WorldWideScience

Sample records for vessel loading operations

  1. EDF studies on PWR vessel internal loading

    International Nuclear Information System (INIS)

    Bellet, S.; Vallat, S.

    1998-01-01

    EDF has undertaken some mechanics and thermal-hydraulics studies with the objective of mastering plant phenomena today and in order to numerically predict the behaviour of vessel internals on units planned for the future. From some justifications already underway after in operation incidents (wear and drop time of RCCA rods, fuel deflection, adapter cracks, baffle bolt cracks) we intend to control reactor vessel flows and mechanical behaviour of internal structures. During normal operation, thermal-hydraulic is the main load of vessel internals. The current approach consists of acquiring the capacity to link different calculations, taking care that codes are qualified for physical phenomena and complex 3D geometries. For baffle assembly, a more simple model of this structure has been used to treat the physical phenomena linked to the LOCA transient. Results are encouraging mainly due to code capacity progression (resolution and models), which allows more and more complex physical phenomena to be treated, like turbulence flow and LOCA. (author)

  2. Vessel Operator System

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Operator cards are required for any operator of a charter/party boat and or a commercial vessel (including carrier and processor vessels) issued a vessel permit from...

  3. Method to moor an offshore operating vessel

    Energy Technology Data Exchange (ETDEWEB)

    Flory, J.F.

    1983-01-24

    A vessel such as a storage vessel is permanently moored, by means such as a yoke pivoted on the forecastle of the vessel, to a mooring leg, e.g. a riser or anchor chain, which is attached to a base located on the ocean floor. Mounted on the vessel is tension exsisting means, for example, counterweights, springs, winches, or the like, operably connected with the mooring leg for applying tension thereto such as by lifting the yoke. The top of the mooring leg is connected to the end of the yoke through a mooring swivel and a gimbaled mooring table or a universal joint. A fluid swivel may be located above the mooring table or about a load-carrying shaft connected to the mooring leg. 8 drawings.

  4. Lifting simulation of an offshore supply vessel considering various operating conditions

    Directory of Open Access Journals (Sweden)

    Dong-Hoon Jeong

    2016-06-01

    Full Text Available Recently, an offshore support vessel is being widely used to install an offshore structure such as a subsea equipment which is laid on its deck. The lifting operation which is one of the installation operations includes lifting off, lifting in the air, splash zone crossing, deep submerging, and finally landing of the structure with an offshore support vessel crane. There are some major considerations during this operation. Especially, when lifting off the structure, if operating conditions such as ocean environmental loads and hoisting (or lowering speed are bad, the excess of tension of wire ropes of the crane and the collision between the offshore support vessel and the structure can be occurred due to the relative motion between them. To solve this problem, this study performs the lifting simulation while the offshore support vessel installs the structure. The simulation includes the calculation of dynamic responses of the offshore support vessel and the equipment, including the wire tension and the collision detection. To check the applicability of the simulation, it is applied to some lifting steps by varying operating conditions. As a result, it is confirmed that the conditions affect the operability of those steps.

  5. Stress analysis in a non axisymmetric loaded reactor pressure vessel

    International Nuclear Information System (INIS)

    Albuquerque, Levi Barcelos; Assis, Gracia Menezes V. de; Miranda, Carlos Alexandre J.; Cruz, Julio Ricardo B.; Mattar Neto, Miguel

    1995-01-01

    In this work we intend to present the stress analysis of a PWR vessel under postulated concentrated loads. The vessel was modeled with Axisymmetric solid 4 nodes harmonic finite elements with the use of the ANSYS program, version 5.0. The bolts connecting the vessel flanges were modeled with beam elements. Some considerations were made to model the contact between the flanges. The perforated part of the vessel tori spherical head was modeled (with reduced properties due to its holes) to introduce its stiffness and loads but was not within the scope of this work. The loading consists of some usual ones, as pressure, dead weight, bolts preload, seismic load and some postulated ones as concentrated loads, over the vessel, modeled by Fourier Series. The results in the axisymmetric model are taken in terms of linearized stresses, obtained in some circumferential positions and for each position, in some sections along the vessel. Using the ASME Code (Section III, Division 1, Sub-section NB) the stresses are within the allowable limits. In order to draw some conclusions about stress linearization, the membrane plus bending stresses (Pl + Pb) are obtained and compared in some sections, using three different methods. (author)

  6. Structural analysis of the ITER vacuum vessel from disruption loading with halo asymmetry

    International Nuclear Information System (INIS)

    Riemer, B.W.; Sayer, R.O.

    1996-01-01

    Static structural analyses of the ITER vacuum vessel were performed with toroidally asymmetric disruption loads. Asymmetric halo current conditions were assumed to modify symmetric disruption loads which resulted in net lateral loading on the vacuum vessel torus. Structural analyses with the asymmetric loading indicated significantly higher vessel stress and blanket support forces than with symmetric disruption loads. A recent change in the vessel support design which provided toroidal constraints at each mid port was found to be effective in reducing torus lateral movement and vessel stress

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

    International Nuclear Information System (INIS)

    Stefanou, G.D.

    1982-01-01

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

  8. Nonlinear response of vessel walls due to short-time thermomechanical loading

    International Nuclear Information System (INIS)

    Pfeiffer, P.A.; Kulak, R.F.

    1994-01-01

    Maintaining structural integrity of the reactor pressure vessel (RPV) during a postulated core melt accident is an important safety consideration in the design of the vessel. This study addresses the failure predictions of the vessel due to thermal and pressure loadings fro the molten core debris depositing on the lower head of the vessel. Different loading combinations were considered based on the dead load, yield stress assumptions, material response and internal pressurization. The analyses considered only short term failure (quasi static) modes, long term failure modes were not considered. Short term failure modes include plastic instabilities of the structure and failure due to exceeding the failure strain. Long term failure odes would be caused by creep rupture that leads to plastic instability of the structure. Due to the sort time durations analyzed, creep was not considered in the analyses presented

  9. Multi-scenario evaluation and specification of electromagnetic loads on ITER vacuum vessel

    International Nuclear Information System (INIS)

    Rozov, Vladimir; Martinez, J.-M.; Portafaix, C.; Sannazzaro, G.

    2014-01-01

    Highlights: • We present the results of multi-scenario analysis of EM loads on ITER vacuum vessel (VV). • The differentiation of models provides the economic way to perform big amount of calculations. • Functional approximation is proposed for distributed data/FE/numerical results specification. • Examples of specification of the load profiles by trigonometric polynomials (DHT) are given. • Principles of accounting for toroidal asymmetry at EM interactions in tokamak are considered. - Abstract: The electro-magnetic (EM) transients cause mechanical forces, which represent one of the most critical loads for the ITER vacuum vessel (VV). The paper is focused on the results of multi-scenario analysis and systematization of these EM loads, including specifically addressed pressures on shells and the net vertical force. The proposed mathematical model and computational technology, based on the use of integral parameters and operational analysis methods, enabled qualitative and quantitative analysis of the problem, time-efficient computations and systematic assessment of a large number of scenarios. The obtained estimates, found envelopes and peak values exemplify the principal loads on the VV and provide a database to support engineering load specifications. Special attention is given to the challenge of specification and documenting of the results in a form, suitable for using the data in engineering applications. The practical aspects of specification of distributed data, such as experimental and finite-element (FE) results, by analytical interpolants are discussed. The example of functional approximation of the load profiles by trigonometric polynomials based on discrete Hartley transform (DHT) is given

  10. Multi-scenario evaluation and specification of electromagnetic loads on ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Rozov, Vladimir, E-mail: vladimir.rozov@iter.org; Martinez, J.-M.; Portafaix, C.; Sannazzaro, G.

    2014-10-15

    Highlights: • We present the results of multi-scenario analysis of EM loads on ITER vacuum vessel (VV). • The differentiation of models provides the economic way to perform big amount of calculations. • Functional approximation is proposed for distributed data/FE/numerical results specification. • Examples of specification of the load profiles by trigonometric polynomials (DHT) are given. • Principles of accounting for toroidal asymmetry at EM interactions in tokamak are considered. - Abstract: The electro-magnetic (EM) transients cause mechanical forces, which represent one of the most critical loads for the ITER vacuum vessel (VV). The paper is focused on the results of multi-scenario analysis and systematization of these EM loads, including specifically addressed pressures on shells and the net vertical force. The proposed mathematical model and computational technology, based on the use of integral parameters and operational analysis methods, enabled qualitative and quantitative analysis of the problem, time-efficient computations and systematic assessment of a large number of scenarios. The obtained estimates, found envelopes and peak values exemplify the principal loads on the VV and provide a database to support engineering load specifications. Special attention is given to the challenge of specification and documenting of the results in a form, suitable for using the data in engineering applications. The practical aspects of specification of distributed data, such as experimental and finite-element (FE) results, by analytical interpolants are discussed. The example of functional approximation of the load profiles by trigonometric polynomials based on discrete Hartley transform (DHT) is given.

