CGH further revealed chromosomal rearrangements that were distinct from those observed in leukaemic cells. Indeed, HEp-2 cells feature a massive amplification of juxtaposed NUP98 and 11q22.1 loci on a chromosome marker derived from chromosome 3. Unexpectedly, minor co-amplifications of NUP98 and 11q22.1 loci were also observed in other HeLa sublines, but on rearranged chromosomes 11. Altogether, this study reveals that distinct genomic rearrangements affecting NUP98 are associated with the formation of GLFG bodies in specific HeLa sublines.

  8. ATR confinement leakage determination

    International Nuclear Information System (INIS)

    Kuan, P.; Buescher, B.J.

    1998-01-01

    The air leakage rate from the Advanced Test Reactor (ATR) confinement is an important parameter in estimating hypothesized accidental releases of radiation to the environment. The leakage rate must be determined periodically to assure that the confinement has not degraded with time and such determination is one of the technical safety requirements of ATR operation. This paper reviews the methods of confinement leakage determination and presents an analysis of leakage determination under windy conditions, which can complicate the interpretation of the determined leakage rates. The paper also presents results of analyses of building air exchange under windy conditions. High wind can enhance air exchange and this could increase the release rates of radioisotopes following an accident

  9. Simulations and imaging algorithm development for a cosmic ray muon tomography system for the detection of special nuclear material in transport containers

    International Nuclear Information System (INIS)

    Jewett, C.; Anghel, V.N.P.; Armitage, J.; Boudjemline, K.; Botte, J.; Bryman, D.; Bueno, J.; Charles, E.; Cousins, T.; Didsbury, R.; Erhardt, L.; Erlandson, A.; Gallant, G.; Jason, A.; Jonkmans, G.; Liu, Z.; McCall, M.; Noel, S.; Oakham, F.G.; Ong, D.; Stocki, T.; Thompson, M.; Waller, D.

    2011-01-01

    The Cosmic Ray Inspection and Passive Tomography (CRIPT) collaboration is developing a cosmic ray muon tomography system to identify Special Nuclear Materials (SNM) in cargo containers. In order to gauge the viability of the technique, and to determine the best detector type, GEANT4 was used to simulate the passage of cosmic ray muons through a cargo container. The scattering density estimation (SDE) algorithm was developed and tested with data from these simulations to determine how well it could reconstruct the interior of a container. The simulation results revealed the ability of cosmic ray muon tomography techniques to image spheres of lead-shielded Special Nuclear Materials (SNM), such as uranium or plutonium, in a cargo container, containing a cargo of granite slabs. (author)

  10. Confined Space Evaluation Student Manual, #19613

    Energy Technology Data Exchange (ETDEWEB)

    Wilmot, David Ezekiel [Los Alamos National Laboratory

    2016-08-29

    Many workplaces contain spaces that are considered to be “confined” because their configuration hinders the activities of employees who must enter into, work in, and exit from them. In general, the permit-required confined spaces (PRCSs) Occupational Safety and Health Administration (OSHA) standard requires that Los Alamos National Laboratory (LANL) evaluate the workplace to determine if any spaces are PRCSs. The standard specifies strict procedures for the evaluation and atmospheric testing of a space before and during an entry by workers. The OSHA PRCS standard provides for alternative (less stringent than full-permit) entry procedures in cases where the only hazard in a space is atmospheric and the hazard can be controlled by forced air. At LANL, all confined spaces or potential confined spaces on LANL-owned or -operated property must be identified and evaluated by a confined space evaluator accompanied by a knowledgeable person. This course provides the information needed by confined space evaluators to make judgements about whether a space is a confined space, and if so, whether the space will require a permit for entry.

  11. Nuclear dynamical deformation induced hetero- and euchromatin positioning

    OpenAIRE

    Awazu, Akinori

    2015-01-01

    The contributions of active deformation dynamics in cell nuclei to the intra-nuclear positioning of hetero- and euchromatin are investigated. We analyzed the behaviors of model chains containing two types of regions, one with high and the other with low mobility, confined in a pulsating container. Here, the regions with high and low mobility represent eu- and heterochromatic regions, respectively, and the pulsating container simulates a nucleus exhibiting dynamic deformations. The Brownian dy...

  12. Confinement for More Space

    DEFF Research Database (Denmark)

    Kipnusu, Wycliffe K.; Elsayed, Mohamed; Kossack, Wilhelm

    2015-01-01

    Broadband dielectric spectroscopy and positron annihilation lifetime spectroscopy are employed to study the molecular dynamics and effective free volume of 2-ethyl-1-hexanol (2E1H) in the bulk state and when confined in unidirectional nanopores with average diameters of 4, 6, and 8 nm. Enhanced α...

  13. 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)

  14. Design and installation of high-temperature ultrasonic measuring system and grinder for nuclear fuel containing trans-uranium elements

    International Nuclear Information System (INIS)

    Serizawa, Hiroyuki; Kikuchi, Hironobu; Iwai, Takashi; Arai, Yasuo; Kurosawa, Makoto; Mimura, Hideaki; Abe, Jiro

    2005-07-01

    A high-temperature ultrasonic measuring system had been designed and installed in a glovebox (711-DGB) to study a mechanical property of nuclear fuel containing trans-uranium (TRU) elements. A figuration apparatus for the cylinder-type sample preparation had also been modified and installed in an established glovebox (142-D). The system consists of an ultrasonic probe, a heating furnace, cooling water-circulating system, a cooling air compressor, vacuum system, gas supplying system and control system. An A/D converter board and an pulsar/receiver board for the measurement of wave velocity were installed in a personal computer. The apparatus was modified to install into the glovebox. Some safety functions were supplied to the control system. The shape and size of the sample was revised to minimize the amount of TRU elements for the use of the measurement. The maximum sample temperature is 1500degC. The performance of the installed apparatuses and the glovebox were confirmed through a series of tests. (author)

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

    International Nuclear Information System (INIS)

    Reviron, Nanthilde

    2009-01-01

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

  16. Effect of microbial action on the corrosion potential of austenitic alloy containers for high-level nuclear waste

    International Nuclear Information System (INIS)

    Angell, P.; Dunn, D.S.; Cragnolino, G.A.

    1996-01-01

    The safe disposal of high-level nuclear waste (HLW) entails the ability to ensure the integrity of waste containers for prolonged time periods. It is generally accepted that under certain conditions, microbial action may change local benign environments to those in which localized corrosion can be actively promoted. The use of repassivation potential (E rp ) in relation to the value of the corrosion potential (E corr ) has been proposed as a means of assessing the propensity of a metallic material to localized corrosion. Microbial activity is known to influence E corr however, the precise mechanism is unresolved. Shewanella putrefaciens, a bacteria with many of the characteristics of sulfate-reducing bacteria (SRB), are being grown under controlled conditions on 316L stainless steel (SS) surfaces to understand the relationship between E corr and metabolic activity. It has been observed that the growth of the bacteria under aerobic conditions, without the production of metabolic sulfide, leads to only minor variation in E corr . These changes possibly correlate to the periods of active bacterial growth

  17. In-situ removal and characterisation of uranium-containing particles from sediments surrounding the Fukushima Daiichi Nuclear Power Plant

    Science.gov (United States)

    Martin, P. G.; Griffiths, I.; Jones, C. P.; Stitt, C. A.; Davies-Milner, M.; Mosselmans, J. F. W.; Yamashiki, Y.; Richards, D. A.; Scott, T. B.

    2016-03-01

    Traditional methods to locate and subsequently study radioactive fallout particles have focused heavily on autoradiography coupled with in-situ analytical techniques. Presented here is the application of a Variable Pressure Scanning Electron Microscope with both backscattered electron and energy dispersive spectroscopy detectors, along with a micromanipulator setup and electron-hardening adhesive to isolate and remove individual particles before synchrotron radiation analysis. This system allows for a greater range of new and existing analytical techniques, at increased detail and speed, to be applied to the material. Using this method, it was possible to erform detailed energy dispersive spectroscopy and synchrotron radiation characterisation of material likely ejected from the Fukushima Daiichi Nuclear Power Plant found within a sediment sample collected from the edge of the 30 km exclusion zone. Particulate material sub-micron in maximum dimension examined during this work via energy dispersive spectroscopy was observed to contain uranium at levels between 19.68 and 28.35 weight percent, with the application of synchrotron radiation spectroscopy confirming its presence as a major constituent. With great effort and cost being devoted to the remediation of significant areas of eastern Japan affected by the incident, it is crucial to gain the greatest possible understanding of the nature of this contamination in order to inform the most appropriate clean-up response.

  18. Event classification related to overflow of solvent containing uranium according to the INES scale (International Nuclear and Radiological Event Scale)

    International Nuclear Information System (INIS)

    Dourado, Eneida R.G.; Assis, Juliana T. de; Lage, Ricardo F.; Lopes, Karina B.

    2013-01-01

    This paper aims to frame the event overflow organic solvent rich in uranium, from a decanter of ore beneficiation plant, caused by the fall in the supply of electricity, according to the criteria established by the International Nuclear Event Scale and radiological (INES), facilitating the understanding of the occurrence and communication with the public regarding the radiation safety aspects involved. With the fall of electricity, routine procedures in situations of installation stop were performed, however, due to operational failure, the valve on the transfer line liquor was not closed. Thus, the mixer continued being fed with liquor, that led the consequent leakage of solvent loaded with uranium. It reached the drainage system, and the box of rainwater harvesting of the plant. However, immediately after the detection of the event, corrective actions were initiated and the overflow was contained. Regulatory agencies followed the removal of the solvent and on the results of the analysis of environmental monitoring, found that the event did not provide exposure to workers or any other impact. Therefore, comparing the characteristics of the event and the guidelines proposed by the INES scale, it is concluded that the classification of the event is below scale/level 0, confirming the absence of risk to the local population, workers and the environment

  19. Long-term performance of structures comprising nuclear power plants PART 1: Deterioration assessment of nuclear power station buildings PART 2: Long-term stability and the leak-tightness of reactor containments

    International Nuclear Information System (INIS)

    Pocock, D.C.; Worthington, J.C.; Oberpichler, R.; Van Exel, H.; Beukelmann, D.; Huth, R.; Rose, B.

    1990-01-01

    The objective of this research was to study the long-term performance of structures comprising nuclear power plants. The time period of interest for this study is 140 years (this figure is based on maximum periods of 40 years for operation and 100 years of storage). It was divided in two parts: - the first based on four UK nuclear power plants examine the principle deterioration mechanism of reinforced structure which is chloride ingress and carbonation penetration - the second based on 2 German nuclear power plants examine the long term behaviour of reinforced and prestressed concrete and also the corrosion of steel containments with particular reference on plastic seals and potential risk areas

  20. Assessment and management of ageing of major nuclear power plant components important to safety: Metal components of BWR containment systems

    International Nuclear Information System (INIS)

    2000-10-01

    At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must therefore be effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed toward technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific

  1. Corrosion kinetics of alloy Ni-22Cr-13Mo-3W as structural material in high level nuclear waste containers

    International Nuclear Information System (INIS)

    Rodriguez, Martin A.

    2004-01-01

    Alloy Ni-22Cr-13Mo-3W (also known as C-22) is one of the candidates to fabricate high level nuclear waste containers. These containers are designed to maintain isolation of the waste for a minimum of 10,000 years. In this period, the material must be resistant to corrosion. If the containers were in contact with water, it is assumed that alloy C-22 may undergo three different corrosion mechanisms: general corrosion, localized corrosion and stress corrosion cracking. This thesis discusses only the first two types of degradation. Electrochemical techniques such as amperometry, potentiometry, potentiodynamic polarization and electrochemical impedance spectroscopy (EIS) and non-electrochemical techniques such as microscopic observation, X-ray fluorescence (XRF) and X-ray photoelectron spectroscopy (XPS) were applied to study the corrosion behavior of alloy C-22 in 1 M NaCl, 25 C degrees saturated NaF (approximately 1 M) and 0,5 M NaCl + 0,5 M NaF solutions. Effects of temperature, pH and alloy thermal aging were analyzed. The corrosion rates obtained at 90 C degrees were low ranging from 0.04 μm/year to 0.48 μm /year. They increased with temperature and decreased with solution pH. Most of the impedance measurements showed a simply capacitive behavior. A second high-frequency time constant was detected in some cases. It was attributed to the formation of a nickel oxide and/or hydroxide at potentials near the reversible potential for this reaction. The active/passive transition detected in some potentiodynamic polarization curves was attributed to the same process. The corrosion potential showed an important increase after 24 hours of immersion. This increase in the corrosion potential was associated with an improvement of the passive film. The corrosion potential was always lower than the re-passivation potential for the corresponding media. The trans passive behavior of alloy C-22 was mainly influenced by temperature and solution chemistry. A clear trans passive peak

  2. Evolution of a novel subfamily of nuclear receptors with members that each contain two DNA binding domains

    Directory of Open Access Journals (Sweden)

    Hirai Hirohisa

    2007-02-01

    Full Text Available Abstract Background Nuclear receptors (NRs are important transcriptional modulators in metazoans which regulate transcription through binding to the promoter region of their target gene by the DNA binding domain (DBD and activation or repression of mRNA synthesis through co-regulators bound to the ligand binding domain (LBD. NRs typically have a single DBD with a LBD. Results Three nuclear receptors named 2DBD-NRs, were identified from the flatworm Schistosoma mansoni that each possess a novel set of two DBDs in tandem with a LBD. They represent a novel NR modular structure: A/B-DBD-DBD-hinge-LBD. The 2DBD-NRs form a new subfamily of NRs, VII. By database mining, 2DBD-NR genes from other flatworm species (Schmidtea mediterranea and Dugesia japonica, from Mollusks (Lottia gigantean and from arthropods (Daphnia pulex were also identified. All 2DBD-NRs possess a P-box sequence of CEACKK in the first DBD, which is unique to 2DBD-NRs, and a P-box sequence of CEGCKG in the second DBD. Phylogenetic analyses of both DBD and ligand binding domain sequences showed that 2DBD-NR genes originate from a common two DBD-containing ancestor gene. A single 2DBD-NR orthologue was found in Arthropoda, Platyhelminths and Mollusca. Subsequent 2DBD-NR gene evolution in Mollusks and Platyhelminths involved gene duplication. Chromosome localization of S. mansoni 2DBD-NR genes by Fluorescent in situ hybridization (FISH suggests that 2DBD-NR genes duplicated on different chromosomes in the Platyhelminths. Dimerization of Sm2DBDα indicates that 2DBD-NRs may act as homodimers, suggesting either that two repeats of a half-site are necessary for each DBD of 2DBD-NRs to bind to its target gene, or that each 2DBD-NR can recognize multiple sites. Conclusion 2DBD-NRs share a common ancestor gene which possessed an extra DBD that likely resulted from a recombination event. After the split of the Arthropods, Mollusks and Platyhelminths, 2DBD-NR underwent a recent duplication in a

  3. Well responses to barometric-pressure fluctuations in confined and semi-confined aquifers

    Science.gov (United States)

    Mohammed, G. A.; Jin, W.; Butler, J. J., Jr.; Reboulet, E. C.