  11. Structural analysis of the ITER Vacuum Vessel regarding 2012 ITER Project-Level Loads

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, J.-M., E-mail: jean-marc.martinez@live.fr [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul lez Durance (France); Jun, C.H.; Portafaix, C.; Choi, C.-H.; Ioki, K.; Sannazzaro, G.; Sborchia, C. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul lez Durance (France); Cambazar, M.; Corti, Ph.; Pinori, K.; Sfarni, S.; Tailhardat, O. [Assystem EOS, 117 rue Jacquard, L' Atrium, 84120 Pertuis (France); Borrelly, S. [Sogeti High Tech, RE2, 180 rue René Descartes, Le Millenium – Bat C, 13857 Aix en Provence (France); Albin, V.; Pelletier, N. [SOM Calcul – Groupe ORTEC, 121 ancien Chemin de Cassis – Immeuble Grand Pré, 13009 Marseille (France)

    2014-10-15

    Highlights: • ITER Vacuum Vessel is a part of the first barrier to confine the plasma. • ITER Vacuum Vessel as Nuclear Pressure Equipment (NPE) necessitates a third party organization authorized by the French nuclear regulator to assure design, fabrication, conformance testing and quality assurance, i.e. Agreed Notified Body (ANB). • A revision of the ITER Project-Level Load Specification was implemented in April 2012. • ITER Vacuum Vessel Loads (seismic, pressure, thermal and electromagnetic loads) were summarized. • ITER Vacuum Vessel Structural Margins with regards to RCC-MR code were summarized. - Abstract: A revision of the ITER Project-Level Load Specification (to be used for all systems of the ITER machine) was implemented in April 2012. This revision supports ITER's licensing by accommodating requests from the French regulator to maintain consistency with the plasma physics database and our present understanding of plasma transients and electro-magnetic (EM) loads, to investigate the possibility of removing unnecessary conservatism in the load requirements and to review the list and definition of incidental cases. The purpose of this paper is to present the impact of this 2012 revision of the ITER Project-Level Load Specification (LS) on the ITER Vacuum Vessel (VV) loads and the main structural margins required by the applicable French code, RCC-MR.

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  13. Load bearing capacities and elastic-plastic behavior of reactor vessel internals

    International Nuclear Information System (INIS)

    Watanabe, Keita; Nagase, Ryuichi

    2017-01-01

    Radial Support Keys (RSKs) are installed at the bottom of Reactor Vessel Internal (RVI) of Pressurized Water Reactor (PWR) and fit into Core Support Lugs of Reactor Pressure Vessel (RPV). This structure provides reactor core horizontal support and transmits the loads between RVI and RPV. RSK is one of the critical parts of RVI from the view point of earthquake-proof safety. In order to assure the structural integrity of Nuclear Reactor in case of massive earthquake, load bearing capacities of RSK are confirmed by static loading tests with reduced-scale mockups. In addition, collapse loads of actual components calculated by Limit Analyses are conservative enough compared to the load bearing capacities confirmed by the test. Thus, the methodology to calculate collapse load by Limit Analysis is applicable to evaluation of structural integrity for RSK. (author)

  14. Thermal and mechanical cyclic loading of thick spherical vessels made of transversely isotropic materials

    International Nuclear Information System (INIS)

    Komijani, M.; Mahbadi, H.; Eslami, M.R.

    2013-01-01

    The aim of this paper is to obtain the dependency of the ratcheting, reversed plasticity, or shakedown behavior of spherical vessels made of some anisotropic materials to the stress category of imposed cyclic loading. The Hill anisotropic yield criterion with the kinematic hardening theories of plasticity based on the Prager and Armstrong–Frederick models are used to predict the yield of the vessel and obtain the plastic strains. An iterative numerical method is used to simulate the cyclic loading behavior of the structure. The effect of mean and amplitude of the mechanical and thermal loads on cyclic behavior and ratcheting rate of the vessel is investigated respectively. The ratcheting rate for the vessels made of transversely isotropic material is evaluated for the various ratios of anisotropy. -- Highlights: ► Cyclic loading analysis of anisotropic spheres is assessed. ► Using the Prager model results in ratcheting. ► Armstrong-Frederick model predicts ratcheting for load controlled cyclic loadings. ► The A-F model predicts ratcheting to a stabilized cycle for thermal loadings

  15. Biaxial Loading Tests for steel containment vessel

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  16. Biaxial Loading Tests for steel containment vessel

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  17. Heat load imposed on reactor vessels during in-vessel retention of core melts

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Su-Hyeon; Chung, Bum-Jin, E-mail: bjchung@khu.ac.kr

    2016-11-15

    Highlights: • Angular heat load on reactor vessel by natural convection of oxide pool was measured. • High Ra was achieved by using mass transfer experiments based on analogy concept. • Measured Nusselt numbers agreed reasonably with the other existing studies. • Three different types of volumetric heat sources were compared. • They didn’t affect the heat flux of the top plate but affected those of the reactor vessel. - Abstract: We measured the heat load imposed on reactor vessels by natural convection of the oxide pool in severe accidents. Based on the analogy between heat and mass transfer, mass transfer experiments were performed using a copper sulfate electroplating system. A modified Rayleigh number of the order 10{sup 14} was achieved in a small facility with a height of 0.1 m. Three different types of volumetric heat sources were compared and the average Nusselt number of the curved surface was 39% lower, whereas in the case of the top plate was 6% higher than in previous studies with a two-dimensional geometry due to the high Sc value of this study. Reliable experimental data on the angular heat flux ratios were reported compared to those of the BALI and SIGMA CP facilities in terms of fluctuations and consistency.

  18. Shock loading of reactor vessel following hypothetical core disruptive accident

    International Nuclear Information System (INIS)

    Srinivas, G.; Doshi, J.B.

    1990-01-01

    Hypothetical Core Disruptive Accident (HCDA) has been historically considered as the maximum credible accident in Fast Breeder Reactor systems. Environmental consequences of such an accident depends to a great extent on the ability of the reactor vessel to maintain integrity during the shock loading following an HCDA. In the present paper, a computational model of the reactor core and the surrounding coolant with a free surface is numerical technique. The equations for conservation of mass, momentum and energy along with an equation of state are considered in two dimensional cylindrical geometry. The reactor core at the end of HCDA is taken as a bubble of hot, vaporized fuel at high temperature and pressure, formed at the center of the reactor vessel and expanding against the surrounding liquid sodium coolant. The free surface of sodium at the top of the vessel and the movement of the core bubble-liquid coolant interface are tracked by Marker and Cell (MAC) procedure. The results are obtained for the transient pressure at the vessel wall and also for the loading on the roof plug by the impact of the slug of liquid sodium. The computer code developed is validated against a benchmark experiment chosen to be ISPRA experiment reported in literature. The computer code is next applied to predict the loading on the Indian Prototype Fast Breeder Reactor (PFBR) being developed at Kalpakkam

  19. Assessment of integrity for the pressure vessel internals of PWRs under blowdown loadings

    International Nuclear Information System (INIS)

    Geiss, M.; Benner, J.; Ludwig, A.