    2009-04-01

    Modern data logger and sensor technology enable well responses to barometric pressure changes to be monitored at a high frequency and precision. In this presentation, we demonstrate that such monitoring data can be utilized not just for the conventional calculation of a well's barometric efficiency but also to provide valuable information for site characterization applications. We investigate the water-level responses of wells in confined and semi-confined aquifers to changes in barometric pressure and show how simple analytical solutions can be fit to experimentally determined barometric response functions to place bounds on the properties of the confining bed. We demonstrate our approach at the Larned Research Site, located along the Arkansas River in south-central Kansas in the Great Plains region of the United States. The site contains monitoring wells tapping an unconsolidated, semi-confined aquifer (High Plains Aquifer) overlain by a clay unit and a shallow, unconfined aquifer. Water levels and atmospheric pressure have been monitored in the wells at 15-minute intervals for up to seven years. The spatial and temporal changes in the barometric response functions provide important insights on the degree of confinement and its change in space and time. Short term (hour or less) response functions indicate a classical confined aquifer, whereas long term responses (day) show semi-confined behavior, an indication that the air pressure exerted on the water table is being transmitted downward through the confining bed. The barometric response functions vary little in space, indicating the homogeneity of the confining bed, but do vary temporally as a function of the water table elevation and the pneumatic diffusivity of the vadose zone.

  4. eEF1A Mediates the Nuclear Export of SNAG-Containing Proteins via the Exportin5-Aminoacyl-tRNA Complex

    Directory of Open Access Journals (Sweden)

    José Manuel Mingot

    2013-11-01

    Full Text Available Exportin5 mediates the nuclear export of double-stranded RNAs, including pre-microRNAs, adenoviral RNAs, and tRNAs. When tRNAs are aminoacylated, the Exportin5-aminoacyl (aa-tRNA complex recruits and coexports the translation elongation factor eEF1A. Here, we show that eEF1A binds to Snail transcription factors when bound to their main target, the E-cadherin promoter, facilitating their export to the cytoplasm in association with the aa-tRNA-Exportin5 complex. Snail binds to eEF1A through the SNAG domain, a protein nuclear export signal present in several transcription factor families, and this binding is regulated by phosphorylation. Thus, we describe a nuclear role for eEF1A and provide a mechanism for protein nuclear export that attenuates the activity of SNAG-containing transcription factors.

  5. Transition metal catalysis in confined spaces.

    Science.gov (United States)

    Leenders, Stefan H A M; Gramage-Doria, Rafael; de Bruin, Bas; Reek, Joost N H

    2015-01-21

    Transition metal catalysis plays an important role in both industry and in academia where selectivity, activity and stability are crucial parameters to control. Next to changing the structure of the ligand, introducing a confined space as a second coordination sphere around a metal catalyst has recently been shown to be a viable method to induce new selectivity and activity in transition metal catalysis. In this review we focus on supramolecular strategies to encapsulate transition metal complexes with the aim of controlling the selectivity via the second coordination sphere. As we will discuss, catalyst confinement can result in selective processes that are impossible or difficult to achieve by traditional methods. We will describe the template-ligand approach as well as the host-guest approach to arrive at such supramolecular systems and discuss how the performance of the catalyst is enhanced by confining it in a molecular container.

  6. Innovative confinement concepts workshop

    International Nuclear Information System (INIS)

    Kirkpatrick, R.C.

    1998-01-01

    The Innovative Confinement Concepts Workshop occurred in California during the week preceding the Second Symposium on Current Trends in International Fusion Research. An informal report was made to the Second Symposium. A summary of the Workshop concluded that some very promising ideas were presented, that innovative concept development is a central element of the restructured US DOE. Fusion Energy Sciences program, and that the Workshop should promote real scientific progress in fusion

  7. Energy confinement in tokamaks

    International Nuclear Information System (INIS)

    Sugihara, M.; Singer, C.

    1986-08-01

    A straightforward generalization is made of the ohmic heating energy confinement scalings of Pfeiffer and Waltz and Blackwell et. al. The resulting model is systematically calibrated to published data from limiter tokamaks with ohmic, electron cyclotron, and neutral beam heating. With considerably fewer explicitly adjustable free parameters, this model appears to give a better fit to the available data for limiter discharges than the combined ohmic/auxiliary heating model of Goldston

  8. Confined exciton spectroscopy

    International Nuclear Information System (INIS)

    Torres, Clivia M.S.

    1998-01-01

    Full text: In this work, the exciton is considered as a sensor of the electronic and optical properties of materials such as semiconductors, which have size compared to the exciton De Broglie wavelength, approximately 20 nm, depending on the semiconductor. Examples of electron-phonon, electron-electron, photon-electron, exciton-polariton, phonon-plasmon, are presented, under different confinement conditions such as quantum wells, superlattices

  9. Neutron imaging with bubble chambers for inertial confinement fusion

    Science.gov (United States)

    Ghilea, Marian C.

    One of the main methods to obtain energy from controlled thermonuclear fusion is inertial confinement fusion (ICF), a process where nuclear fusion reactions are initiated by heating and compressing a fuel target, typically in the form of a pellet that contains deuterium and tritium, relying on the inertia of the fuel mass to provide confinement. In inertial confinement fusion experiments, it is important to distinguish failure mechanisms of the imploding capsule and unambiguously diagnose compression and hot spot formation in the fuel. Neutron imaging provides such a technique and bubble chambers are capable of generating higher resolution images than other types of neutron detectors. This thesis explores the use of a liquid bubble chamber to record high yield 14.1 MeV neutrons resulting from deuterium-tritium fusion reactions on ICF experiments. A design tool to deconvolve and reconstruct penumbral and pinhole neutron images was created, using an original ray tracing concept to simulate the neutron images. The design tool proved that misalignment and aperture fabrication errors can significantly decrease the resolution of the reconstructed neutron image. A theoretical model to describe the mechanism of bubble formation was developed. A bubble chamber for neutron imaging with Freon 115 as active medium was designed and implemented for the OMEGA laser system. High neutron yields resulting from deuterium-tritium capsule implosions were recorded. The bubble density was too low for neutron imaging on OMEGA but agreed with the model of bubble formation. The research done in here shows that bubble detectors are a promising technology for the higher neutron yields expected at National Ignition Facility (NIF).

  10. Nuclear law - Nuclear safety

    International Nuclear Information System (INIS)

    Pontier, Jean-Marie; Roux, Emmanuel; Leger, Marc; Deguergue, Maryse; Vallar, Christian; Pissaloux, Jean-Luc; Bernie-Boissard, Catherine; Thireau, Veronique; Takahashi, Nobuyuki; Spencer, Mary; Zhang, Li; Park, Kyun Sung; Artus, J.C.

    2012-01-01

    This book contains the contributions presented during a one-day seminar. The authors propose a framework for a legal approach to nuclear safety, a discussion of the 2009/71/EURATOM directive which establishes a European framework for nuclear safety in nuclear installations, a comment on nuclear safety and environmental governance, a discussion of the relationship between citizenship and nuclear, some thoughts about the Nuclear Safety Authority, an overview of the situation regarding the safety in nuclear waste burying, a comment on the Nome law with respect to electricity price and nuclear safety, a comment on the legal consequences of the Fukushima accident on nuclear safety in the Japanese law, a presentation of the USA nuclear regulation, an overview of nuclear safety in China, and a discussion of nuclear safety in the medical sector

  11. Nuclear fusion. (Latest citations from the NTIS bibliographic database). Published Search

    International Nuclear Information System (INIS)

    1993-08-01

    The bibliography contains citations concerning research, development, and assessment of nuclear fusion for applications in reactor engineering and technology. Citations discuss various engineering problems associated with reactor design, magnetic systems, nuclear materials, plasma generation and control, blankets, environments, economics, and safety. Also discussed are tokamak devices, stellarators, inertial confinement, reflectometry, and magnetohydrodynamics. Studies sponsored by the Department of Energy are not included. (Contains a minimum of 249 citations and includes a subject term index and title list.)

  12. Pre-operational proof and leakage rate testing requirements for concrete containment structures for CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1994-02-01

    This Standard provides the requirements for pre-operational proof tests and leakage rate tests of concrete containment structures of a containment system designed as Class Containment components. 1 fig

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

    International Nuclear Information System (INIS)

    Benboudjema, F.

    2002-12-01

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

  14. Three-dimensional all-speed CFD code for safety analysis of nuclear reactor containment: Status of GASFLOW parallelization, model development, validation and application

    Energy Technology Data Exchange (ETDEWEB)

    Xiao, Jianjun, E-mail: jianjun.xiao@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Travis, John R., E-mail: jack_travis@comcast.com [Engineering and Scientific Software Inc., 3010 Old Pecos Trail, Santa Fe, NM 87505 (United States); Royl, Peter, E-mail: peter.royl@partner.kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Necker, Gottfried, E-mail: gottfried.necker@partner.kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Svishchev, Anatoly, E-mail: anatoly.svishchev@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Jordan, Thomas, E-mail: thomas.jordan@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2016-05-15

    Highlights: • 3-D scalable semi-implicit pressure-based CFD code for containment safety analysis. • Robust solution algorithm valid for all-speed flows. • Well validated and widely used CFD code for hydrogen safety analysis. • Code applied in various types of nuclear reactor containments. • Parallelization enables high-fidelity models in large scale containment simulations. - Abstract: GASFLOW is a three dimensional semi-implicit all-speed CFD code which can be used to predict fluid dynamics, chemical kinetics, heat and mass transfer, aerosol transportation and other related phenomena involved in postulated accidents in nuclear reactor containments. The main purpose of the paper is to give a brief review on recent GASFLOW code development, validations and applications in the field of nuclear safety. GASFLOW code has been well validated by international experimental benchmarks, and has been widely applied to hydrogen safety analysis in various types of nuclear power plants in European and Asian countries, which have been summarized in this paper. Furthermore, four benchmark tests of a lid-driven cavity flow, low Mach number jet flow, 1-D shock tube and supersonic flow over a forward-facing step are presented in order to demonstrate the accuracy and wide-ranging capability of ICE’d ALE solution algorithm for all-speed flows. GASFLOW has been successfully parallelized using the paradigms of Message Passing Interface (MPI) and domain decomposition. The parallel version, GASFLOW-MPI, adds great value to large scale containment simulations by enabling high-fidelity models, including more geometric details and more complex physics. It will be helpful for the nuclear safety engineers to better understand the hydrogen safety related physical phenomena during the severe accident, to optimize the design of the hydrogen risk mitigation systems and to fulfill the licensing requirements by the nuclear regulatory authorities. GASFLOW-MPI is targeting a high

  15. Transport properties of supercooled confined water

    International Nuclear Information System (INIS)

    Mallamace, F.; Baglioni, P.; Corsaro, C.; Spooren, J.; Stanley, H.E.; Chen, S.-H.

    2011-01-01

    We present an overview of recent experiments performed on water in the deeply supercooled region, a temperature region of fundamental importance in the science of water. We examine data generated by nuclear magnetic resonance, quasi-elastic neutron scattering, Fourier-transform infrared spectroscopy, and Raman spectroscopy, and study water confined in nanometer-scale environments. When contained within small pores, water does not crystallize and can be supercooled well below its homogeneous nucleation temperature T H. On this basis, it is possible to carry out a careful analysis of the well-known thermodynamic anomalies of water. Studying the temperature and pressure dependencies of water dynamics, we show that the liquid-liquid phase transition (LLPT) hypothesis represents a reliable model for describing liquid water. In this model, liquid water is a mixture of two different local structures: a low density liquid (LDL) and a high-density liquid (HDL). The LLPT line terminates at a low-T liquid-liquid critical point. We discuss the following experimental findings: 1.) the crossover from non-Arrhenius behavior at high T to Arrhenius behavior at low T in transport parameters; 2.) the breakdown of the Stokes-Einstein relation; 3.) the existence of a Widom line, which is the locus of points corresponding to a maximum correlation length in the P-T phase diagram and which ends in the liquid-liquid critical point; 4.) the direct observation of the LDL phase; and 5.) the minimum in the density at approximately 70 K below the temperature of the density maximum. In our opinion these results strongly support the LLPT hypothesis. All of the basic science and technology community should be impressed by the fact that, although the few ideas (apparently elementary) developed concerning water approximately 27 centuries ago have changed very little up to now, because of the current expansion in our knowledge in this area, they can begin to change in the near future.

  16. Inertial confinement fusion (ICF)

    International Nuclear Information System (INIS)

    Nuckolls, J.

    1977-01-01

    The principal goal of the inertial confinement fusion program is the development of a practical fusion power plant in this century. Rapid progress has been made in the four major areas of ICF--targets, drivers, fusion experiments, and reactors. High gain targets have been designed. Laser, electron beam, and heavy ion accelerator drivers appear to be feasible. Record-breaking thermonuclear conditions have been experimentally achieved. Detailed diagnostics of laser implosions have confirmed predictions of the LASNEX computer program. Experimental facilities are being planned and constructed capable of igniting high gain fusion microexplosions in the mid 1980's. A low cost long lifetime reactor design has been developed

  17. Hadrosynthesis and Quark Confinement

    Directory of Open Access Journals (Sweden)

    Satz Helmut

    2014-04-01

    Full Text Available Multihadron production in high energy collisions, from e+e− annihilation to heavy ion interactions, shows remarkable thermal behaviour, specified by a universal “Hagedorn” temperature. We argue that this hadronic radiation is formed by tunnelling through the event horizon of colour confinement, i.e., that it is the QCD counterpart of Hawking-Unruh radiation from black holes. It is shown to be emitted at a universal temperature TH ≃ (σ/2π1/2, where σ denotes the string tension. Since the event horizon does not allow information transfer, the radiation is thermal “at birth”.

  18. Confinement and 4-manifolds

    CERN Multimedia

    CERN. Geneva

    2014-01-01

    In this talk I will survey a connection between two very challenging problems, one in physics and one in math. The physics problem involves quantitative understanding of confinement in a system with least amount of supersymmetry that has been studied so far and that has a wide range of applications, from semi-realistic string models to qualitatively new examples of gauge-gravity duality. Surprisingly, the rich physics of this system translates into incredibly rich mathematics of the only remaining unsolved case of the Poincare conjecture.