    1984-01-01

    In safety analysis of pressurized water reactors the loss-of-coolant accident plays a central role. Thereby a sudden break of a cold primary coolant pipe close to the reactor pressure vessel is postulated. The sudden pressure release of the primary system (blowdown) causes high dynamic loading on the pressure vessel internals. The resulting deformations must not impair shut down of the reactor and decay heat removal in an inadmissible way. For this assessment a blowdown analysis for a 1300 MW pressurized water reactor is carried out. These investigations are completed with a detailed stress analysis for the highly loaded core barrel clamping. The results show that the reactor pressure vessel internals are able to withstand blowdown loading. Even in case of a sudden and complete break of the primary coolant pipe the loading has to be twice as high to endanger the structural integrity. (orig.) [de

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

    Science.gov (United States)

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

    2012-12-01

    Autonomous platforms are increasingly used as components of Integrated Ocean Observing Systems and oceanographic research cruises. Systems deployed can include gliders or propeller-driven autonomous underwater vessels (AUVs), autonomous surface vessels (ASVs), and unmanned aircraft systems (UAS). Prior field campaigns have demonstrated successful communication, sensor data fusion and visualization for studies using gliders and AUVs. However, additional requirements exist for incorporating ASVs and UASs into ship operations. For these systems to be optimally integrated into research vessel data management and operational planning systems involves addressing three key issues: real-time field data availability, platform coordination, and data archiving for later analysis. A fleet of AUVs, ASVs and UAS deployed from a research vessel is best operated as a system integrated with the ship, provided communications among them can be sustained. For this purpose, Disruptive Tolerant Networking (DTN) software protocols for operation in communication-challenged environments help ensure reliable high-bandwidth communications. Additionally, system components need to have considerable onboard autonomy, namely adaptive sampling capabilities using their own onboard sensor data stream analysis. We discuss Oceanographic Decision Support System (ODSS) software currently used for situational awareness and planning onshore, and in the near future event detection and response will be coordinated among multiple vehicles. Results from recent field studies from oceanographic research vessels using AUVs, ASVs and UAS, including the Rapid Environmental Picture (REP-12) cruise, are presented describing methods and results for use of multi-vehicle communication and deliberative control networks, adaptive sampling with single and multiple platforms, issues relating to data management and archiving, and finally challenges that remain in addressing these technological issues. Significantly, the

  1. 46 CFR 173.025 - Additional intact stability standards: Counterballasted vessels.

    Science.gov (United States)

    2010-10-01

    ...) SUBDIVISION AND STABILITY SPECIAL RULES PERTAINING TO VESSEL USE Lifting § 173.025 Additional intact stability standards: Counterballasted vessels. (a) Each vessel equipped to counterballast while lifting must be shown... loading and operation and at each combination of hook load and crane radius. (b) When doing the...

  2. Fourier series analysis of a cylindrical pressure vessel subjected to axial end load and external pressure

    International Nuclear Information System (INIS)

    Brar, Gurinder Singh; Hari, Yogeshwar; Williams, Dennis K.

    2013-01-01

    This paper presents the comparison of a reliability technique that employs a Fourier series representation of random axisymmetric and asymmetric imperfections in a cylindrical pressure vessel subjected to an axial end load and external pressure, with evaluations prescribed by the ASME Boiler and Pressure Vessel Code, Section VIII, Division 2 Rules. The ultimate goal of the reliability technique described herein is to predict the critical buckling load associated with the subject cylindrical pressure vessel. Initial geometric imperfections are shown to have a significant effect on the calculated load carrying capacity of the vessel. Fourier decomposition was employed to interpret imperfections as structural features that can be easily related to various other types of defined imperfections. The initial functional description of the imperfections consists of an axisymmetric portion and a deviant portion, which are availed in the form of a double Fourier series. Fifty simulated shells generated by the Monte Carlo technique are employed in the final prediction of the critical buckling load. The representation of initial geometrical imperfections in the cylindrical pressure vessel requires the determination of respective Fourier coefficients. Multi-mode analyses are expanded to evaluate a large number of potential buckling modes for both predefined geometries in combination with asymmetric imperfections as a function of position within the given cylindrical shell. The probability of the ultimate buckling stress exceeding a predefined threshold stress is also calculated. The method and results described herein are in stark contrast to the “knockdown factor” approach as applied to compressive stress evaluations currently utilized in industry. Further effort is needed to improve on the current design rules regarding column buckling of large diameter pressure vessels subjected to an axial end load and external pressure designed in accordance with ASME Boiler and

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

    Science.gov (United States)

    Storhaug, Gaute

    2014-12-01

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

  4. 46 CFR 180.202 - Survival craft-vessels operating on oceans routes.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Survival craft-vessels operating on oceans routes. 180... VESSELS (UNDER 100 GROSS TONS) LIFESAVING EQUIPMENT AND ARRANGEMENTS Number and Type of Survival Craft § 180.202 Survival craft—vessels operating on oceans routes. (a) Each vessel certificated to operate on...

  5. 43 CFR 423.38 - Operating vessels on Reclamation waters.

    Science.gov (United States)

    2010-10-01

    ... 43 Public Lands: Interior 1 2010-10-01 2010-10-01 false Operating vessels on Reclamation waters... WATERBODIES Rules of Conduct § 423.38 Operating vessels on Reclamation waters. (a) You must comply with... Reclamation waters, and with any restrictions established by an authorized official. (b) You must not operate...

  6. Test of 6-in.-thick pressure vessels. Series 3: intermediate test vessel V-7A under sustained loading

    International Nuclear Information System (INIS)

    Bryan, R.H.; Cate, T.M.; Holz, P.P.; King, T.A.; Merkle, J.G.; Robinson, G.C.; Smith, G.C.; Smith, J.E.; Whitman, G.D.

    1978-01-01

    HSST intermediate test vessel V-7 was repaired after being tested hydrostatically to leakage and was retested pneumatically as vessel V-7A. Except for the method of applying the load, the conditions in both tests were nearly identical. In each case, a sharp outside surface flaw 547 mm long (18 in.) by about 135 mm deep (5.3 in.) was prepared in the 152-mm-thick (6-in.) test cylinder of A533, grade B, class 1 steel. The inside surface of vessel V-7A was sealed in the region of the flaw by a thin metal patch so that pressure could be sustained after rupture. Vessel V-7A failed by rupture of the flaw ligament without burst, as expected. Rupture occurred at 144.3 MPa (20.92 ksi), after which pressure was sustained for 30 min without any indication of instability. The rupture pressure of vessel V-7A was about 2 percent less than that of vessel V-7

  7. Stresses and displacements in vessels due to loads imposed by single and multiple piping attachments

    International Nuclear Information System (INIS)

    Wong, F.M.G.; Craft, W.J.; East, G.H.

    1985-01-01

    The Fourier solution for thin shell equations models pressure vessels as continuous simply connected surfaces with local loads. The technique allows placement of tractions with combinations of radial, shear, and axial components. Unlike Bijlaard, the solution in this paper includes loads placed at any position along the cylinder. Stiffness and the enhanced load-carrying capacity that internal pressure gives to thin vessels can be simulated. Numerical convergence problems are reduced by an improved displacement-load algorithm, and by use of load sites that allow the circular functions to be compactly grouped. A variety of loading distributions may be analyzed including large and small nozzles near and away from centerlines. Both rectangular and circular attachments are simulated. Through superposition, multiple attachments with their own loads may be examined. The attachments to the vessel may be either rigid or soft. A comparison to analytical results from Bijlaard shows excellent agreement. Comparisons with experimental tests on an API-650 nozzle on a storage tank are in good agreement. Variations between experimental and calculated results are primarily caused by assuming a simply supported base in the calculation, whereas in the experimental test, the base is more nearly fixed

  8. Pressure vessels supported in the soil submitted to axissymetrical loads

    International Nuclear Information System (INIS)

    Gouvea, J.P. de; Bevilacqua, L.

    1982-01-01

    A pressure vessel, spherical segment or vertical cylinder, is supported in the soil and submitted to axissymetrical loads. The soil is considered as a semi-infinite elastic solid and the support as a lattice. The method of rigidity is used. (E.G.) [pt

  9. Dynamic loads on reactor vessel components by low pressure waves

    International Nuclear Information System (INIS)

    Benkert, J.; Mika, C.; Stegemann, D.; Valero, M.

    1978-01-01

    Starting from the conservation theorems for mass and impulses the code DRUWE has been developed enabling the calculation of dynamic loads of the reactor shell on the basis of simplified assumptions for the first period shortly after rupture. According to the RSK-guidelines it can be assumed that the whole weld size is opened within 15 msec. This time-dependent opening of the fractured plane can be taken into account in the computer program. The calculation is composed in a way that for a reactor shell devided into cross and angle sections the local, chronological pressure and strength curves, the total dynamic load as well as the moments acting on the fastenings of the reactor shell can be calculated. As input data only geometrical details concerning the concept of the pressure vessel and its components as well as the effective subcooling of the fluid are needed. By means of several parameters the program can be operated in a way that the results are available in form of listings or diagrams, respectively, but also as card pile for further examinations, e.g. strength analysis. (orig./RW) [de

  10. 46 CFR 180.208 - Survival craft-vessels operating on rivers routes.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Survival craft-vessels operating on rivers routes. 180... VESSELS (UNDER 100 GROSS TONS) LIFESAVING EQUIPMENT AND ARRANGEMENTS Number and Type of Survival Craft § 180.208 Survival craft—vessels operating on rivers routes. (a) Except as allowed by paragraphs (c), (d...