  19. Dynamical properties of nimodipine molecules confined in SBA-15 matrix

    International Nuclear Information System (INIS)

    Kiwilsza, A.; Pajzderska, A.; Mielcarek, J.; Jenczyk, J.; Wąsicki, J.

    2016-01-01

    Highlights: • Dynamics of confined nimodipine (drug) in SBA-15 studied by ssNMR. • Higher mobility of confined molecules in comparison with bulk drug. • Decreasing of the energy barrier for reorientation of methyl groups in confined drug. - Abstract: The paper reports results of 13 C and 1 H ssNMR for nimodipine confined in mesopores of SBA-15 for the samples (i) containing nimodipine molecules inside and on the external surface of silica, (ii) containing nimodipine only inside pores forming an incomplete monolayer on the surface (iii) for bulk nimodipine. The measurements permitted comparison of the dynamics of nimodipine bulk and confined in pores. The confined nimodipine is in an amorphous state and has additional degrees of rotational freedom with respect to the bulk one. The height of the energy barrier related to the rotation of methyl groups in confined nimodipine is lower than in bulk nimodipine. The higher mobility of nimodipine molecules confined in silica pores can explain the higher release rate of nimodipine from silica matrix than dissolution rate of bulk drug.

  20. Dynamical properties of nimodipine molecules confined in SBA-15 matrix

    Energy Technology Data Exchange (ETDEWEB)

    Kiwilsza, A. [Faculty of Physics, Adam Mickiewicz University, Umultowska 85, 61-614 Poznan (Poland); NanoBioMedical Center, Adam Mickiewicz University, Umultowska 85, 61-614 Poznan (Poland); Pajzderska, A. [Faculty of Physics, Adam Mickiewicz University, Umultowska 85, 61-614 Poznan (Poland); Mielcarek, J. [Department of Inorganic and Analytical Chemistry, Poznan University of Medical Sciences, Grunwaldzka 6, 60-780 Poznan (Poland); Jenczyk, J. [NanoBioMedical Center, Adam Mickiewicz University, Umultowska 85, 61-614 Poznan (Poland); Wąsicki, J., E-mail: jwasicki@amu.edu.pl [Faculty of Physics, Adam Mickiewicz University, Umultowska 85, 61-614 Poznan (Poland); NanoBioMedical Center, Adam Mickiewicz University, Umultowska 85, 61-614 Poznan (Poland)

    2016-08-22

    Highlights: • Dynamics of confined nimodipine (drug) in SBA-15 studied by ssNMR. • Higher mobility of confined molecules in comparison with bulk drug. • Decreasing of the energy barrier for reorientation of methyl groups in confined drug. - Abstract: The paper reports results of {sup 13}C and {sup 1}H ssNMR for nimodipine confined in mesopores of SBA-15 for the samples (i) containing nimodipine molecules inside and on the external surface of silica, (ii) containing nimodipine only inside pores forming an incomplete monolayer on the surface (iii) for bulk nimodipine. The measurements permitted comparison of the dynamics of nimodipine bulk and confined in pores. The confined nimodipine is in an amorphous state and has additional degrees of rotational freedom with respect to the bulk one. The height of the energy barrier related to the rotation of methyl groups in confined nimodipine is lower than in bulk nimodipine. The higher mobility of nimodipine molecules confined in silica pores can explain the higher release rate of nimodipine from silica matrix than dissolution rate of bulk drug.

  1. Analysis of Dust Samples Collected from an Unused Spent Nuclear Fuel Interim Storage Container at Hope Creek, Delaware.

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Enos, David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-03-01

    In July, 2014, the Electric Power Research Institute and industry partners sampled dust on the surface of an unused canister that had been stored in an overpack at the Hope Creek Nuclear Generating Station for approximately one year. The foreign material exclusion (FME) cover that had been on the top of the canister during storage, and a second recently - removed FME cover, were also sampled. This report summarizes the results of analyses of dust samples collected from the unused Hope Creek canister and the FME covers. Both wet and dry samples of the dust/salts were collected, using SaltSmart(TM) sensors and Scotch - Brite(TM) abrasive pads, respectively. The SaltSmart(TM) samples were leached and the leachate analyzed chemically to determine the composition and surface load per unit area of soluble salts present on the canister surface. The dry pad samples were analyzed by X-ray fluorescence and by scanning electron microscopy to determine dust texture and mineralogy; and by leaching and chemical analysis to deter mine soluble salt compositions. The analyses showed that the dominant particles on the canister surface were stainless steel particles, generated during manufacturing of the canister. Sparse environmentally - derived silicates and aluminosilicates were also present. Salt phases were sparse, and consisted of mostly of sulfates with rare nitrates and chlorides. On the FME covers, the dusts were mostly silicates/aluminosilicates; the soluble salts were consistent with those on the canister surface, and were dominantly sulfates. It should be noted that the FME covers were w ashed by rain prior to sampling, which had an unknown effect of the measured salt loads and compositions. Sulfate salts dominated the assemblages on the canister and FME surfaces, and in cluded Ca - SO4 , but also Na - SO4 , K - SO4 , and Na - Al - SO4 . It is likely that these salts were formed by particle - gas conversion reactions, either

  2. Quark Confinement and Strings

    Science.gov (United States)

    't Hooft, Gerardus

    QCD was proposed as a theory for the strong interactions long before we had any idea as to how it could be that its fundamental constituents, the quarks, are never seen as physical particles. Massless gluons also do not exist as free particles. How can this be explained? The first indication that this question had to be considered in connection with the topological structure of a gauge theory came when Nielsen and Olesen observed the occurrence of stable magnetic vortex structures [1] in the Abelian Higgs model. Expanding on such ideas, the magnetic monopole solution was found [2]. Other roundabout attempts to understand confinement involve instantons. Today, we have better interpretations of these topological structures, including a general picture of the way they do lead to unbound potentials confining quarks. It is clear that these unbound potentials can be ascribed to a string-like structure of the vortices formed by the QCD field lines. Can string theory be used to analyze QCD? Many researchers think so. The leading expert on this is Sacha Polyakov. In his instructive account he adds how he experienced the course of events in Gauge Theory, emphasizing the fact that quite a few discoveries often ascribed to researchers from the West, actually were made independently by scientists from the Soviet Union…

  3. Confinement from correlation functions

    Science.gov (United States)

    Fister, Leonard; Pawlowski, Jan M.

    2013-08-01

    We compute the Polyakov loop potential in Yang-Mills theory from the fully dressed primitively divergent correlation functions only. This is done in a variety of functional approaches ranging from functional renormalization group equations over Dyson-Schwinger equations to two-particle irreducible functionals. We present a confinement criterion that links the infrared behavior of propagators and vertices to the Polyakov loop expectation value. The present work extends the works of [J. Braun , Phys. Lett. B 684, 262 (2010)PYLBAJ0370-2693; F. Marhauser and J. M. Pawlowski, arXiv:0812.1144; J. Braun , Eur. Phys. J. C 70, 689 (2010)EPCFFB1434-6044] to general functional methods and sharpens the confinement criterion presented there. The computations are based on the thermal correlation functions in the Landau gauge calculated in [L. Fister and J. M. Pawlowski, arXiv:1112.5440; L. Fister and J. M. Pawlowski, arXiv:1112.5429; L. Fister, Ph.D. thesis, Heidelberg University, 2012].

  4. NEW APPROACHES TO CONFINED ALPHA DIAGNOSTICS

    Energy Technology Data Exchange (ETDEWEB)

    FISHER,R.K

    2004-04-01

    Three new approaches to obtain information on the confined fast alphas in International Thermonuclear Experimental Reactor (ITER) are proposed. The first technique measures the energetic charge exchange (CX) neutrals that result from the alpha collision-induced knock-on fuel ion tails undergoing electron capture on the MeV D neutral beams planned for heating and current drive. The second technique measures the energetic knock-on neutron tail due to alphas using the lengths of the proton recoil tracks produced by neutron collisions in nuclear emulsions. The range of the 14 to 20 MeV recoil protons increases by {approx}140 microns per MeV. The third approach would measure the CX helium neutrals resulting from confined alphas capturing two electrons in the ablation cloud surrounding a dense gas jet that has been proposed for disruption mitigation in ITER.

  5. NEW APPROACHES TO CONFINED ALPHA DIAGNOSTICS

    International Nuclear Information System (INIS)

    FISHER, R.K.

    2004-01-01

    Three new approaches to obtain information on the confined fast alphas in International Thermonuclear Experimental Reactor (ITER) are proposed. The first technique measures the energetic charge exchange (CX) neutrals that result from the alpha collision-induced knock-on fuel ion tails undergoing electron capture on the MeV D neutral beams planned for heating and current drive. The second technique measures the energetic knock-on neutron tail due to alphas using the lengths of the proton recoil tracks produced by neutron collisions in nuclear emulsions. The range of the 14 to 20 MeV recoil protons increases by ∼140 microns per MeV. The third approach would measure the CX helium neutrals resulting from confined alphas capturing two electrons in the ablation cloud surrounding a dense gas jet that has been proposed for disruption mitigation in ITER

  6. Study of high gain spherical shell ICF targets containing uniform layers of liquid deuterium tritium fuel. A numericial model for analyzing thermal layering of liquid mixtures of hydrogen isotopes inside a spherical inertial confinement fusion target: Final report

    International Nuclear Information System (INIS)

    Simpson, E.M.; Kim, Kyekyoon

    1994-05-01

    A numerical model has been developed to describe the thermally induced behavior of a liquid layer of hydrogen isotopes inside a spherical Inertial Confinement Fusion (ICF) target and to calculate the far-field temperature gradient which will sustain a uniform liquid layer. This method is much faster than the trial-and-error method previously employed. The governing equations are the equations of continuity, momentum, energy, mass diffusion-convection, and conservation of the individual isotopic species. Ordinary and thermal diffusion equations for the diffusion of fluxes of the species are included. These coupled equations are solved by a finite-difference method using upwind schemes, variable mesh, and rigorous boundary conditions. The solution methodology unique to the present problem is discussed in detail. in particular, the significance of the surface tension gradient driven flows (also called Marangoni flows) in forming uniform liquid layers inside ICF targets is demonstrated. Using the theoretical model, the values of the externally applied thermal gradients that give rise to uniform liquid layers of hydrogen inside a cryogenic spherical-shell ICF target are calculated, and the results compared with the existing experimental data

  7. Use of computational fluid dynamics codes for safety analysis of nuclear reactor systems, including containment. Summary report of a technical meeting

    International Nuclear Information System (INIS)

    2003-11-01

    Safety analysis is an important tool for justifying the safety of nuclear power plants. Typically, this type of analysis is performed by means of system computer codes with one dimensional approximation for modelling real plant systems. However, in the nuclear area there are issues for which traditional treatment using one dimensional system codes is considered inadequate for modelling local flow and heat transfer phenomena. There is therefore increasing interest in the application of three dimensional computational fluid dynamics (CFD) codes as a supplement to or in combination with system codes. There are a number of both commercial (general purpose) CFD codes as well as special codes for nuclear safety applications available. With further progress in safety analysis techniques, the increasing use of CFD codes for nuclear applications is expected. At present, the main objective with respect to CFD codes is generally to improve confidence in the available analysis tools and to achieve a more reliable approach to safety relevant issues. An exchange of views and experience can facilitate and speed up progress in the implementation of this objective. Both the International Atomic Energy Agency (IAEA) and the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA) believed that it would be advantageous to provide a forum for such an exchange. Therefore, within the framework of the Working Group on the Analysis and Management of Accidents of the NEA's Committee on the Safety of Nuclear Installations, the IAEA and the NEA agreed to jointly organize the Technical Meeting on the Use of Computational Fluid Dynamics Codes for Safety Analysis of Reactor Systems, including Containment. The meeting was held in Pisa, Italy, from 11 to 14 November 2002. The publication constitutes the report of the Technical Meeting. It includes short summaries of the presentations that were made and of the discussions as well as conclusions and

  8. ICENES '91:Sixth international conference on emerging nuclear energy systems

    Energy Technology Data Exchange (ETDEWEB)

    1991-01-01

    This document contains the program and abstracts of the sessions at the Sixth International Conference on Emerging Nuclear Energy Systems held June 16--21, 1991 at Monterey, California. These sessions included: The plenary session, fission session, fission and nonelectric session, poster session 1P; (space propulsion, space nuclear power, electrostatic confined fusion, fusion miscellaneous, inertial confinement fusion, [mu]-catalyzed fusion, and cold fusion); Advanced fusion session, space nuclear session, poster session 2P, (nuclear reactions/data, isotope separation, direct energy conversion and exotic concepts, fusion-fission hybrids, nuclear desalting, accelerator waste-transmutation, and fusion-based chemical recycling); energy policy session, poster session 3P (energy policy, magnetic fusion reactors, fission reactors, magnetically insulated inertial fusion, and nuclear explosives for power generation); exotic energy storage and conversion session; and exotic energy storage and conversion; review and closing session.

  9. ICENES `91:Sixth international conference on emerging nuclear energy systems. Program and abstracts

    Energy Technology Data Exchange (ETDEWEB)

    1991-12-31

    This document contains the program and abstracts of the sessions at the Sixth International Conference on Emerging Nuclear Energy Systems held June 16--21, 1991 at Monterey, California. These sessions included: The plenary session, fission session, fission and nonelectric session, poster session 1P; (space propulsion, space nuclear power, electrostatic confined fusion, fusion miscellaneous, inertial confinement fusion, {mu}-catalyzed fusion, and cold fusion); Advanced fusion session, space nuclear session, poster session 2P, (nuclear reactions/data, isotope separation, direct energy conversion and exotic concepts, fusion-fission hybrids, nuclear desalting, accelerator waste-transmutation, and fusion-based chemical recycling); energy policy session, poster session 3P (energy policy, magnetic fusion reactors, fission reactors, magnetically insulated inertial fusion, and nuclear explosives for power generation); exotic energy storage and conversion session; and exotic energy storage and conversion; review and closing session.

  10. Specialist meeting on selected containment severe accident management strategies. Summary and conclusions

    International Nuclear Information System (INIS)

    1994-01-01

    The CSNI Specialist Meeting on Selected Containment Severe Accident Management Strategies held in Stockholm, Sweden in June 1994 was organised by the Task Group on Containment Aspects of Severe Accident Management (CAM) of CSNI's Principal Working Group on the Confinement of Accidental Radioactive Releases (PWG4) in collaboration with the Swedish Nuclear Power Inspectorate (SKI). Conclusions and recommendations are given for each of the sessions of the workshops: Containment accident management strategies (general aspects); hydrogen management techniques and other containment accident management techniques; surveillance and protection of containment function

  11. Experimental study on the shrinkage properties and cracking potential of high strength concrete containing industrial by-products for nuclear power plant concrete

    Energy Technology Data Exchange (ETDEWEB)

    KIm, Baek Joong; Yi, Chong Ku [School of Civil, Environmental and Architectural Engineering, Korea University, Seoul (Korea, Republic of)

    2017-02-15

    In Korea, attempts have been made to develop high strength concrete for the safety and design life improvement of nuclear power plants. In this study, the cracking potentials of nuclear power plant-high strength concretes (NPP-HSCs) containing industrial by-products with W/B 0.34 and W/B 0.28, which are being reviewed for their application in the construction of containment structures, were evaluated through autogenous shrinkage, unrestrained drying shrinkage, and restrained drying shrinkage experiments. The cracking potentials of the NPP-HSCs with W/B 0.34 and W/B 0.28 were in the order of 0.34FA25 > 0.34FA25BFS25 > 0.34BFS50 > 0.34BFS65SF5 and 0.28FA25SF5 >> 0.28BFS65SF5 > 0.28BFS45SF5 > 0.28 FA20BFS25SF5, respectively. The cracking potentials of the seven mix proportions excluding 0.28FA25SF5 were lower than that of the existing nuclear power plant concrete; thus, the durability of a nuclear power plant against shrinkage cracking could be improved by applying the seven mix proportions with low cracking potentials.

  12. Hydrogen countermeasures and activity retention by filtered venting for WWER-440/V230 NPP confinement

    International Nuclear Information System (INIS)

    Feuerbach, R.

    2001-01-01

    In order to prevent loss of confinement integrity caused by steam and hydrogen generation nuclear power plants in the Federal Republic of Germany as well as in most other European countries have been or will be back-fitted with a system for filtering the confinement atmosphere prior to release to the environment and a system for reducing and measuring the H 2 concentration inside the confinement. For these tasks systems for confinement atmosphere control are presented, capable of: handling high H 2 production rates and cleaning of high contaminated confinement atmosphere. (author)

  13. The study of the container types used for transport and final disposal of the radioactive wastes resulting from decommissioning of nuclear facilities

    International Nuclear Information System (INIS)

    Postelnicu, C.

    1998-01-01

    The purpose of the present paper is to select from a variety of package forms and capacities some containers which will be used for transport and disposal of the radioactive wastes resulting from decommissioning of nuclear facilities into the National Repository for Radioactive Waste - Baita, Bihor county. Taken into account the possibilities of railway and / or road transport and waste disposal in our country, detailed container classification was given in order to use them for radioactive waste transport and final disposal from decommissioning of IFIN-HH Research Reactor. (author)

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

    International Nuclear Information System (INIS)

    Kumpf, H.

    1976-01-01

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

  15. Trinucleotide-repeat expanded and normal DMPK transcripts contain unusually long poly(A) tails despite differential nuclear residence.

    Science.gov (United States)

    Gudde, Anke E E G; van Kessel, Ingeborg D G; André, Laurène M; Wieringa, Bé; Wansink, Derick G

    2017-06-01

    In yeast and higher eukaryotes nuclear retention of transcripts may serve in control over RNA decay, nucleocytoplasmic transport and premature cytoplasmic appearance of mRNAs. Hyperadenylation of RNA is known to be associated with nuclear retention, but the cause-consequence relationship between hyperadenylation and regulation of RNA nuclear export is still unclear. We compared polyadenylation status between normal and expanded DMPK transcripts in muscle cells and tissues derived from unaffected individuals and patients with myotonic dystrophy type 1 (DM1). DM1 is an autosomal dominant disorder caused by (CTG)n repeat expansion in the DMPK gene. DM1 etiology is characterized by an almost complete block of nuclear export of DMPK transcripts carrying a long (CUG)n repeat, including aberrant sequestration of RNA-binding proteins. We show here by use of cell fractionation, RNA size separation and analysis of poly(A) tail length that a considerable fraction of transcripts from the normal DMPK allele is also retained in the nucleus (~30%). They carry poly(A) tails with an unusually broad length distribution, ranging between a few dozen to >500 adenosine residues. Remarkably, expanded DMPK (CUG)n transcripts from the mutant allele, almost exclusively nuclear, carry equally long poly(A) tails. Our findings thus suggest that nuclear retention may be a common feature of regulation of DMPK RNA expression. The typical forced nuclear residence of expanded DMPK transcripts affects this regulation in tissues of DM1 patients, but not through hyperadenylation. Copyright © 2017 The Authors. Published by Elsevier B.V. All rights reserved.

  16. Development of 3D models of buildings for containment of the nuclear power plant of Almaraz and of the Trillo Nuclear with the GOTHIC 8.0 code; Desarrollo de modelos 3D de los edificios de conten cion de la Central Nuclear de Almaraz y de la Central Nuclear de Trillo con el codigo GOTHIC 8.0

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Bocanegra Melian, R.; Fernandez Cosils, K.; Barreira Pereira, P.; Rey Peinado, L.; Posada Barral, J. M.

    2014-07-01

    The objective of the first phase of the research of CNAT and the UPM project is the construction of several three-dimensional models detailed GOTHIC 8.0 code of containment of a buildings plant type PWR-W and KWU, corresponding to the Central Nuclear de Almaraz (CNA) and Trillo (CNT) respectively. (Author)

  17. Sensitivity of Sump Water Temperature to Containment Integrity

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Misuk; Kim, Seoung Rae [Nuclear Engineering Service and Solution, Daejeon (Korea, Republic of)

    2014-05-15

    This paper is focused on the containment behavior analysis in the above described cases using GOTHIC-IST (generation of thermal-hydraulic information for containments, industry standard toolset). GOTHIC-IST version 7.2a is an integrated, general purpose thermal-hydraulics software package for design, licensing, safety and operating analysis of nuclear power plant containments and other confinement buildings. In this study, we perform the sensitivity the sump water temperature to containment integrity. For 35% RIH break accident with the malfunction of spray system, local air coolers, ECC(emergency core cooling) pump and heat exchanger, the peak pressure at boiler room do not exceed the design pressure 124kPa(g) of the containment and containment integrity is secured. If accompanied the malfunction of heat exchanger or pump in the time of low pressure safety injection, of ECCS, it will be one of the aggravating factors to the integrity of core and containment.

  18. Section 1. Confinement systems

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    Major experimental and theoretical results achieved by the Controlled Thermonuclear Research (CTR) program at Lawrence Livermore Laboratory during FY 1975 gave the greatest encouragement to date that the ultimate goal of a deuterium-tritium-fueled mirror reactor can be reached. In the experimental program, the year was characterized by unusually important physics results from the 2XIIB experiment and by significant steps in the plan to change the Baseball II mode of operation. The stabilization of ion-cyclotron instabilities in the 2XIIB experiment by the introduction of an auxiliary warm plasma permitted the buildup of a high-temperature, high-density plasma with an n tau parameter an order of magnitude larger than the 2XII experiment.I In the Baseball II experiment, preliminary tests and computer predictions indicated that a dense, transient, target plasma can be created by laser irradiation of a pellet in midflight through the center of the Baseball confinement zone

  19. Confinement effects in sesquioxydes

    Energy Technology Data Exchange (ETDEWEB)

    Mercier, B. [Laboratoire de Physico-Chimie des Materiaux Luminescents, UMR 5620 du CNRS, Universite Claude Bernard Lyon I, 10 rue A.M. Ampere 69622 Villeurbanne Cedex (France); Dujardin, C. [Laboratoire de Physico-Chimie des Materiaux Luminescents, UMR 5620 du CNRS, Universite Claude Bernard Lyon I, 10 rue A.M. Ampere 69622 Villeurbanne Cedex (France); Ledoux, G. [Laboratoire de Physico-Chimie des Materiaux Luminescents, UMR 5620 du CNRS, Universite Claude Bernard Lyon I, 10 rue A.M. Ampere 69622 Villeurbanne Cedex (France)]. E-mail: ledoux@pcml.univ-lyon1.fr; Louis, C. [Laboratoire de Physico-Chimie des Materiaux Luminescents, UMR 5620 du CNRS, Universite Claude Bernard Lyon I, 10 rue A.M. Ampere 69622 Villeurbanne Cedex (France); Tillement, O. [Laboratoire de Physico-Chimie des Materiaux Luminescents, UMR 5620 du CNRS, Universite Claude Bernard Lyon I, 10 rue A.M. Ampere 69622 Villeurbanne Cedex (France); Perriat, P. [Groupe d' Etude de Metallurgie Physique et de Physique des Materiaux, UMR 5510 du CNRS, INSA de Lyon, 69622 Villeurbanne Cedex (France)

    2006-07-15

    When the size of a particle is decreased to nano-scales, confinement effects induce strong changes. The luminescence of Eu{sup 3+}-doped Gd{sub 2}O{sub 3} nanoparticles has been studied as a function of both the temperature and the size of the nanoparticles. Two consequences of the small sizes in Eu{sup 3+}-doped Gd{sub 2}O{sub 3} are observed. First the linewidth of the Eu{sup 3+} transitions are inhomogeneously broadened at low temperature, and second the population of the ground and the first excited state of Eu{sup 3+} as a function of temperature differs depending on the size: at low temperatures, the smaller the particles, the less populated the first excited state is. The explanation of this second phenomenon could be the absence of very low-energy phonons in small particles.

  20. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Saba, Kazuhisa.

    1979-01-01

    Purpose: To improve the earthquake resistance as well as reduce the size of a container for a nuclear reactor with no adverse effects on the decrease of impact shock to the container and shortening of construction step. Constitution: Reinforcing profile steel materials are welded longitudinally and transversely to the inner surface of a container, and inner steel plates are secured to the above profile steel materials while keeping a gap between the materials and the container. Reactor shielding wall planted to the base concrete of the container is mounted to the pressure vessel, and main steam pipeways secured by the transverse beams and led to the outside of container is connected. This can improve the rigidity earthquake strength and the safetiness against the increase in the inside pressure upon failures of the container. (Yoshino, Y.)

  1. Effects of Confinement on Conventional Spin Problems

    DEFF Research Database (Denmark)

    Marchukov, Oleksandr

    2015-01-01

    , Hubbard model, etc., which are nevertheless used to describe physical phenomena in various fields, such as condensed matter physics, nuclear physics, etc. This dissertation discusses the effects of the external confinement on some con- ventional spin problems. It consists of two parts: In the first part...... the effects of spin-orbit coupling on particles trapped in a two-dimensional harmonic oscilla- tor are considered. The influences of the deformation of the trap, interparticle interaction and external magnetic field are analyzed. The statistical analysis of the single-particle energy spectrum and its relation...

  2. CLASSIFICATION OF SYSTEMS FOR PASSIVE AFTERHEAT REMOVAL FROM REACTOR CONTAINMENT OF NUCLEAR POWER PLANT WITH WATER-COOLED POWER REACTOR

    Directory of Open Access Journals (Sweden)

    N. Khaled

    2014-01-01

    Full Text Available A classification on systems for passive afterheat removal from reactor containment has been developed in the paper.  The classification permits to make a detailed analysis of various concepts pertaining to systems for passive afterheat removal from reactor containment of new generation. The paper considers main classification features of the given systems.

  3. Theoretical studies in nuclear reactions and nuclear structure

    International Nuclear Information System (INIS)

    1992-05-01

    Research in the Maryland Nuclear Theory Group focusses on problems in four basic areas of current relevance. Hadrons in nuclear matter; the structure of hadrons; relativistic nuclear physics and heavy ion dynamics and related processes. The section on hadrons in nuclear matter groups together research items which are aimed at exploring ways in which the properties of nucleons and the mesons which play a role in the nuclear force are modified in the nuclear medium. A very interesting result has been the finding that QCD sum rules supply a new insight into the decrease of the nucleon's mass in the nuclear medium. The quark condensate, which characterizes spontaneous chiral symmetry breaking of the late QCD vacuum, decreases in nuclear matter and this is responsible for the decrease of the nucleon's mass. The section on the structure of hadrons contains progress reports on our research aimed at understanding the structure of the nucleon. Widely different approaches are being studied, e.g., lattice gauge calculations, QCD sum rules, quark-meson models with confinement and other hedgehog models. A major goal of this type of research is to develop appropriate links between nuclear physics and QCD. The section on relativistic nuclear physics represents our continuing interest in developing an appropriate relativistic framework for nuclear dynamics. A Lorentz-invariant description of the nuclear force suggests a similar decrease of the nucleon's mass in the nuclear medium as has been found from QCD sum rules. Work in progress extends previous successes in elastic scattering to inelastic scattering of protons by nuclei. The section on heavy ion dynamics and related processes reports on research into the e + e - problem and heavy ion dynamics

  4. ORF73 LANA homologs of RRV and MneRV2 contain an extended RGG/RG-rich nuclear and nucleolar localization signal that interacts directly with importin β1 for non-classical nuclear import.

    Science.gov (United States)

    Howard, Kellie; Cherezova, Lidia; DeMaster, Laura K; Rose, Timothy M

    2017-11-01

    The latency-associated nuclear antigens (LANA) of KSHV and macaque RFHVMn, members of the RV1 rhadinovirus lineage, are closely related with conservation of complex nuclear localization signals (NLS) containing bipartite KR-rich motifs and RG-rich domains, which interact distinctly with importins α and ß1 for nuclear import via classical and non-classical pathways, respectively. RV1 LANAs are expressed in the nucleus of latently-infected cells where they inhibit replication and establish a dominant RV1 latency. Here we show that LANA homologs of macaque RRV and MneRV2 from the more distantly-related RV2 lineage, lack the KR-rich NLS, and instead have a large RG-rich NLS with multiple RG dipeptides and a conserved RGG motif. The RG-NLS interacts uniquely with importin β1, which mediates nuclear import and accumulation of RV2 LANA in the nucleolus. The alternative nuclear import and localization of RV2 LANA homologs may contribute to the dominant RV2 lytic replication phenotype. Copyright © 2017. Published by Elsevier Inc.

  5. Understanding the Risk of Chloride Induced Stress Corrosion Cracking of Interim Storage Containers for the Dry Storage of Spent Nuclear Fuel: Evolution of Brine Chemistry on the Container Surface

    International Nuclear Information System (INIS)

    Enos, David; Bryan, Charles R.

    2015-01-01

    Although the susceptibility of austenitic stainless steels to chloride-induced stress corrosion cracking is well known, uncertainties exist in terms of the environmental conditions that exist on the surface of the storage containers. While a diversity of salts is present in atmospheric aerosols, many of these are not stable when placed onto a heated surface. Given that the surface temperature of any container storing spent nuclear fuel will be well above ambient, it is likely that salts deposited on its surface may decompose or degas. To characterize this effect, relevant single and multi-salt mixtures are being evaluated as a function of temperature and relative humidity to establish the rates of degassing, as well as the likely final salt and brine chemistries that will remain on the canister surface.