  11. 46 CFR 117.208 - Survival craft-vessels operating on rivers routes.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Survival craft-vessels operating on rivers routes. 117... LIFESAVING EQUIPMENT AND ARRANGEMENTS Number and Type of Survival Craft § 117.208 Survival craft—vessels... vessel certificated to operate on a rivers route in warm water is not required to carry survival craft...

  12. Thermal structural analysis of SST-1 vacuum vessel and cryostat assembly using ANSYS

    International Nuclear Information System (INIS)

    Santra, Prosenjit; Bedakihale, Vijay; Ranganath, Tata

    2009-01-01

    Steady state super-conducting tokamak-1 (SST-1) is a medium sized tokamak, which has been designed to produce a 'D' shaped double null divertor plasma and operate in quasi steady state (1000 s). SST-1 vacuum system comprises of plasma chamber (vacuum vessel, interconnecting rings, baking and cooling channels), and cryostat all made of SS 304L material designed to meet ultra high vacuum requirements for plasma generation and confinement. Prior to plasma shot and operation the vessel assembly is baked to 250/150 deg. C from room temperature and discharge cleaned to remove impurities/trapped gases from wall surfaces. Due to baking the non-uniform temperature pattern on the vessel assembly coupled with atmospheric pressure loading and self-weight give rise to high thermal-structural stresses, which needs to be analyzed in detail. In addition the vessel assembly being a thin shell vessel structure needs to be checked for critical buckling load caused by atmospheric and baking thermal loads. Considering symmetry of SST-1, 1/16th of the geometry is modeled for finite element (FE) analysis using ANSYS for different loading scenarios, e.g. self-weight, pressure loading considering normal operating conditions, and off-normal loads coupled with baking of vacuum vessel from room temperature 250 deg. C to 150 deg. C, buckling and modal analysis for future dynamic analysis. The paper will discuss details about SST-1 vacuum system/cryostat, solid and FE model of SST-1, different loading scenarios, material details and the stress codes used. We will also present the thermal structural results of FE analysis using ANSYS for various load cases being investigated and our observations under different loading conditions.

  13. 46 CFR 180.205 - Survival craft-vessels operating on limited coastwise routes.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Survival craft-vessels operating on limited coastwise... Craft § 180.205 Survival craft—vessels operating on limited coastwise routes. (a) Except as allowed by... survival craft required by § 180.204(d). (e) Each vessel certificated to operate on a limited coastwise...

  14. Cyclic loading of thick vessels based on the Prager and Armstrong-Frederick kinematic hardening models

    International Nuclear Information System (INIS)

    Mahbadi, H.; Eslami, M.R.

    2006-01-01

    The aim of this paper is to relate the type of stress category in cyclic loading to ratcheting or shakedown behaviour of the structure. The kinematic hardening theory of plasticity based on the Prager and Armstrong-Frederick models is used to evaluate the cyclic loading behaviour of thick spherical and cylindrical vessels under load and deformation controlled stresses. It is concluded that kinematic hardening based on the Prager model under load and deformation controlled conditions, excluding creep, results in shakedown or reversed plasticity for spherical and cylindrical vessels with the isotropy assumption of the tension/compression curve. Under an anisotropy assumption of the tension/compression curve, this model predicts ratcheting. On the other hand, the Armstrong-Frederick model predicts ratcheting under load controlled cyclic loading and reversed plasticity for deformation controlled stress. The interesting conclusion is that the Armstrong-Frederick model is well capable to predict the experimental data under the assumed type of stresses, wherever experimental data are available

  15. 33 CFR 161.12 - Vessel operating requirements.

    Science.gov (United States)

    2010-07-01

    ....0′ N. extending eastward through the Golden Gate, and the navigable waters of San Francisco Bay and... safety beyond that provided by other means. The bridge-to-bridge navigational frequency, 156.650 MHz (Ch... Measures, and Operating Requirements § 161.12 Vessel operating requirements. (a) Subject to the exigencies...

  16. Transient temperature response of in-vessel components due to pulsed operation in tokamak fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    Minato, Akio; Tone, Tatsuzo

    1985-12-01

    A transient temperature response of the in-vessel components (first wall, blanket, divertor/limiter and shielding) surrounding plasma in Tokamak Fusion Experimental Reactor (FER) has been analysed. Transient heat load during start up/shut down and pulsed operation cycles causes the transient temperature response in those components. The fatigue lifetime of those components significantly depends upon the resulting cyclic thermal stress. The burn time affects the temperature control in the solid breeder (Li 2 O) and also affects the thermo-mechanical design of the blanket and shielding which are constructed with thick structure. In this report, results of the transient temperature response obtained by the heat transfer and conduction analyses for various pulsed operation scenarios (start up, shut down, burn and dwell times) have been investigated in view of thermo-mechanical design of the in-vessel components. (author)

  17. Design concept for vessels and heat exchangers

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  18. 46 CFR 105.45-1 - Loading or dispensing petroleum products.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Loading or dispensing petroleum products. 105.45-1... VESSELS COMMERCIAL FISHING VESSELS DISPENSING PETROLEUM PRODUCTS Special Operating Requirements § 105.45-1 Loading or dispensing petroleum products. (a) A commercial fishing vessel must have aboard a letter of...

  19. 46 CFR 180.206 - Survival craft-vessels operating on Great Lakes routes.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Survival craft-vessels operating on Great Lakes routes... Craft § 180.206 Survival craft—vessels operating on Great Lakes routes. (a) Except as allowed by... with the survival craft required by § 180.205 (a) through (e), as appropriate. (b) Each vessel...

  20. LOFT reactor vessel 290/sup 0/ downcomer stalk instrument penetration flange stress analysis

    Energy Technology Data Exchange (ETDEWEB)

    Finicle, D.P.

    1978-06-06

    The LOFT Reactor Vessel 290/sup 0/ Downcomer Stalk Instrument Penetration Flange Stress Analysis has been completed using normal operational and blowdown loading. A linear elastic analysis was completed using simplified hand analysis techniques. The analysis was in accordance with the 1977 ASME Boiler and Pressure Vessel Code, Section III, for a Class 1 component. Loading included internal pressure, bolt preload, and thermal gradients due to normal operating and blowdown.

  1. A system for rotatably mounting a vessel to a loading buoy. System for dreibar tilkopling av et flytende farty til en lastebye

    Energy Technology Data Exchange (ETDEWEB)

    Breivik, K.; Smedal, A.; Syvertsen, K.

    1994-07-04

    The invention relates to a system for rotatable mounting of a floating vessel to a submerged loading/unloading buoy which is anchored to the sea bed. The buoy is adapted to be introduced into and fastened in a releasable manner in a submerged downwardly open receiving space in the vessel, and is during operation connected to at least one transfer line and forming a transfer connection between this line and a tube system on the vessel. The buoy comprises an outer member which is arranged to be rigidly fastened in the receiving space, and a central inner member which is rotatably mounted in the outer member, so that the vessel is able to turn about the central member when the buoy is fastened in the receiving space. Further, the upper end of the central member is connected to the tube system of the vessel through a swivel means and through at least one flexible joint means respectively. 3 figs.

  2. [Key vessels assessment and operation highlights in laparoscopic extended right hemicolectomy].