  6. Understanding the Risk of Chloride Induced Stress Corrosion Cracking of Interim Storage Containers for the Dry Storage of Spent Nuclear Fuel: Evolution of Brine Chemistry on the Container Surface.

    Energy Technology Data Exchange (ETDEWEB)

    Enos, David; Bryan, Charles R.

    2015-10-01

    Although the susceptibility of austenitic stainless steels to chloride-induced stress corrosion cracking is well known, uncertainties exist in terms of the environmental conditions that exist on the surface of the storage containers. While a diversity of salts is present in atmospheric aerosols, many of these are not stable when placed onto a heated surface. Given that the surface temperature of any container storing spent nuclear fuel will be well above ambient, it is likely that salts deposited on its surface may decompose or degas. To characterize this effect, relevant single and multi-salt mixtures are being evaluated as a function of temperature and relative humidity to establish the rates of degassing, as well as the likely final salt and brine chemistries that will remain on the canister surface.

  7. West European magnetic confinement fusion research

    International Nuclear Information System (INIS)

    McKenney, B.L.; McGrain, M.; Hogan, J.T.; Porkolab, M.; Thomassen, K.I.

    1990-01-01

    This report presents a technical assessment and review of the West European program in magnetic confinement fusion by a panel of US scientists and engineers active in fusion research. Findings are based on the scientific and technical literature, on laboratory reports and preprints, and on the personal experiences and collaborations of the panel members. Concerned primarily with developments during the past 10 years, from 1979 to 1989, the report assesses West European fusion research in seven technical areas: tokamak experiments; magnetic confinement technology and engineering; fusion nuclear technology; alternate concepts; theory; fusion computations; and program organization. The main conclusion emerging from the analysis is that West European fusion research has attained a position of leadership in the international fusion program. This distinction reflects in large measure the remarkable achievements of the Joint European Torus (JET). However, West European fusion prominence extends beyond tokamak experimental physics: the program has demonstrated a breadth of skill in fusion science and technology that is not excelled in the international effort. It is expected that the West European primacy in central areas of confinement physics will be maintained or even increased during the early 1990s. The program's maturity and commitment kindle expectations of dramatic West European advances toward the fusion energy goal. For example, achievement of fusion breakeven is expected first in JET, before 1995

  8. The use of contained nuclear explosions to create underground reservoirs, and experience of operating these for gas condensate storage

    International Nuclear Information System (INIS)

    Kedrovskij, O.L.; Myasnikov, K.V.; Leonov, E.A.; Romadin, N.M.; Dorodnov, V.F.; Nikiforov, G.A.

    1975-01-01

    Investigations on the creation of underground reservoirs by means of nuclear explosions have been going on in the Soviet Union for many years. In this paper the authors consider three main kinds of sites or formations that can be used for constructing reservoirs by this method, namely, low-permeable rocks, worked-out mines and rock salt formations. Formulae are given for predicting the mechanical effect of an explosion in rocks, taking their strength characteristics into account. Engineering procedures are described for sealing and restoring the emplacement holes, so that they can be used for operating the underground reservoir. Experience with the contruction and operation of a 50 000 m 3 gas-condensate reservoir in a rock salt formation is described. In the appendix to the paper a method is presented for calculating the stability of spherical cavities created by nuclear explosions in rock salt, allowing for the development of elasto-plastic deformations and creep

  9. Thermostating highly confined fluids.

    Science.gov (United States)

    Bernardi, Stefano; Todd, B D; Searles, Debra J

    2010-06-28

    In this work we show how different use of thermostating devices and modeling of walls influence the mechanical and dynamical properties of confined nanofluids. We consider a two dimensional fluid undergoing Couette flow using nonequilibrium molecular dynamics simulations. Because the system is highly inhomogeneous, the density shows strong fluctuations across the channel. We compare the dynamics produced by applying a thermostating device directly to the fluid with that obtained when the wall is thermostated, considering also the effects of using rigid walls. This comparison involves an analysis of the chaoticity of the fluid and evaluation of mechanical properties across the channel. We look at two thermostating devices with either rigid or vibrating atomic walls and compare them with a system only thermostated by conduction through vibrating atomic walls. Sensitive changes are observed in the xy component of the pressure tensor, streaming velocity, and density across the pore and the Lyapunov localization of the fluid. We also find that the fluid slip can be significantly reduced by rigid walls. Our results suggest caution in interpreting the results of systems in which fluid atoms are thermostated and/or wall atoms are constrained to be rigid, such as, for example, water inside carbon nanotubes.

  10. 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)

  11. System of large transport containers for waste from dismantling light water and gas-cooled nuclear reactors. Volume 2

    International Nuclear Information System (INIS)

    Price, M.S.T.; Lafontaine, I.

    1985-01-01

    The purpose of this volume is to assess the means of transportation of decommissioning wastes, costs of transport, radiological detriment attributable to transport and develops conceptual designs of large transport containers. The document ends with Conclusions and Recommendations

  12. On the buckling of elastic rings by external confinement.

    Science.gov (United States)

    Hazel, Andrew L; Mullin, Tom

    2017-05-13

    We report the results of an experimental and numerical investigation into the buckling of thin elastic rings confined within containers of circular or regular polygonal cross section. The rings float on the surface of water held in the container and controlled removal of the fluid increases the confinement of the ring. The increased compressive forces can cause the ring to buckle into a variety of shapes. For the circular container, finite perturbations are required to induce buckling, whereas in polygonal containers the buckling occurs through a linear instability that is closely related to the canonical Euler column buckling. A model based on Kirchhoff-Love beam theory is developed and solved numerically, showing good agreement with the experiments and revealing that in polygons increasing the number of sides means that buckling occurs at reduced levels of confinement.This article is part of the themed issue 'Patterning through instabilities in complex media: theory and applications.' © 2017 The Author(s).

  13. Radio Frequency (RF) Trap for Confinement of Antimatter Plasmas Using Rotating Wall Electric Fields

    Science.gov (United States)

    Sims, William Herbert, III; Pearson, J. Boise

    2004-01-01

    Perturbations associated with a rotating wall electric field enable the confinement of ions for periods approaching weeks. This steady state confinement is a result of a radio frequency manipulation of the ions. Using state-of-the-art techniques it is shown that radio frequency energy can produce useable manipulation of the ion cloud (matter or antimatter) for use in containment experiments. The current research focuses on the improvement of confinement systems capable of containing and transporting antimatter.

  14. A nuclear criticality safety assessment of the loss of moderation control in 2 1/2 and 10-ton cylinders containing enriched UF{sub 6}

    Energy Technology Data Exchange (ETDEWEB)

    Newvahner, R.L. [Martin Marietta Energy Systems, Inc., Piketon, OH (United States); Pryor, W.A. [PAI Corp., Oak Ridge, TN (United States)

    1991-12-31

    Moderation control for maintaining nuclear criticality safety in 2 {1/2}-ton, 10-ton, and 14-ton cylinders containing enriched uranium hexafluoride (UF{sub 6}) has been used safely within the nuclear industry for over thirty years, and is dependent on cylinder integrity and containment. This assessment evaluates the loss of moderation control by the breaching of containment and entry of water into the cylinders. The first objective of this study was to estimate the required amounts of water entering these large UF{sub 6} cylinders to react with, and to moderate the uranium compounds sufficiently to cause criticality. Hypothetical accident situations were modeled as a uranyl fluoride (UO{sub 2}F{sub 2}) slab above a UF{sub 6} hemicylinder, and a UO{sub 2}F{sub 2} sphere centered within a UF{sub 6} hemicylinder. These situations were investigated by computational analyses utilizing the KENO V.a Monte Carlo Computer Code. The results were used to estimate both the masses of water required for criticality, and the limiting masses of water that could be considered safe. The second objective of the assessment was to calculate the time available for emergency control actions before a criticality would occur, i.e., a {open_quotes}safetime{close_quotes}, for various sources of water and different size openings in a breached cylinder. In the situations considered, except the case for a fire hose, the safetime appears adequate for emergency control actions. The assessment shows that current practices for handling moderation controlled cylinders of low enriched UF{sub 6}, along with the continuation of established personnel training programs, ensure nuclear criticality safety for routine and emergency operations.

  15. BY FRUSTUM CONFINING VESSEL

    Directory of Open Access Journals (Sweden)

    Javad Khazaei

    2016-09-01

    Full Text Available Helical piles are environmentally friendly and economical deep foundations that, due to environmental considerations, are excellent additions to a variety of deep foundation alternatives available to the practitioner. Helical piles performance depends on soil properties, the pile geometry and soil-pile interaction. Helical piles can be a proper alternative in sensitive environmental sites if their bearing capacity is sufficient to support applied loads. The failure capacity of helical piles in this study was measured via an experimental research program that was carried out by Frustum Confining Vessel (FCV. FCV is a frustum chamber by approximately linear increase in vertical and lateral stresses along depth from top to bottom. Due to special geometry and applied bottom pressure, this apparatus is a proper choice to test small model piles which can simulate field stress conditions. Small scale helical piles are made with either single helix or more helixes and installed in fine grained sand with three various densities. Axial loading tests including compression and tension tests were performed to achieve pile ultimate capacity. The results indicate the helical piles behavior depends essentially on pile geometric characteristics, i.e. helix configuration and soil properties. According to the achievements, axial uplift capacity of helical model piles is about equal to usual steel model piles that have the helixes diameter. Helical pile compression bearing capacity is too sufficient to act as a medium pile, thus it can be substituted other piles in special geoenvironmental conditions. The bearing capacity also depends on spacing ratio, S/D, and helixes diameter.

  16. Porter with nuclear reservations

    International Nuclear Information System (INIS)

    Patterson, W.

    1978-01-01

    The Ontario Royal Commission on Electric Power Planning, chaired by Arthur Porter is reviewed. This interim report on nuclear power is relevant to nuclear power planning far beyond the confines of Ontario and discusses nuclear issues in the explicit context of electricity generation and use. Problems considered in the report include safety aspects of uranium mining, milling, and spent fuel disposal, the economic issues which affect nuclear planning and nuclear industry, and the proliferation issue. (U.K.)

  17. The H1 histone-specific proteinase is associated with nuclear matrix and stimulated by DNA containing breaks of denatured sites

    International Nuclear Information System (INIS)

    Gaziev, A.I.; Kutsyj, M.P.

    1988-01-01

    Discovery of proteinase in nuclear matrix specific of H1 histone and dependent presence of breaks or denatured sites in DNA permits to assume that the given enzyme, obviously, participates in replication and DNA repair, in regulation of genes expression. Removal of H1 histone by proteinase is, probably, necessary for procedure of these processes, and, obviously, this proteinase suffers conformational changes in the composition of the DNA-histone complex. H1 histone disintegration in nucleohistone containing damaged sites of DNA by specific proteinase, probably, represents one of the mechanisms for providing DNA repair in cells of higher organisms

  18. Technical update on pressure suppression type containments in use in U.S. light water reactor nuclear power plants

    International Nuclear Information System (INIS)

    1978-07-01

    In 1972, Dr. S. H. Hanauer (Technical Advisor to the NRC's Executive Director for Operations) wrote a memorandum that raised several questions on the viability of pressure suppression containment concepts. The concerns raised by Dr. Hanauer have recently become the subject of considerable discussion by several members of the U.S. Congress and public. The report provides a response to these expressed concerns and a status summary for various technical matters that relate to the safety of pressure suppression type containments for light water cooled reactor plants

  19. Evaluation of internal occupational exposure of workers from nuclear medicine services by aerosol analysis containing 131I

    International Nuclear Information System (INIS)

    Carneiro, Luana Gomes; Sampaio, Camilla da Silva; Dantas, Ana Leticia Almeida; Lucena, Eder Augusto; Santos, Maristela Souza; Dantas, Bernardo Maranhao; Paula, Gustavo Affonso de

    2014-01-01

    This study evaluated the risk of internal occupational exposure associated with the incorporation of 131 I via inhalation, in Nuclear Medicine Services, using aerosol analysis techniques. Occupationally Exposed Individuals (IOE) involved in handling this radionuclide are subject to chronic exposure, which can lead to an increase in the committed effective dose. Results obtained in preliminary studies indicate the occurrence of incorporation of 131 I by workers involved in handling solutions for radioiodine therapy procedures. The evaluation was carried out in radiopharmacy lab (nuclear medicine service) of a public hospital located in the city of Rio de Janeiro. After confirmed the presence of the radioisotope, by a qualitative assessment, it was determined an experimental arrangement for sample collection and were detected and quantitated the presence of steam 131 I during routine work. The average concentration of activity obtained in this study was 3 Bq / m 3 . This value is below of Derived Concentration in Air (DCA) of 8.4 x 10 3 Bq of 131 I / m 3 corresponding to a committed effective dose of 1.76 x 10 -4 mSv. These results demonstrate that the studied area is safe in terms of internal exposure of workers. However, the presence of 131 I should be periodically reevaluated, since this type of exposure contributes to the increase of the individual effective doses. Based on the data obtained improvements were suggested in the exhaust system and the use of good work practices in order to optimize the exposures

  20. Magnetic-confinement fusion

    Science.gov (United States)

    Ongena, J.; Koch, R.; Wolf, R.; Zohm, H.

    2016-05-01

    Our modern society requires environmentally friendly solutions for energy production. Energy can be released not only from the fission of heavy nuclei but also from the fusion of light nuclei. Nuclear fusion is an important option for a clean and safe solution for our long-term energy needs. The extremely high temperatures required for the fusion reaction are routinely realized in several magnetic-fusion machines. Since the early 1990s, up to 16 MW of fusion power has been released in pulses of a few seconds, corresponding to a power multiplication close to break-even. Our understanding of the very complex behaviour of a magnetized plasma at temperatures between 150 and 200 million °C surrounded by cold walls has also advanced substantially. This steady progress has resulted in the construction of ITER, a fusion device with a planned fusion power output of 500 MW in pulses of 400 s. ITER should provide answers to remaining important questions on the integration of physics and technology, through a full-size demonstration of a tenfold power multiplication, and on nuclear safety aspects. Here we review the basic physics underlying magnetic fusion: past achievements, present efforts and the prospects for future production of electrical energy. We also discuss questions related to the safety, waste management and decommissioning of a future fusion power plant.

  1. Alternative approaches to plasma confinement

    Science.gov (United States)

    Roth, J. R.

    1978-01-01

    The paper discusses 20 plasma confinement schemes each representing an alternative to the tokamak fusion reactor. Attention is given to: (1) tokamak-like devices (TORMAC, Topolotron, and the Extrap concept), (2) stellarator-like devices (Torsatron and twisted-coil stellarators), (3) mirror machines (Astron and reversed-field devices, the 2XII B experiment, laser-heated solenoids, the LITE experiment, the Kaktus-Surmac concept), (4) bumpy tori (hot electron bumpy torus, toroidal minimum-B configurations), (5) electrostatically assisted confinement (electrostatically stuffed cusps and mirrors, electrostatically assisted toroidal confinement), (6) the Migma concept, and (7) wall-confined plasmas. The plasma parameters of the devices are presented and the advantages and disadvantages of each are listed.

  2. Assessing confinement in coastal lagoons.

    Science.gov (United States)

    Canu, Donata Melaku; Solidoro, Cosimo; Umgiesser, Georg; Cucco, Andrea; Ferrarin, Christian

    2012-11-01

    Measures of transport scale in aquatic systems can contribute to the formulation of definitions of indicators of the system's ecological properties. This paper addresses confinement, a specific transport scale proposed by biological scientists as a parameter that can capture and synthesize the principal properties that determine the spatial structure of biological communities in transitional environments. Currently, there is no direct experimental measure of confinement. In this study, a methodology based on the accumulation rate within a lagoon of a passive tracer of marine origin is proposed, the influences of different factors in the calculation of confinement are analyzed, and general recommendations are derived. In particular, we analyze the spatial and the temporal variability of confinement and its sensitivity to the seasonal variability of climatic forcing, the inputs from rivers and the parameterization of the tidal exchanges. The Lagoon of Venice is used as a case study. Copyright © 2012 Elsevier Ltd. All rights reserved.

  3. Impurities confined in quantum structures

    CERN Document Server

    Holtz, Per Olof

    2004-01-01

    The introduction of impurities, even in very small concentrations, in a semiconductor can change its optical and electrical properties entirely. This attribute of the semiconductor is utilized in the manifoldness of their applications. In this book, the progress on elucidating the physical properties of impurities confined in quantum structures are reviewed with an emphasis on the experimental aspects. The major results of various kinds of characterization, such as infrared spectroscopy, Raman measurements, luminescence characterization, perturbation spectroscopy and dynamical studies of the confined impurities are reviewed, but also the theoretical basis to calculate the electronic structure of the confined donors and acceptors are presented. This monograph also describes more specific aspects of the confined impurities such as the properties in the high doping regime and the effects of hydrogen passivation.

  4. Ensuring preservation of nuclear materials at RT plant storage facilities by upgrading physical protection, containment (TID) and TV surveillance measures

    International Nuclear Information System (INIS)

    Savin, A.A.

    1999-01-01

    The upgrading of nuclear material (NM) physical protection system of storage facilities for plutonium dioxide has been carried out in the frame of contract between Sandia National Laboratories and PA Mayak. The implementation of this project is of great importance for ensuring the preservation of NM in storage. The project envisages a complex approach to the solution of the task: the use of engineered fences, detection means, observation devices, access and control thereof for the personnel; the organization of security force in such way that to give a possibility to take adequate and purposeful actions towards possible adversaries. The design includes upgrading the system for physical protection through the introduction of modern, efficient engineering devices for security, integrated system for access control and security, the system for TV surveillance [ru

  5. Radiation heat transfer calculations for the uranium fuel-containment region of the nuclear light bulb engine.

    Science.gov (United States)

    Rodgers, R. J.; Latham, T. S.; Krascella, N. L.

    1971-01-01

    Calculation results are reviewed of the radiant heat transfer characteristics in the fuel and buffer gas regions of a nuclear light bulb engine based on the transfer of energy by thermal radiation from gaseous uranium fuel in a neon vortex, through an internally cooled transparent wall, to seeded hydrogen propellant. The results indicate that the fraction of UV energy incident on the transparent walls increases with increasing power level. For the reference engine power level of 4600 megw, it is necessary to employ space radiators to reject the UV radiated energy absorbed by the transparent walls. This UV energy can be blocked by employing nitric oxide and oxygen seed gases in the fuel and buffer gas regions. However, this results in increased UV absorption in the buffer gas which also requires space radiators to reject the heat load.

  6. Effects of steam generator sectioning on the reliability of a nuclear power station containing a fast reactor

    International Nuclear Information System (INIS)

    Klemin, A.I.; Frolov, E.V.; Samoilov, O.B.

    1984-01-01

    Adequate reliability in the individual units and in the station as a whole must be provided in order to meet current specifications for the economic performance and safety of nuclear power stations. In the case of sodium-cooled fast reactors, a current problem is to improve the reliability in the steam generators of sodium-water type. Various Soviet and foreign designs for such reactors use sectional steam generators, which raise the reliability and safety of the station because they localize possible faults caused by leaks in the SG heatexchanger tubes to a single steam-generating section. The SG sectioning makes it necessary to incorporate factors into the reliability model such as the steam generator repair strategy, the number of steamgenerating sections in one cooling loop, and the working constraints on the available unit power. The authors present a model that has been used to examine the effects of these factors on the reliability of a fast-reactor system

  7. Nuclear magnetic resonance studies on bacterial dihydrofolate reductase containing (guanidino-/sup 13/C)arginine. [Streptococcus

    Energy Technology Data Exchange (ETDEWEB)

    Cocco, L.; Blakley, R.L.; Walker, T.E.; London, R.E.; Matwiyoff, N.A.

    1978-10-03

    Dihydrofolate reductase labeled with (guanidino-/sup 13/C)arginine has been purified from Streptococcus faecium and /sup 13/C nuclear magnetic resonance spectra of the enzyme and its complexes with various ligands have been recorded. Resonances of the eight residues are resolved into 4 to 6 peaks with chemical shifts over a range of 1.2 ppM. There appear to be two classes of residues: those with chemical shifts very close to that of free (guanidino-/sup 13/C)arginine (class 1); and those with significantly different shifts (class 2). Spin-lattice relaxation times (T/sub 1/), measured in H/sub 2/O, for residues of class 1 are approximately 50% greater than the values for residues of the second class. In D/sub 2/O the T/sub 1/ values for both classes of residues are essentially the same and approximately twice the values obtained in H/sub 2/O for residues of class 1. The temperature-dependent behavior of T/sub 1/ for residues of class 2, together with the small nuclear Overhauser enhancement values, and the difference in line width in H/sub 2/O vs. D/sub 2/O are consistent with the assumption that the internal motion of these residues is slow relative to the overall rotational motion of the protein. An overall rotational correlation time for the protein of 20 ns has been estimated from the data for these immobilized residues. Class 1 residues appear to have a significant degree of internal motion and are probably accessible to solvent, whereas class 2 residues are probably inaccessible.

  8. Infrared slavery and quark confinement

    CERN Document Server

    Alabiso, C

    1976-01-01

    The question is considered of whether the so-called infrared slavery mechanism as, e.g., being manifest in non-Abelian gauge theories, necessarily confines quarks. Making a specific ansatz for the long- range forces, the Schwinger-Dyson equation is solved for the quark Green function. Besides having a confining solution, it appears that quarks may by-pass the long-range forces and be produced. (20 refs).

  9. Overexpression of a soybean nuclear localized type-III DnaJ domain-containing HSP40 reveals its roles in cell death and disease resistance.

    Science.gov (United States)

    Liu, Jian-Zhong; Whitham, Steven A

    2013-04-01

    Heat-shock proteins such as HSP70 and HSP90 are important molecular chaperones that play critical roles in biotic and abiotic stress responses; however, the involvement of their co-chaperones in stress biology remains largely uninvestigated. In a screen for candidate genes stimulating cell death in Glycine max (soybean), we transiently overexpressed full-length cDNAs of soybean genes that are highly induced during soybean rust infection in Nicotiana benthamiana leaves. Overexpression of a type-III DnaJ domain-containing HSP40 (GmHSP40.1), a co-chaperone of HSP70, caused hypersensitive response (HR)-like cell death. The HR-like cell death was dependent on MAPKKKα and WIPK, because silencing each of these genes suppressed the HR. Consistent with the presence of a nuclear localization signal (NLS) motif within the GmHSP40.1 coding sequence, GFP-GmHSP40.1 was exclusively present in nuclear bodies or speckles. Nuclear localization of GmHSP40.1 was necessary for its function, because deletion of the NLS or addition of a nuclear export signal abolished its HR-inducing ability. GmHSP40.1 co-localized with HcRed-SE, a protein involved in pri-miRNA processing, which has been shown to be co-localized with SR33-YFP, a protein involved in pre-mRNA splicing, suggesting a possible role for GmHSP40.1 in mRNA splicing or miRNA processing, and a link between these processes and cell death. Silencing GmHSP40.1 enhanced the susceptibility of soybean plants to Soybean mosaic virus, confirming its positive role in pathogen defense. Together, the results demonstrate a critical role of a nuclear-localized DnaJ domain-containing GmHSP40.1 in cell death and disease resistance in soybean. © 2013 The Authors The Plant Journal © 2013 Blackwell Publishing Ltd.

  10. The use of repassivation potential in predicting the performance of high-level nuclear waste container materials

    International Nuclear Information System (INIS)

    Sridhar, N.; Dunn, D.; Cragnolino, G.

    1995-01-01

    Localized corrosion in aqueous environments forms an important bounding condition for the performance assessment of high-level waste (HLW) container materials. A predictive methodology using repassivation potential is examined in this paper. It is shown, based on long-term (continuing for over 11 months) testing of alloy 825, that repassivation potential of deep pits or crevices is a conservative and robust parameter for the prediction of localized corrosion. In contrast, initiation potentials measured by short-term tests are non-conservative and highly sensitive to several surface and environmental factors. Corrosion data from various field tests and plant equipment performance are analyzed in terms of the applicability of repassivation potential. The applicability of repassivation potential for predicting the occurrence of stress corrosion cracking (SCC) and intergranular corrosion in chloride containing environments is also examined

  11. BEACON/MOD2A: a computer program for subcompartment analysis of nuclear reactor containment. A user's manual. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Wells, R. A.

    1979-03-01

    The BEACON code is a Best Estimate Advanced Containment code which being developed by EG and G, Idaho, Inc., at the Idaho National Engineering Laboratory. The program is designed to perform a best estimate analysis of the flow of a mixture of air, water, and steam in a nuclear reactor containment system under loss-of-coolant accident conditions. The code can simulate two-component, two-phase fluid flow in complex geometries using a combination of two-dimensional, one-dimensional, and lumped-parameter representations for the various parts of the system. The current version of BEACON, which is designated BEACON/MOD2A, contains mass and heat transfer models for wall film and for wall conduction. It is suitable for the evaluation of short term transients in PWR dry containment systems. This manual describes the models employed in BEACON/MOD2A and specifies code implementation requirements. It provides application information for input data preparation and for output data interpretation.

  12. Understanding long-term corrosion of Alloy 22 container in the potential Yucca Mountain repository for high-level nuclear waste disposal

    Science.gov (United States)

    Ahn, T.; Jung, H.; He, X.; Pensado, O.

    2008-09-01

    Alloy 22 (Ni-22Cr-13Mo-3W-4Fe) is the candidate material for the waste package outer container in a potential geologic repository for high-level nuclear waste disposal at Yucca Mountain, Nevada. This alloy exhibits very low corrosion rates in the absence of environmental conditions promoting crevice corrosion. However, there are uncertainties regarding Alloy 22's corrosion performance when general corrosion rates and susceptibility to crevice corrosion are extrapolated to a geological time period (e.g. 10 5 years). This paper presents an analysis of available literature information relevant to the long-term extrapolation of general corrosion processes and the crevice corrosion behavior of Alloy 22, under potential repository environments. For assessment of general corrosion rates, potential degradation processes causing the loss of the long-term persistence of passive film formed are considered. For crevice corrosion, induction time, and the extent of susceptibility and opening area, are considered. Disclaimer: The US Nuclear Regulatory Commission (NRC) staff views expressed herein are preliminary and do not constitute a final judgment or determination of the matters addressed nor of the acceptability of a license application for a geologic repository at Yucca Mountain. The paper describes work performed by the Center for Nuclear Waste Regulatory Analyses (CNWRA) for NRC under Contract Number NRC-02-02-012. The activities reported here were performed by CNWRA on behalf of the NRC office of Nuclear Material Safety and Safeguards, Division of High Level Waste Repository Safety. This paper is an independent product of the CNWRA and does not necessarily reflect the view or regulatory position of the NRC.

  13. Two-electron atomic systems confined within spheroidal boxes

    Energy Technology Data Exchange (ETDEWEB)

    Corella-Madueno, A.; Rosas, R.A.; Marin, J.L.; Riera, R.

    2000-03-15

    The direct variational method is used to estimate some interesting physical properties of the He atom and the Li{sup +} ion confined within impenetrable spheroidal boxes. A comparative investigation f the ground=state energy, pressure, polarizability, dipole, and quadrupole moments with those of the He atom inside boxes with paraboloidal walls is made. The overall results show a similar qualitative behavior. However, for Li{sup +} there are quantitative differences on such properties due to its major nuclear charge, as expected. The trial wave function is constructed as a product of two hydrogenic wave functions adapted to the geometry of the confining boxes.

  14. A Review of Quantum Confinement

    Science.gov (United States)

    Connerade, Jean-Patrick

    2009-12-01

    A succinct history of the Confined Atom problem is presented. The hydrogen atom confined to the centre of an impenetrable sphere counts amongst the exactly soluble problems of physics, alongside much more noted exact solutions such as Black Body Radiation and the free Hydrogen atom in absence of any radiation field. It shares with them the disadvantage of being an idealisation, while at the same time encapsulating in a simple way particular aspects of physical reality. The problem was first formulated by Sommerfeld and Welker [1]—henceforth cited as SW—in connection with the behaviour of atoms at very high pressures, and the solution was published on the occasion of Pauli's 60th birthday celebration. At the time, it seemed that there was not much other connection with physical reality beyond a few simple aspects connected to the properties of atoms in solids, for which more appropriate models were soon developed. Thus, confined atoms attracted little attention until the advent of the metallofullerene, which provided the first example of a confined atom with properties quite closely related to those originally considered by SW. Since then, the problem has received much more attention, and many more new features of quantum confinement, quantum compression, the quantum Faraday cage, electronic reorganisation, cavity resonances, etc have been described, which are relevant to real systems. Also, a number of other situations have been uncovered experimentally to which quantum confinement is relevant. Thus, studies of the confined atom are now more numerous, and have been extended both in terms of the models used and the systems to which they can be applied. Connections to thermodynamics are explored through the properties of a confined two-level atom adapted from Einstein's celebrated model, and issues of dynamical screening of electromagnetic radiation by the confining shell are discussed in connection with the Faraday cage produced by a confining conducting shell

  15. Assessment and management of ageing of major nuclear power plant components important to safety: Concrete containment buildings

    International Nuclear Information System (INIS)

    1998-06-01

    The report presents the results of the Co-ordinated Research Programme (CRP) on the Management of Ageing of Concrete Containment Buildings (CCBs) addressing current practices and techniques for assessing fitness-for-service and the inspection, monitoring and mitigation of ageing degradation of selected components of CANDU reactor, BWR reactor, PWR reactor and WWER plants. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age-related licensing issues

  16. Nuclear structure

    International Nuclear Information System (INIS)

    Eastham, D.A.; Joy, T.