    Science.gov (United States)

    Wang, Hao; Zhao, Quanquan

    2018-03-25

    Laparoscopic radical colectomies have been more widely used gradually, among which laparoscopic extended right hemicolectomy is considered as the most difficult procedure. The difficulty of extended right hemicolectomy lies in the need to dissect lymph nodes along the superior mesenteric vein (SMV) and disconnect numerous and possible aberrant vessels. To address this problem, we emphasize two points in key vessel assessment: getting familiar with the anatomy along the medial-to-lateral approach and having a good understanding about the preoperative imaging presentations. An accurately preoperative imaging assessment by abdominal enhanced CT can help the surgeon understand the relative position of the key vessels to be dealt with during operation and the situation of the possible aberrant vessels so as to guide the procedure more effectively and facilitate the prevention and management of the intraoperative complications. During operation, the operator should pay special attention to the management of the vessels in the ileocolic vessel region, Henle's trunk and middle colon vessels. The operation highlights of the key vessels are as follows: (1) The ileocolic vessels: identifying the Toldt's gap correctly and opening the vascular sheath of the SMV securely; making sure that the duodenum is well protected. (2) Henle's trunk: dissecting along the surface of the Henle's trunk; preserving the anterior superior pancreaticoduodenal vein (ASPDV) and main trunk of the Henle's trunk; disconnecting the roots of the right colic vein (RCV) and right gastroepiploic vein (RGEV), and then dissecting lymph nodes along the surface of the pancreas. (3) The middle colon vessels: identifying the root of the middle colon vessel along the lower edge of the pancreas; avoiding entering behind the pancreas; mobilizing the transverse mesocolon sufficiently along the surface of the pancreas. Finally, we discuss and analyze the disputes currently existing in laparoscopic extended right

  3. Offshore wind transport and installation vessel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    The initial objective of the project was to complete a feasibility study to determine the viability of an innovative transportation vessel to be deployed in the installation of offshore wind farms. This included the feasibility of providing a stable-working platform that can be used in harsh offshore environments. A study of current installation contractors and their installation equipment was used to provide a preliminary specification for the installation vessel. A typical barge was selected and a number of hydrodynamic analyses were carried out in order to establish it's on course and operational stability. The analysis proved the stability of the vessel during operation was critical and that in order to utilise the crane's full potential a stabilisation system must be employed. The main aim of the work to date was to establish whether it was feasible to use a stabilisation system on the installation vessel. The spud leg FEED study established that it was feasible to use spud legs to stabilise the vessel. In order to achieve the degree of stability required it is necessary to lift the vessel completely out of the water. This was not the original aim of the study but due to the external loads on the hull it was the only viable option. Lifting the vessel out of the water results in the legs and leg casings becoming very large. This has a number of consequences for the final design. Due to large loads on the legs spud cans must be used to avoid bottom penetration, the spud cans increase the draft of the vessel by 2m. The large loads require larger winches and more reeving to be used, this results in larger pumps and motors, all of which have to be housed. The stabilisation system has been proved to be feasible for a large installation vessel, the cost and physical size are however more excessive than first anticipated. (Author)

  4. Thermal Load Analysis of Multilayered Corium in the Lower Head of Reactor Pressure Vessel during Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Whang, Seok Won; Park, Hyun Sun [POSTECH, Pohang (Korea, Republic of); Hwang, Tae Suk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-05-15

    In-Vessel Retention (IVR) is one of the severe accident management strategies to terminate or mitigate the severe accident which is also called 'core-melt accident'. The reactor vessel would be cooled by flooding the cavity with water. The molten core mixture is divided into two or three layers due to the density difference. Light metal layer which contains Fe and Zr is on the oxide layer which is consist of UO{sub 2} and ZrO{sub 2}. Heavy metal layer which contains U, Fe and Zr is located under the oxide layer. In oxide layer, the crust which is solidified material is formed along the boundary. The assessment of IVR for nuclear power plant has been conducted with lumped parameter method by Theofanous, Rempe and Esmaili. In this paper, the numerical analysis was performed and verified with the Esmaili's work to analyze thermal load of multilayered corium in pressurized reactor vessel and also to examine the condition of in-vessel corium characteristic before the vessel failure that lead to ex-vessel severe accident progression for example, ex-vessel debris bed cooling. The in-vessel coolability analysis for several scenarios is conducted for the plant which has higher power than AP1000. Two sensitivity analyses are conducted, the first is emissivity of light metal layer and the second is the heat transfer coefficient correlations of oxide layer. The effect of three layered system also investigated. In this paper, the numerical analysis was performed and verified with Esmaili's model to analyze thermal load of multilayered corium in pressurized reactor vessel. For two layered system, thermal load was analyzed according to the severe accident scenarios, emissivity of the light metal layer and heat transfer correlations of the.

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

    International Nuclear Information System (INIS)

    Fanning, D.N.

    1978-09-01

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

  6. 46 CFR 180.204 - Survival craft-vessels operating on coastwise routes.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Survival craft-vessels operating on coastwise routes. 180.204 Section 180.204 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) SMALL... Craft § 180.204 Survival craft—vessels operating on coastwise routes. (a) Except as allowed by paragraph...

  7. Safety assessment of in-vessel vapor explosion loads in next generation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Kwang Hyun; Cho, Jong Rae; Choi, Byung Uk; Kim, Ki Yong; Lee, Kyung Jung [Korea Maritime University, Busan (Korea); Park, Ik Kyu [Seoul National University, Seoul (Korea)

    1998-12-01

    A safety assessment of the reactor vessel lower head integrity under in-vessel vapor explosion loads has been performed. The premixing and explosion calculations were performed using TRACER-II code. Using the calculated explosion pressures imposed on the lower head inner wall, strain calculations were performed using ANSYS code. The explosion analyses show that the explosion impulses are not altered significantly by the uncertain parameters of triggering location and time, fuel and vapor volume fractions in uniform premixture bounding calculations within the conservative ranges. Strain analyses using the calculated pressure loads on the lower head inner wall show that the vapor explosion-induced lower head failure is physically unreasonable. The static analysis using the conservative explosion-end pressure of 7,246 psia shows that the maximum equivalent strain is 4.3% at the bottom of lower head, which is less than the allowable threshold value of 11%. (author). 24 refs., 40 figs., 3 tabs.

  8. Marine Vessel Models in Changing Operational Conditions - A Tutorial

    DEFF Research Database (Denmark)

    Perez, Tristan; Sørensen, Asgeir; Blanke, Mogens

    2006-01-01

    conditions (VOC). However, since marine systems operate in changing VOCs, there is a need to adapt the models. To date, there is no theory available to describe a general model valid across different VOCs due to the complexity of the hydrodynamic involved. It is believed that system identification could......This tutorial paper provides an introduction, from a systems perspective, to the topic of ship motion dynamics of surface ships. It presents a classification of parametric models currently used for monitoring and control of marine vessels. These models are valid for certain vessel operational...

  9. Vessel Operating Units (Vessels)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains data for vessels that are greater than five net tons and have a current US Coast Guard documentation number. Beginning in1979, the NMFS...

  10. 46 CFR 180.207 - Survival craft-vessels operating on lakes, bays, and sounds routes.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Survival craft-vessels operating on lakes, bays, and... Survival Craft § 180.207 Survival craft—vessels operating on lakes, bays, and sounds routes. (a) Except as... warm water is not required to carry survival craft. (d) A vessel certificated to operate on lakes, bays...

  11. 36 CFR 3.8 - What vessel operations are prohibited?

    Science.gov (United States)

    2010-07-01

    .... (4) Operating a vessel in excess of flat wake speed within 100 feet of: (i) A downed water skier; (ii... the endangering of the life, limb, or property of a person(s) through the operator's lack of knowledge...

  12. 36 CFR 3.15 - What is the maximum noise level for the operation of a vessel?

    Science.gov (United States)

    2010-07-01

    ... level for the operation of a vessel? 3.15 Section 3.15 Parks, Forests, and Public Property NATIONAL PARK... level for the operation of a vessel? (a) A person may not operate a vessel at a noise level exceeding... vessel is being operated in excess of the noise levels established in paragraph (a) of this section may...

  13. Models and Algorithms for Container Vessel Stowage Optimization

    DEFF Research Database (Denmark)

    Delgado-Ortegon, Alberto

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

  14. Structural failure analysis of reactor vessels due to molten core debris

    International Nuclear Information System (INIS)

    Pfeiffer, P.A.

    1993-01-01

    Maintaining structural integrity of the reactor vessel during a postulated core melt accident is an important safety consideration in the design of the vessel. This paper addresses the failure predictions of the vessel due to thermal and pressure loadings from the molten core debris depositing on the lower head of the vessel. Different loading combinations were considered based on a wet or dry cavity and pressurization of the vessel based on operating pressure or atmospheric (pipe break). The analyses considered both short term (minutes) and long term (days) failure modes. Short term failure modes include creep at elevated temperatures and plastic instabilities of the structure. Long term failure modes are caused by creep rupture that lead to plastic instability of the structure. The analyses predict the reactor vessel will remain intact after the core melt has deposited on the lower vessel head

  15. Load-following operation of PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jong Hwa; Oh, Soo Yul; Koo, Yang Hyun; Lee, Jae Han [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-12-01

    The load-following operation of nuclear power plants will become inevitable due to the increased nuclear share in the total electricity generation. As a groundwork for the load-following capability of the Korean next generation PWRs, the state-of-the-art has been reviewed. The core control principles and methods are the main subject in this review as well as the impact of load-following operations on the fuel performance and on the mechanical integrity of components. To begin with, it was described what the load-following operation is and in what view point the technology should be reviewed. Afterwards the load-following method, performance and problems in domestic 900 MWe class PWRs were discussed, and domestic R and D works were summarized. Foreign technologies were also reviewed. They include Mode G and Mode X of Foratom, D and L bank method of KWU, the method using PSCEA of ABB-CE, and MSHIM of Westinghouse. The load-following related special features of Foratom`s N4 plant, KWU`s plants, ABB-CE`s Systems 80+, and Westinghouse`s AP600 were described in each technology review. The review concluded that the capability of N4 plant with Mode X is the best and the methods in System, 80+ and AP600 would require verifications for the continued and usual load-following operation. It was recommended that the load-following operation experiences in domestic PWRs under operation be required to settle down the capability for the future. In addition, a more enhanced technology is required for the Korean next generation PWR regardless what the reference plant concept is. 30 figs., 19 tabs., 75 refs. (Author).