    1986-01-01

    The paper on 'nuclear structure' is the Appendix to the Daresbury (United Kingdom) Annual Report 1985/86, and contains the research work carried out at the Nuclear Structure Facility, Daresbury, within that period. During the year a total of 74 experiments were scheduled covering the main areas of activity including: nuclear collective motion, nuclei far from stability, and nuclear collisions. The Appendix contains brief reports on these experiments and associated theory. (U.K.)

  17. Analysis of dust samples collected from spent nuclear fuel interim storage containers at Hope Creek, Delaware, and Diablo Canyon, California

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Enos, David George [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-07-01

    Potentially corrosive environments may form on the surface of spent nuclear fuel dry storage canisters by deliquescence of deposited dusts. To assess this, samples of dust were collected from in-service dry storage canisters at two near-marine sites, the Hope Creek and Diablo Canyon storage installations, and have been characterized with respect to mineralogy, chemistry, and texture. At both sites, terrestrially-derived silicate minerals, including quartz, feldspars, micas, and clays, comprise the largest fraction of the dust. Also significant at both sites were particles of iron and iron-chromium metal and oxides generated by the manufacturing process. Soluble salt phases were minor component of the Hope Creek dusts, and were compositionally similar to inland salt aerosols, rich in calcium, sulfate, and nitrate. At Diablo Canyon, however, sea-salt aerosols, occurring as aggregates of NaCl and Mg-sulfate, were a major component of the dust samples. The seasalt aerosols commonly occurred as hollow spheres, which may have formed by evaporation of suspended aerosol seawater droplets, possibly while rising through the heated annulus between the canister and the overpack. The differences in salt composition and abundance for the two sites are attributed to differences in proximity to the open ocean and wave action. The Diablo Canyon facility is on the shores of the Pacific Ocean, while the Hope Creek facility is on the shores of the Delaware River, several miles from the open ocean.

  18. The meganuclease I-SceI containing nuclear localization signal (NLS-I-SceI) efficiently mediated mammalian germline transgenesis via embryo cytoplasmic microinjection.

    Science.gov (United States)

    Wang, Yong; Zhou, Xiao-Yang; Xiang, Peng-Ying; Wang, Lu-Lu; Tang, Huan; Xie, Fei; Li, Liang; Wei, Hong

    2014-01-01

    The meganuclease I-SceI has been effectively used to facilitate transgenesis in fish eggs for nearly a decade. I-SceI-mediated transgenesis is simply via embryo cytoplasmic microinjection and only involves plasmid vectors containing I-SceI recognition sequences, therefore regarding the transgenesis process and application of resulted transgenic organisms, I-SceI-mediated transgenesis is of minimal bio-safety concerns. However, currently no transgenic mammals derived from I-SceI-mediated transgenesis have been reported. In this work, we found that the native I-SceI molecule was not capable of facilitating transgenesis in mammalian embryos via cytoplasmic microinjection as it did in fish eggs. In contrast, the I-SceI molecule containing mammalian nuclear localization signal (NLS-I-SceI) was shown to be capable of transferring DNA fragments from cytoplasm into nuclear in porcine embryos, and cytoplasmic microinjection with NLS-I-SceI mRNA and circular I-SceI recognition sequence-containing transgene plasmids resulted in transgene expression in both mouse and porcine embryos. Besides, transfer of the cytoplasmically microinjected mouse and porcine embryos into synchronized recipient females both efficiently resulted in transgenic founders with germline transmission competence. These results provided a novel method to facilitate mammalian transgenesis using I-SceI, and using the NLS-I-SceI molecule, a simple, efficient and species-neutral transgenesis technology based on embryo cytoplasmic microinjection with minimal bio-safety concerns can be established for mammalian species. As far as we know, this is the first report for transgenic mammals derived from I-SceI-mediated transgenesis via embryo cytoplasmic microinjection.

  19. Gas solubility in hydrophobic confinement.

    Science.gov (United States)

    Luzar, Alenka; Bratko, Dusan

    2005-12-01

    Measured forces between apolar surfaces in water have often been found to be sensitive to exposure to atmospheric gases despite low gas solubilities in bulk water. This raises questions as to how significant gas adsorption is in hydrophobic confinement, whether it is conducive to water depletion at such surfaces, and ultimately if it can facilitate the liquid-to-gas phase transition in the confinement. Open Ensemble molecular simulations have been used here to determine saturated concentrations of atmospheric gases in water-filled apolar confinements as a function of pore width at varied gas fugacities. For paraffin-like confinements of widths barely exceeding the mechanical instability threshold (spinodal) of the liquid-to-vapor transition of confined water (aqueous film thickness between three and four molecular diameters), mean gas concentrations in the pore were found to exceed the bulk values by a factor of approximately 30 or approximately 15 in cases of N2 and CO2, respectively. At ambient conditions, this does not result in visible changes in the water density profile next to the surfaces. Whereas the barrier to capillary evaporation has been found to decrease in the presence of dissolved gas (Leung, K.; Luzar, A.; and Bratko, D. Phys. Rev. Lett. 2003, 90, 065502), gas concentrations much higher than those observed at normal atmospheric conditions would be needed to produce noticeable changes in the kinetics of capillary evaporation. In simulations, dissolved gas concentrations corresponding to fugacities above approximately 40 bar for N2, or approximately 2 bar for CO2, were required to trigger expulsion of water from a hydrocarbon slit as narrow as 1.4 nm. For nanosized pore widths corresponding to the mechanical instability threshold or above, no significant coupling between adsorption layers at opposing confinement walls was observed. This finding explains the approximately linear increase in gas solubility with inverse confinement width and the

  20. Defense High Level Waste Disposal Container System Description Document

    International Nuclear Information System (INIS)

    Pettit, N. E.

    2001-01-01

    The Defense High Level Waste 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 and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms [IPWF]) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. US Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as co-disposal. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister inserted in the center and/or one or more DOE SNF canisters displacing a HLW canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by

  1. Sulphate solubility and sulphate diffusion in oxide glasses: implications for the containment of sulphate-bearing nuclear wastes

    International Nuclear Information System (INIS)

    Lenoir, M.

    2009-09-01

    The thesis deals with sulphate solubility and sulphate diffusion in oxide glasses, in order to control sulphate incorporation and sulphate volatilization in nuclear waste glasses. It was conducted on simplified compositions, in the SiO 2 -B 2 O 3 -R 2 O (R = Li, Na, K, Cs), SiO 2 -B 2 O 3 -BaO and V 2 O 5 -B 2 O 3 -BaO systems. These compositions allowed us to study the influence of the nature of network-modifying ions (Li + , Na + , K + , Cs + or Ba 2+ ) and also of former elements (Si, B, V), on structure and properties of glasses. Sulphate volatility is studied in sodium borosilicate melts using an innovative technique of sulphate quantitation with Raman spectroscopy. This technique is useful to obtain kinetic curves of sulphate volatilization. The establishment of a model to fit these curves leads to the determination of diffusion coefficients of sulphate. These diffusion coefficients can thus be compared to diffusion coefficients of other species, determined by other techniques and presented in the literature. They are also linked to diffusion coefficients in relation with the viscosity of the melts. Concerning sulphate solubility in glasses, it depends on glass composition and on the nature of sulphate incorporated. Sulphate incorporation in alkali borosilicate glasses leads to the formation of a sulphate layer floating on top of the melt. Sulphate incorporation in barium borosilicate and boro-vanadate glasses leads to the crystallization of sulphate species inside the vitreous matrix. Moreover, sulphate solubility is higher in these glasses than in alkali borosilicates. Finally, exchanges between cations present in glasses and cations present in the sulphate phase are also studied. (author)

  2. Nuclear light bulb

    Science.gov (United States)

    Latham, Tom

    1991-01-01

    The nuclear light bulb engine is a closed cycle concept. The nuclear light bulb concept provides containment by keeping the nuclear fuel fluid mechanically suspended in a cylindrical geometry. Thermal heat passes through an internally cooled, fused-silica, transparent wall and heats hydrogen propellant. The seeded hydrogen propellant absorbs radiant energy and is expanded through a nozzle. Internal moderation was used in the configuration which resulted in a reduced critical density requirement. This result was supported by criticality experiments. A reference engine was designed that had seven cells and was sized to fit in what was then predicted to be the shuttle bay mass and volume limitations. There were studies done of nozzle throat cooling schemes to remove the radiant heat. Elements of the nuclear light bulb program included closed loop critical assembly tests done at Los Alamos with UF6 confined by argon buffer gas. It was shown that the fuel region could be seeded with constituents that would block UV radiation from the uranium plasma. A combination of calculations and experiments showed that internal moderation produced a critical mass reduction. Other aspects of the research are presented.

  3. Analysis of mitigating measures during steam/hydrogen distributions in nuclear reactor containments with the 3D field code gasflow

    International Nuclear Information System (INIS)

    Royl, P.; Travis, J.R.; Haytcher, E.A.; Wilkening, H.

    1997-01-01

    This paper reports on the recent model additions to the 3D field code GASFLOW and on validation and application analyses for steam/hydrogen transport with inclusion of mitigation measures. The results of the 3D field simulation of the HDR test E11.2 are summarized. Results from scoping analyses that simulate different modes of CO2 inertization for conditions from the HDR test T31.5 are presented. The last part discusses different ways of recombiner modeling during 3D distribution simulations and gives the results from validation calculations for the HDR recombiner test E11.8.1 and the Battelle test MC3. The results demonstrate that field code simulations with computer codes like GASFLOW are feasible today for complex containment geometries and that they are necessary for a reliable prediction of hydrogen/steam distribution and mitigation effects. (author)

  4. Effects of surface chromium depletion on localized corrosion of alloy 825 as a high-level nuclear waste container material

    International Nuclear Information System (INIS)

    Dunn, D.S.; Sridhar, N.; Cragnolino, G.A.

    1995-01-01

    Effects of the chromium-depleted, mill-finished surface on the localized corrosion resistance of alloy 825 (UNS N08825) were investigated. Tests were conducted in solutions based on the ground water at Yucca mountain, Nevada, but with a higher concentration of chloride. Results indicated that breakdown (E p ) and repassivation (E rp ) potentials for mill-finished surfaces were more active than those for polished surfaces. Potentiodynamic polarization tests indicated pits could be initiated on the chromium-depleted surface at potentials of 220 mV SCE in a solution containing 1,000 ppm Cl - at 95 C. Potentiostatic tests identified a similar pit initiation potential for the mill-finished surface. However, under longterm potentiostatic tests, a higher potential of 300 mV SCE was needed to sustain stable pit growth beyond the chromium-depleted layer. An increase in surface roughness also was observed to decrease localized corrosion resistance of the material

  5. Multi-scale approach of the mechanical behaviour of reinforced concrete structures - Application to nuclear plant containment buildings

    International Nuclear Information System (INIS)

    David, M.

    2012-01-01

    This thesis develops a multi-scale strategy to describe the mechanical behaviour of steel reinforcements and prestressing tendons in a reinforced concrete structure. This strategy is declined in several steps, which allow gradual integration of new physical phenomena. The first asymptotic model represents the effective elastic behaviour of heterogeneities periodically distributed on a surface. It combines an elastic interface behaviour and a membrane behaviour. A second asymptotic model then focuses on the behaviour of rigid fibers distributed on a surface, which may slide with respect to the surrounding volume. These models induce less stress concentrations than the usual truss models. They are implemented in the finite element code Code-Aster, and validated with respect to reference three-dimensional simulations. Their interaction with a macroscopic crack is studied. Finally, this strategy allows the modeling of experimental tests carried out on a portion of a containment building in real scale. (author)

  6. Spherical microwave confinement and ball lightning

    Science.gov (United States)

    Robinson, William Richard

    This dissertation presents the results of research done on unconventional energy technologies from 1995 to 2009. The present civilization depends on an infrastructure that was constructed and is maintained almost entirely using concentrated fuels and ores, both of which will run out. Diffuse renewable energy sources rely on this same infrastructure, and hence face the same limitations. I first examined sonoluminescence directed toward fusion, but demonstrated theoretically that this is impossible. I next studied Low Energy Nuclear Reactions and developed methods for improving results, although these have not been implemented. In 2000, I began Spherical Microwave Confinement (SMC), which confines and heats plasma with microwaves in a spherical chamber. The reactor was designed and built to provide the data needed to investigate the possibility of achieving fusion conditions with microwave confinement. A second objective was to attempt to create ball lightning (BL). The reactor featured 20 magnetrons, which were driven by a capacitor bank and operated in a 0.2 s pulse mode at 2.45 GHz. These provided 20 kW to an icosahedral array of 20 antennas. Video of plasmas led to a redesign of the antennas to provide better coupling of the microwaves to the plasma. A second improvement was a grid at the base of the antennas, which provided corona electrons and an electric field to aid quick formation of plasmas. Although fusion conditions were never achieved and ball lightning not observed, experience gained from operating this basic, affordable system has been incorporated in a more sophisticated reactor design intended for future research. This would use magnets that were originally planned. The cusp geometry of the magnetic fields is suitable for electron cyclotron resonance in the same type of closed surface that in existing reactors has generated high-temperature plasmas. Should ball lightning be created, it could be a practical power source with nearly ideal

  7. Thermodynamics of confined gallium clusters.

    Science.gov (United States)

    Chandrachud, Prachi

    2015-11-11

    We report the results of ab initio molecular dynamics simulations of Ga13 and Ga17 clusters confined inside carbon nanotubes with different diameters. The cluster-tube interaction is simulated by the Lennard-Jones (LJ) potential. We discuss the geometries, the nature of the bonding and the thermodynamics under confinement. The geometries as well as the isomer spectra of both the clusters are significantly affected. The degree of confinement decides the dimensionality of the clusters. We observe that a number of low-energy isomers appear under moderate confinement while some isomers seen in the free space disappear. Our finite-temperature simulations bring out interesting aspects, namely that the heat capacity curve is flat, even though the ground state is symmetric. Such a flat nature indicates that the phase change is continuous. This effect is due to the restricted phase space available to the system. These observations are supported by the mean square displacement of individual atoms, which are significantly smaller than in free space. The nature of the bonding is found to be approximately jellium-like. Finally we note the relevance of the work to the problem of single file diffusion for the case of the highest confinement.

  8. New results on structure of low beta confinement Polywell cusps simulated by comsol multiphysics

    Directory of Open Access Journals (Sweden)

    B. Mahdavipour

    Full Text Available The Inertial electrostatic confinement (IEC is one of the ways for fusion approaches. It is one of the various methods which can be used to confine hot fusion plasma. The advantage of IEC is that the IEC experiments could be done in smaller size facilities than ITER or NIF, costing less money and moving forward faster. In IEC fusion, we need to trap adequate electrons to confine the desired ion density which is needed for a fusion reactor. Polywell is a device which uses the magnetic cusp system and traps the required amount of electrons for fusion reactions. The purpose of this device is to create a virtual cathode in order to achieve nuclear fusion using inertial electrostatic confinement (Miley and Krupakar Murali, 2014. In this paper, we have simulated the low beta Polywell. Then, we examined the effects of coil spacing, coils current, electron injection energy on confinement time. Keywords: Low beta confinement, Polywell, IEC, Comsol multiphysics

  9. High Explosive Detonation–Confiner Interactions

    Science.gov (United States)

    Short, Mark; Quirk, James J.

    2018-01-01

    The primary purpose of a detonation in a high explosive (HE) is to provide the energy to drive a surrounding confiner, typically for mining or munitions applications. The details of the interaction between an HE detonation and its confinement are essential to achieving the objectives of the explosive device. For the high pressures induced by detonation loading, both the solid HE and confiner materials will flow. The structure and speed of a propagating detonation, and ultimately the pressures generated in the reaction zone to drive the confiner, depend on the induced flow both within the confiner and along the HE–confiner material interface. The detonation–confiner interactions are heavily influenced by the material properties and, in some cases, the thickness of the confiner. This review discusses the use of oblique shock polar analysis as a means of characterizing the possible range of detonation–confiner interactions. Computations that reveal the fluid mechanics of HE detonation–confiner interactions for finite reaction-zone length detonations are discussed and compared with the polar analysis. This includes cases of supersonic confiner flow; subsonic, shock-driven confiner flow; subsonic, but shockless confiner flow; and sonic flow at the intersection of the detonation shock and confiner material interface. We also summarize recent developments, including the effects of geometry and porous material confinement, on detonation–confiner interactions.

  10. Experiments to investigate direct containment heating phenomena with scaled models of the Zion Nuclear Power Plant in the Surtsey Test Facility

    International Nuclear Information System (INIS)

    Allen, M.D.; Pilch, M.M.; Blanchat, T.K.; Griffith, R.O.; Nichols, R.T.

    1994-05-01

    The Surtsey Facility at Sandia National Laboratories (SNL) is used to perform scaled experiments that simulate hypothetical high-pressure melt ejection (HPME) accidents in a nuclear power plant (NPP). These experiments are designed to investigate the effect of specific phenomena associated with direct containment heating (DCH) on the containment load, such as the effect of physical scale, prototypic subcompartment structures, water in the cavity, and hydrogen generation and combustion. In the Integral Effects Test (IET) series, 1:10 linear scale models of the Zion NPP structures were constructed in the Surtsey vessel. The RPV was modeled with a steel pressure vessel that had a hemispherical bottom head, which had a 4-cm hole in the bottom head that simulated the final ablated hole that would be formed by ejection of an instrument guide tube in a severe NPP accident. Iron/alumina/chromium thermite was used to simulate molten corium that would accumulate on the bottom head of an actual RPV. The chemically reactive melt simulant was ejected by high-pressure steam from the RPV model into the scaled reactor cavity. Debris was then entrained through the instrument tunnel into the subcompartment structures and the upper dome of the simulated reactor containment building. The results of the IET experiments are given in this report

  11. Experiments to investigate direct containment heating phenomena with scaled models of the Zion Nuclear Power Plant in the Surtsey Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Allen, M.D.; Pilch, M.M.; Blanchat, T.K.; Griffith, R.O. [Sandia National Labs., Albuquerque, NM (United States); Nichols, R.T. [Ktech Corp., Albuquerque, NM (United States)

    1994-05-01

    The Surtsey Facility at Sandia National Laboratories (SNL) is used to perform scaled experiments that simulate hypothetical high-pressure melt ejection (HPME) accidents in a nuclear power plant (NPP). These experiments are designed to investigate the effect of specific phenomena associated with direct containment heating (DCH) on the containment load, such as the effect of physical scale, prototypic subcompartment structures, water in the cavity, and hydrogen generation and combustion. In the Integral Effects Test (IET) series, 1:10 linear scale models of the Zion NPP structures were constructed in the Surtsey vessel. The RPV was modeled with a steel pressure vessel that had a hemispherical bottom head, which had a 4-cm hole in the bottom head that simulated the final ablated hole that would be formed by ejection of an instrument guide tube in a severe NPP accident. Iron/alumina/chromium thermite was used to simulate molten corium that would accumulate on the bottom head of an actual RPV. The chemically reactive melt simulant was ejected by high-pressure steam from the RPV model into the scaled reactor cavity. Debris was then entrained through the instrument tunnel into the subcompartment structures and the upper dome of the simulated reactor containment building. The results of the IET experiments are given in this report.

  12. Topical Anti-Nuclear Factor-Kappa B Small Interfering RNA with Functional Peptides Containing Sericin-Based Hydrogel for Atopic Dermatitis

    Directory of Open Access Journals (Sweden)

    Takanori Kanazawa

    2015-09-01

    Full Text Available The small interfering RNA (siRNA is suggested to offer a novel means of treating atopic dermatitis (AD because it allows the specific silencing of genes related to AD pathogenesis. In our previous study, we found that siRNA targeted against RelA, an important nuclear factor-kappa B (NF-κB subdomain, with functional peptides, showed therapeutic effects in a mouse model of AD. In the present study, to develop a topical skin application against AD, we prepared a hydrogel containing anti-RelA siRNA and functional peptides and determined the intradermal permeation and the anti-AD effects in an AD mouse model. We selected the silk protein, sericin (SC, which is a versatile biocompatible biomaterial to prepare hydrogel as an aqueous gel base. We found that the siRNA was more widely delivered to the site of application in AD-induced ear skin of mice after topical application via the hydrogel containing functional peptides than via the preparation without functional peptides. In addition, the ear thickness and clinical skin severity of the AD-induced mice treated with hydrogel containing anti-RelA siRNA with functional peptides improved more than that of mice treated with the preparation formulated with negative siRNA.

  13. Confirmation test on confinement performance of improved glove box

    International Nuclear Information System (INIS)

    Miura, S.; Kanazawa, J.; Nakajima, M.; Sakuno, K.; Miyata, H.

    1995-01-01

    Glove boxes are used at nuclear facilities to confine radioactive materials by ensuring a high level of airtightness and negative internal pressure. The allowable rate of air leakage is 0.1% vol/hr or less at the pre-service inspection. The negative pressure value is normally maintained at about -30 mm H 2 O. Structural strength and confinement reliability of glove boxes during earthquake are major concerns, and most important glove boxes are designed to withstand seismic class A events is Japan. This paper describes vibration tests done to confirm that improve large-sized glove boxes maintain their confinement performance and structural strength even during earthquake and that the design analysis methods used are appropriate. (author). 1 ref., 6 figs., 3 tabs

  14. Thermodynamic properties of bulk and confined water

    Energy Technology Data Exchange (ETDEWEB)

    Mallamace, Francesco, E-mail: francesco.mallamace@unime.it [Dipartimento di Fisica e Scienza della Terra Università di Messina and CNISM, I-98168 Messina (Italy); Department of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Center for Polymer Studies and Department of Physics, Boston University, Boston, Massachusetts 02215 (United States); Corsaro, Carmelo [Dipartimento di Fisica e Scienza della Terra Università di Messina and CNISM, I-98168 Messina (Italy); Mallamace, Domenico [Dipartimento di Scienze dell' Ambiente, della Sicurezza, del Territorio, degli Alimenti e della Salute, Università di Messina, I-98166 Messina (Italy); Vasi, Sebastiano; Vasi, Cirino [IPCF-CNR, I-98166 Messina (Italy); Stanley, H. Eugene [Center for Polymer Studies and Department of Physics, Boston University, Boston, Massachusetts 02215 (United States)

    2014-11-14

    The thermodynamic response functions of water display anomalous behaviors. We study these anomalous behaviors in bulk and confined water. We use nuclear magnetic resonance (NMR) to examine the configurational specific heat and the transport parameters in both the thermal stable and the metastable supercooled phases. The data we obtain suggest that there is a behavior common to both phases: that the dynamics of water exhibit two singular temperatures belonging to the supercooled and the stable phase, respectively. One is the dynamic fragile-to-strong crossover temperature (T{sub L} ≃ 225 K). The second, T{sup *} ∼ 315 ± 5 K, is a special locus of the isothermal compressibility K{sub T}(T, P) and the thermal expansion coefficient α{sub P}(T, P) in the P–T plane. In the case of water confined inside a protein, we observe that these two temperatures mark, respectively, the onset of protein flexibility from its low temperature glass state (T{sub L}) and the onset of the unfolding process (T{sup *})

  15. Supplement analysis for a container system for the management of DOE spent nuclear fuel located at the INEEL

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-03-12

    The Council on Environmental Quality (CEQ) regulations for implementing the NEPA, 40 CFR 1502.9 (c), directs federal agencies to prepare a supplement to an environmental impact statement when an agency makes substantial changes in the Proposed Action that are relevant to environmental concerns, or there are significant new circumstances or information relevant to environmental concerns and bearing on the Proposed Action or impacts. When it is unclear whether a supplemental environmental impact statement is required, DOE regulations (10 CFR 1021.314) direct the preparation of a supplement analysis to assist in making that determination. This supplement analysis evaluates the impacts of employing dual-purpose canisters (DPCs) to prepare DOE SNF located at the INEEL for interim onsite storage and transport outside the State of Idaho. Impacts associated with DPC manufacturing, loading and storage of DOE-ID SNF into DPCs, transport of loaded DPCs outside Idaho, and the cumulative impacts are compared with the impacts previously analyzed in the SNF and INEL EIS and the Navy Container System EIS. This SA provides information to determine whether: (1) an existing EIS should be supplemented; (2) a new EIS should be prepared; or (3) no further NEPA documentation is required.

  16. Correlations In Confined Quantum Plasmas

    International Nuclear Information System (INIS)

    Dufty, J.W.

    2012-01-01

    This is the final report for the project 'Correlations in Confined Quantum Plasmas', NSF-DOE Partnership Grant DE FG02 07ER54946, 8/1/2007 - 7/30/2010. The research was performed in collaboration with a group at Christian Albrechts University (CAU), Kiel, Germany. That collaboration, almost 15 years old, was formalized during the past four years under this NSF-DOE Partnership Grant to support graduate students at the two institutions and to facilitate frequent exchange visits. The research was focused on exploring the frontiers of charged particle physics evolving from new experimental access to unusual states associated with confinement. Particular attention was paid to combined effects of quantum mechanics and confinement. A suite of analytical and numerical tools tailored to the specific inquiry has been developed and employed

  17. Developing a virtual reality application for training Nuclear Power Plant operators: Setting up a database containing dose rates in the refuelling plant

    International Nuclear Information System (INIS)

    Rodenas, J.; Zarza, I.; Burgos, M. C.; Felipe, A.; Sanchez-Mayoral, M. L.

    2004-01-01

    Operators in Nuclear Power Plants can receive high doses during refuelling operations. A training programme for simulating refuelling operations will be useful in reducing the doses received by workers as well as minimising operation time. With this goal in mind, a virtual reality application is developed within the framework of the CIPRES project. The application requires doses, both instantaneous and accumulated, to be displayed at all times during operator training. Therefore, it is necessary to set up a database containing dose rates at every point in the refuelling plant. This database is based on radiological protection surveillance data measured in the plant during refuelling operations. Some interpolation routines have been used to estimate doses through the refuelling plant. Different assumptions have been adopted in order to perform the interpolation and obtain consistent data. In this paper, the procedures developed to set up the dose database for the virtual reality application are presented and analysed. (authors)

  18. Structural analysis of mixed alkali borosilicate glasses containing Cs+ and Na+ using strong magnetic field magic angle spinning nuclear magnetic resonance

    Directory of Open Access Journals (Sweden)

    S. Kaneko

    2017-03-01

    Full Text Available We have investigated the local structure of alkali atoms in mixed alkali silicate, borate, and borosilicate glasses, which contain Cs+ and Na+, using strong magnetic field magic angle spinning nuclear magnetic resonance (MAS NMR spectroscopy of 133Cs and 23Na. The spectral peaks of 133Cs in borosilicate (Si:B = 1:1 and Si-rich borosilicate (Si:B = 2:1 glasses shifted to upfield with increasing Cs+/(Na+ + Cs+ ratio, which implies that the coordination number of Cs+ decreased as in the case of silicate and borate glasses. However, this trend was not observed in the 23Na spectra of either borosilicate glass. This might be because the chemical shift of 23Na in borosilicate glass is strongly affected by nearby species such as Si or B, and not by the coordination number of Na+.

  19. Conducting water chemistry of the secondary coolant circuit of VVER-based nuclear power plant units constructed without using copper containing alloys

    Science.gov (United States)

    Tyapkov, V. F.

    2014-07-01

    The secondary coolant circuit water chemistry with metering amines began to be put in use in Russia in 2005, and all nuclear power plant units equipped with VVER-1000 reactors have been shifted to operate with this water chemistry for the past seven years. Owing to the use of water chemistry with metering amines, the amount of products from corrosion of structural materials entering into the volume of steam generators has been reduced, and the flow-accelerated corrosion rate of pipelines and equipment has been slowed down. The article presents data on conducting water chemistry in nuclear power plant units with VVER-1000 reactors for the secondary coolant system equipment made without using copper-containing alloys. Statistical data are presented on conducting ammonia-morpholine and ammonia-ethanolamine water chemistries in new-generation operating power units with VVER-1000 reactors with an increased level of pH. The values of cooling water leaks in turbine condensers the tube system of which is made of stainless steel or titanium alloy are given.

  20. Hermitian relativity, chromodynamics and confinement

    International Nuclear Information System (INIS)

    Treder, H.J.

    1983-01-01

    The extension of the Riemann metrics of General Relativity to the complex domain (substitution of the symmetry conditions for the fundamental tensor, the affinity and the Ricci curvature by the conditions of hermicity) leads to a 'Generalized Theory of Gravity' (Einstein) describing the Newton-Einstein gravodynamics combined with the chromodynamics of quarks. The interaction of gravodynamics and chromodynamics implied by the Einstein-Schroedinger field equations of the hermitian relativity theory enforces the 'confinement'. The 'confinement' prevents the gravitational potential from divergence which would result in the lack of a Riemann space-time metric