  16. 46 CFR 117.206 - Survival craft-vessels operating on Great Lakes routes.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Survival craft-vessels operating on Great Lakes routes... PASSENGERS LIFESAVING EQUIPMENT AND ARRANGEMENTS Number and Type of Survival Craft § 117.206 Survival craft... vessel certificated to operate on a Great Lakes route must be provided with the survival craft required...

  17. 46 CFR 117.207 - Survival craft-vessels operating on lakes, bays, and sounds routes.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Survival craft-vessels operating on lakes, bays, and... 49 PASSENGERS LIFESAVING EQUIPMENT AND ARRANGEMENTS Number and Type of Survival Craft § 117.207 Survival craft—vessels operating on lakes, bays, and sounds routes. (a) Each vessel with overnight...

  18. Filament wound pressure vessels with load sharing liners for space shuttle orbiter applications

    International Nuclear Information System (INIS)

    Ecord, G.M.

    1976-01-01

    Early in the development of orbiter propulsion and environmental control subsystems it was recognized that use of overwrapped pressure vessels with load sharing liners may provide significant weight savings for high pressure gas containment. A program is described which was undertaken by Rockwell International to assess the utility for orbiter applications of titanium 6Al--4V and Inconel 718 liners overwrapped with Kevlar fibers. Also briefly described are programs administered by the NASA Lewis Research Center to evaluate cryoformed steel liners overwrapped with Kevlar fibers and to establish a method that can guarantee cyclic life of the vessels

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

    International Nuclear Information System (INIS)

    Finicle, D.P.

    1977-01-01

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

  20. In vivo experimental study on laser welded ICG-loaded chitosan patches for vessel repair

    Science.gov (United States)

    Rossi, Francesca; Matteini, Paolo; Esposito, Giuseppe; Albanese, Alessio; Puca, Alfredo; Maira, Giulio; Rossi, Giacomo; Pini, Roberto

    2011-03-01

    Laser welding of microvessels provides several advantages over conventional suturing techniques: surgical times reduction, vascular healing process improvement, tissue damage reduction. We present the first application of biopolymeric patches in an in vivo laser assisted procedure for vessel repair. The study was performed in 20 New Zealand rabbits. After anesthesia, a 3-cm segment of the right common carotid artery was exposed and clamped proximally and distally. A linear lesion 3 mm in length was carried out. We used a diode laser emitting at 810 nm and equipped with a 300 μm diameter optical fiber. To close the cut, ICG-loaded chitosan films were prepared: chitosan is characterized by biodegradability, biocompatibility, antimicrobial, haemostatic and wound healing-promoting activity. ICG is an organic chromophore commonly used in the laser welding procedures to mediate the photothermal conversion at the basis of the welding effect. The membranes were used to wrap the whole length of the cut, and then they were welded in the correct position by delivering single laser spots to induce local patch/tissue adhesion. The result is an immediate closure of the wound, with no bleeding at clamps release. The animals were observed during follow-up and sacrificed after 2, 7, 30 and 90 days. All the repaired vessels were patent, no bleeding signs were documented. The carotid samples underwent histological examinations. The advantages of the proposed technique are: simplification of the surgical procedure and shortening of the operative time; good strength of the vessel repair; decreased foreign-body reaction, reduced inflammatory response and improved vascular healing process.

  1. Loads on reactor pressure vessel internals induced by low-pressure waves

    International Nuclear Information System (INIS)

    Benkert, J.; Mika, C.; Stegemann, D.; Valero, M.

    1978-02-01

    Departing from the conservation theorems for mass and impulse the computer code DRUWE has been developed which allows to calculate loads on the core shell with simplifying assumptions for the first period just after the rupture has opened. It can be supposed that the whole rupture cross section is set free within 15 msec. The calculation progresses in a way that for a core shell the local, timely pressure- and load development, respectively, the total dynamic load as well as the moments acting on the fixing of the core shell, can be calculated. The required input data are merely geometric data on the concept of the pressure vessel and its components as well as the effective subcooling of the fluid. By means of some parameters the programm development can be controlled in a way that the results are available in form of listings or diagrams, respectively, as well as in form of card decks for following investigations, e.g. solidity calculations. (orig./RW) [de

  2. The need to pressure test prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  3. Thermal loads on the TJ-II Vacuum Vessel under Neutral Beam Injection

    International Nuclear Information System (INIS)

    Guasp, J.; Fuentes, C.; Liniers, M.

    1996-01-01

    In this study a numerical analysis of power loads on the complex 3D structure of the TJ-II Vacuum Vessel, moderated with reasonable accuracy, under NBI, is done. To do this it has been necessary to modify deeply the DENSB code for power loads in order to include the TJ-II VV wall parts as targets and as beam scrapers, allowing the possibility of self-shadowing. After a short description of the primitive version of the DENSB code (paragraph 2) and of the visualisation code MOVIE(paragraph 3), the DENSB upgrading are described (paragraphs 4,5) and finally the results are presented (paragraph 6). These code modifications and the improving on the visualization tools provide more realistic load evaluations, both with and without plasma, validating former results and showing clearly the VV zones that will need new protections. (Author)

  4. Part-load pumping operation, control and behaviour

    International Nuclear Information System (INIS)

    1988-01-01

    Twenty one papers are printed. Their main concern is with pump performance under abnormal operating conditions or when operating at part-load. The effect of part-load operation on pump performance, pump internal flow and pump cavitation and noise were considered. The pumps considered are used in a variety of situations and some case studies were discussed. One paper about part-load operation of the boiler pumps for French pressurized water reactors and one paper on pressure pulsations of centrifugal pumps at very low flowrate, are indexed separately. (U.K.)

  5. Transient analysis of house load operation for LNPP

    International Nuclear Information System (INIS)

    Shi Junying; Zheng Bin

    2000-01-01

    The author analysis the transient of house load operation for Ling'ao Nuclear Power Plant by using the methods of dynamic simulation and closed loops of primary and secondary system. The transient of house load operation from 100% FP is the most severe that can occur on the unit in normal operation because it causes immediately shedding of 95% of turbine load and requires the unit to operate steadily at reduced power. The results show that the transient can be successful both at beginning of core life and manual house load operation. However, more attentions must be paid to automatic house load operation caused by grid fault at toward end of core life because the success of the transient could be threatened by the actuation of the protection of high flux and high flux rate

  6. 46 CFR 117.202 - Survival craft-vessels operating on oceans routes.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Survival craft-vessels operating on oceans routes. 117... LIFESAVING EQUIPMENT AND ARRANGEMENTS Number and Type of Survival Craft § 117.202 Survival craft—vessels... number of overnight persons allowed, the survival craft requirements contained in paragraph (e) of this...

  7. In-vessel maintenance remote manipulator system

    International Nuclear Information System (INIS)

    Jimenez, E.

    1978-01-01

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

  8. Research program plan: reactor vessels. Volume 1

    International Nuclear Information System (INIS)

    Vagins, M.; Taboada, A.

    1985-07-01

    The ability of the licensing staff of the NRC to make decisions concerning the present and continuing safety of nuclear reactor pressure vessels under both normal and abnormal operating conditions is dependent upon the existence of verified analysis methods and a solid background of applicable experimental data. It is the role of this program to provide both the analytical methods and the experimental data needed. Specifically, this program develops fracture mechanics analysis methods and design criteria for predicting the stress levels and flaw sizes required for crack initiation, propagation, and arrest in LWR pressure vessels under all known and postulated operations conditions. To do this, not only must the methods be developed but they must be experimentally validated. Further, the materials data necessary for input to these analytical methods must be developed. Thus, in addition to methods development and large scale experimental verification this program also develops data to show that slow-load fracture toughness, rapid-load fracture toughness, and crack arrest toughness obtained from small laboratory specimens are truly representative of the toughness characteristics of the material behavior in pressure vessels in both the unirradiated and the irradiated conditions

  9. Reactor vessel supported by flexure member

    International Nuclear Information System (INIS)

    Crawford, J.D.; Pankow, B.

    1977-01-01

    According to the present invention there is provided an improved arrangement for supporting a reactor vessel within a containment structure against static and dynamic vertical loadings capable of being imposed as a result of a serious accident as well as during periods of normal plant operation. The support arrangement of the invention is, at the same time, capable of accommodating radial displacements that normally occur between the reactor vessel and the containment structure due to operational transients. The arrangement comprises a plurality of vertical columns connected between the reactor vessel and a support base within the containment structure. The columns are designed to accommodate relative displacements between the vessel and the containment structure by flexing. This eliminates the need for relative sliding movements and thus enables the columns to be securely fixed to the vessel. This elimination of a provision for relative sliding movements avoids the spaces or gaps between the retention members and the retained elements as occurred in prior art arrangements and, concomitantly, the danger of establishing impact forces on the retention members in the event of an accident is reduced. (author)

  10. Upper and Lower Bound Limit Loads for Thin-Walled Pressure Vessels Used for Aerosol Cans

    Directory of Open Access Journals (Sweden)

    Stephen John Hardy

    2009-01-01

    Full Text Available The elastic compensation method proposed by Mackenzie and Boyle is used to estimate the upper and lower bound limit (collapse loads for one-piece aluminium aerosol cans, which are thin-walled pressure vessels subjected to internal pressure loading. Elastic-plastic finite element predictions for yield and collapse pressures are found using axisymmetric models. However, it is shown that predictions for the elastic-plastic buckling of the vessel base require the use of a full three-dimensional model with a small unsymmetrical imperfection introduced. The finite element predictions for the internal pressure to cause complete failure via collapse fall within the upper and lower bounds. Hence the method, which involves only elastic analyses, can be used in place of complex elastic-plastic finite element analyses when upper and lower bound estimates are adequate for design purposes. Similarly, the lower bound value underpredicts the pressure at which first yield occurs.

  11. The response of pressure vessel steel specimens on drop weight loading

    International Nuclear Information System (INIS)

    Winkler, S.; Kalthoff, J.F.; Gerscha, A.

    1979-01-01

    Load records obtained in instrumented impact tests in general are disturbed by inertia effects. The influence of mechanical damping provisions on these disturbing inertia effects is investigated. Precracked bend specimens are dynamically loaded in a drop weight testing system. The specimens of size 620 mm x 150 mm (25 mm or 50 mm thick) were machined from the pressure vessel steel 22 NiMoCr 37 which was heat treated to achieve a specially hardened condition. The tests were performed at two different low temperatures. The impact velocity was about 4 m/s. As it is usual in instrumented impact testing, the load at the tup of the impining striker is recorded as a function of time during the impact process. In addition the specimen is instrumented by a strain gage close to the crack tip in order to directly measure the stress intensification. Experiments were performed under pure and damped impact conditions. Damping was achieved by utilizing a soft aluminum plate between the striker and the specimen. (orig.)

  12. Conceptual Design of Electrical Propulsion System for Nuclear Operated Vessel Adventurer

    International Nuclear Information System (INIS)

    Halimi, B.; Suh, K. Y.

    2009-01-01

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

  13. Load Extrapolation During Operation for Wind Turbines

    DEFF Research Database (Denmark)

    Toft, Henrik Stensgaard; Sørensen, John Dalsgaard

    2008-01-01

    In the recent years load extrapolation for wind turbines has been widely considered in the wind turbine industry. Loads on wind turbines during operations are normally dependent on the mean wind speed, the turbulence intensity and the type and settings of the control system. All these parameters...... must be taken into account when characteristic load effects during operation are determined. In the wind turbine standard IEC 61400-1 a method for load extrapolation using the peak over threshold method is recommended. In this paper this method is considered and some of the assumptions are examined...

  14. Structural analysis of the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Sannazzaro, G.; Ioki, K.; Johnson, G.; Onozuka, M.; Utin, Y. [ITER Joint Work Site, Garching (Germany); Nelson, B. [Oak Ridge National Lab., TN (United States); Swanson, J. [USHT, Raytheon, Princeton (United States)

    1998-07-01

    The ITER Vacuum Vessel (VV) must withstand a large number of loading conditions including electromagnetic, seismic, operational and upset pressure, thermal and test loads. All of the loading conditions and load combinations have been categorized and classified to permit the allowable stress to be defined in accordance with the recommendations of the ASME code. The most severe loading conditions for the VV are the toroidal field coil fast discharge (TFCFD) and the load combination of seismic and electromagnetic loads due to a plasma vertical instability. The areas of high stress are the regions around the VV and the blanket supports, and the attachment of the ports to the main shell. In all of the loading conditions and load combinations the calculated stresses are below the allowable values. (authors)

  15. Dynamic simulation of a planar flexible boom for tokamak in-vessel operations

    International Nuclear Information System (INIS)

    Ambrosino, G.; Celentano, G.; Garofalo, F.; Maisonnier, D.

    1991-01-01

    In this paper we present a dynamic model for the analysis of the vibrations of a planar articulated flexible boom to be used for tokamak in-vessel maintenance operations. The peculiarity of the mechanical structure of the boom enables us to consider separately the oscillations in the horizontal and vertical planes so that two separate models can be constructed for describing these phenomena. The results of simulations based on booms like that proposed for NET in-vessel operations are presented. (orig.)

  16. Scenario based optimization of a container vessel with respect to its projected operating conditions

    Science.gov (United States)

    Wagner, Jonas; Binkowski, Eva; Bronsart, Robert

    2014-06-01

    In this paper the scenario based optimization of the bulbous bow of the KRISO Container Ship (KCS) is presented. The optimization of the parametrically modeled vessel is based on a statistically developed operational profile generated from noon-to-noon reports of a comparable 3600 TEU container vessel and specific development functions representing the growth of global economy during the vessels service time. In order to consider uncertainties, statistical fluctuations are added. An analysis of these data lead to a number of most probable upcoming operating conditions (OC) the vessel will stay in the future. According to their respective likeliness an objective function for the evaluation of the optimal design variant of the vessel is derived and implemented within the parametrical optimization workbench FRIENDSHIP Framework. In the following this evaluation is done with respect to vessel's calculated effective power based on the usage of potential flow code. The evaluation shows, that the usage of scenarios within the optimization process has a strong influence on the hull form.

  17. Evaluation of Thermal Load to the Lower Head Vessel Using the ASTEC Computer Code

    International Nuclear Information System (INIS)

    Park, Raejoon; Ahn, Kwangil

    2013-01-01

    The thermal load from the corium to the lower head vessel in the APR (Advanced Power reactor) 1400 during a small break loss of coolant accident (SBLOCA) without a safety injection (SI) has been evaluated using the ASTEC (Accident Source Term Evaluation Code) computer code, which has been developed as a part of the EU (European Union)-SARNET (Severe Accident Research NET work) program. The ASTEC results predict that the reactor vessel did not fail by using an ERVC, in spite of the large melting of the reactor vessel wall in a two-layer formation case of the SBLOCA in the APR1400. The outer surface conditions of the temperature and heat transfer coefficient are not effective on the vessel geometry change, which are preliminary results. A more detailed analysis of the main parameter effects on the corium behavior in the lower plenum is necessary to evaluate the IVR-ERVC in the APR1400, in particular, for a three-layer formation of the TLFW. Comparisons of the present results with others are necessary to verify and apply them to the actual IVR-ERVC evaluation in the APR1400

  18. 46 CFR 117.205 - Survival craft-vessels operating on limited coastwise routes.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Survival craft-vessels operating on limited coastwise... PASSENGERS LIFESAVING EQUIPMENT AND ARRANGEMENTS Number and Type of Survival Craft § 117.205 Survival craft... the survival craft required by §§ 117.204 (a) through (d) of this part, as applicable. (b) Each vessel...

  19. Stresses from pressure, radial, and moment loads in cylinder-to-cylinder vessel by a finite plate method

    International Nuclear Information System (INIS)

    Brown, S.J.; Fox, M.E.

    1977-08-01

    A structural problem that has received continued interest and development over the last several decades is the determination of stresses in two normally intersecting cylindrical shells subjected to internal pressure and external loading. In nuclear pressure vessels the external loading of the vessel through the attachment is encountered in thermal interaction, seismic loading and various postulated rupture or failure mechanisms. A simple technique, the Finite Plate Method, (FPM) is presented to analyze stresses in cylinder-to-cylinder junctures. The approach uses shallow shell formulations and a three term series expansion plate formulation, which limits the range of applicability. It is felt that the value of the method is its accuracy, economy, and ease in modeling a structure which falls within the range of applicability. Another appealing feature of the method is that its simplistic approach of superposition of results permits an easy extension to include additional loads not treated. For those mechanical loadings not developed, it is felt that their effect can either be accounted for by the mechanisms discussed or by simple calculations. Generally, the stresses resulting from torsional or transverse shear are small compared to the loads discussed, however, these shear effects may be included. Finally, in the instance of thermal stress within the cylinder-to-cylinder structure, it has been shown in an unpublished study by Brown that the FPM yields very good results for the range of curvatures discussed

  20. 46 CFR 520.11 - Non-vessel-operating common carriers.

    Science.gov (United States)

    2010-10-01

    ... CARRIER AUTOMATED TARIFFS § 520.11 Non-vessel-operating common carriers. (a) Financial responsibility. An... its tariff publication: (1) That it has furnished the Commission proof of its financial responsibility..., insurance policy, or guaranty; (5) The number of the bond, insurance policy or guaranty; and (6) Where...

  1. Biaxial loading effects on fracture toughness of reactor pressure vessel steel

    International Nuclear Information System (INIS)

    McAfee, W.J.; Bass, B.R.; Bryson, J.W. Jr.; Pennell, W.E.

    1995-03-01

    The preliminary phases of a program to develop and evaluate fracture methodologies for assessing crack-tip constraint effects on fracture toughness of reactor pressure vessel (RPV) steels have been completed by the Heavy-Section Steel Technology (HSST) Program. Objectives were to investigate effect of biaxial loading on fracture toughness, quantify this effect through existing stress-based, dual-parameter, fracture-toughness correlations, or propose and verify alternate correlations. A cruciform beam specimen with 2-D, shallow, through-thickness flaw and a special loading fixture was designed and fabricated. Tests were performed using biaxial loading ratios of 0:1 (uniaxial), 0.6:1, and 1:1 (equi-biaxial). Critical fracture-toughness values were calculated for each test. Biaxial loading of 0.6:1 resulted in a reduction in the lower bound fracture toughness of ∼12% as compared to that from the uniaxial tests. The biaxial loading of 1:1 yielded two subsets of toughness values; one agreed well with the uniaxial data, while one was reduced by ∼43% when compared to the uniaxial data. Results were evaluated using J-Q theory and Dodds-Anderson (D-A) micromechanical scaling model. The D-A model predicted no biaxial effect, while the J-Q method gave inconclusive results. When applied to the 1:1 biaxial data, these constraint methodologies failed to predict the observed reduction in fracture toughness obtained in one experiment. A strain-based constraint methodology that considers the relationship between applied biaxial load, the plastic zone width in the crack plane, and fracture toughness was formulated and applied successfully to the data. Evaluation of this dual-parameter strain-based model led to the conclusion that it has the capability of representing fracture behavior of RPV steels in the transition region, including the effects of out-of-plane loading on fracture toughness. This report is designated as HSST Report No. 150

  2. Scenario based optimization of a container vessel with respect to its projected operating conditions

    Directory of Open Access Journals (Sweden)

    Jonas Wagner

    2014-06-01

    Full Text Available In this paper the scenario based optimization of the bulbous bow of the KRISO Container Ship (KCS is presented. The optimization of the parametrically modeled vessel is based on a statistically developed operational profile generated from noon-to-noon reports of a comparable 3600 TEU container vessel and specific development functions representing the growth of global economy during the vessels service time. In order to consider uncertainties, statistical fluctuations are added. An analysis of these data lead to a number of most probable upcoming operating conditions (OC the vessel will stay in the future. According to their respective likeliness an objective function for the evaluation of the optimal design variant of the vessel is derived and implemented within the parametrical optimization workbench FRIENDSHIP Framework. In the following this evaluation is done with respect to vessel's calculated effective power based on the usage of potential flow code. The evaluation shows, that the usage of scenarios within the optimization process has a strong influence on the hull form.

  3. Sealing analysis for nuclear vessels of PWR

    International Nuclear Information System (INIS)

    Qu Jiadi; Dou Yikang

    1988-01-01

    The fundamental equations of sealing analysis for vessels are given and a computer program named SMEC, which considers the change of stud loading, the elastic contact between flange mating surfaces and the transient thermal effects, is developed accordingly. The SMEC is verified by several test. On the basis of analysis, a new concept of classifying vessels into three types according to increasing or decreasing of bolt loading with increasing pressure is suggested. Type-A vessel is that in which the bolt loading increases monotonically with increasing pressure, while in type-B, the bolt loading decreases monotonically, and in type-C, the bolt loading changes nonmonotonically. It is important for vessel design to distinguish the types through analysis. The sealing mechanism is also discussed

  4. Numerical modelling of electromagnetic loads on fusion device structures

    International Nuclear Information System (INIS)

    Bettini, Paolo; Palumbo, Maurizio Furno; Specogna, Ruben

    2014-01-01

    In magnetic confinement fusion devices, during abnormal operations (disruptions) the plasma begins to move rapidly towards the vessel wall in a vertical displacement event (VDE), producing plasma current asymmetries, vessel eddy currents and open field line halo currents, each of which can exert potentially damaging forces upon the vessel and in-vessel components. This paper presents a methodology to estimate electromagnetic loads, on three-dimensional conductive structures surrounding the plasma, which arise from the interaction of halo-currents associated to VDEs with a magnetic field of the order of some Tesla needed for plasma confinement. Lorentz forces, calculated by complementary formulations, are used as constraining loads in a linear static structural analysis carried out on a detailed model of the mechanical structures of a representative machine

  5. Numerical modelling of electromagnetic loads on fusion device structures

    Science.gov (United States)

    Bettini, Paolo; Furno Palumbo, Maurizio; Specogna, Ruben

    2014-03-01

    In magnetic confinement fusion devices, during abnormal operations (disruptions) the plasma begins to move rapidly towards the vessel wall in a vertical displacement event (VDE), producing plasma current asymmetries, vessel eddy currents and open field line halo currents, each of which can exert potentially damaging forces upon the vessel and in-vessel components. This paper presents a methodology to estimate electromagnetic loads, on three-dimensional conductive structures surrounding the plasma, which arise from the interaction of halo-currents associated to VDEs with a magnetic field of the order of some Tesla needed for plasma confinement. Lorentz forces, calculated by complementary formulations, are used as constraining loads in a linear static structural analysis carried out on a detailed model of the mechanical structures of a representative machine.

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  7. Elmo Bumpy Torus proof of principle, Phase II: Title 1 report. Volume II. Toroidal vessel

    International Nuclear Information System (INIS)

    1982-01-01

    The Toroidal Vessel provides the vacuum enclosure for containing the high temperature steady state plasma. In addition, the Toroidal Vessel must provide several viewing ports for plasma diagnostics, vacuum pumping ports for both high vacuum and roughing vacuum, feed-through ports for ECRH waveguides, limiter feed throughs for cooling and supporting the limiters, and ports for ion gages. The vessel must operate in an intense environment comprised of x-rays, microwaves, magnetic fields and plasma heat loads as well as the atmosphere pressure and gravity loads and the internal thermal stress loads due to heating and cooling of the torus. A key issue addressed was the choice of vacuum vessel seal and wall materials. In addition, during the course of the study, ORNL requested that horsecollar diagnostic ports be incorporated in the design. A comprehensive trade study was performed considering the vessel material issues in concert with the impact of the horsecollar port design. A change in baseline from an aluminum vessel with elastomer seals and circular diagnostic ports to austenitic stainless steel vessel with metal seals and horsecollar ports was agreed upon by both MDAC and ORNL towards the end of Title I

  8. After-operating properties of nuclear reactor vessel materials of Lenin atomic ice breaker and prospective of reactor vessels radiation life prolongation

    International Nuclear Information System (INIS)

    Platonov, P.A.; Shtrombakh, Ya.I.; Amaev, A.D.; Krasikov, E.A.; Korolev, Yu.N.; Zabusov, O.O.; Glushakov, G.M.

    2001-01-01

    A post-operational state of the icebreaker Lenin reactor vessel metal is investigated. It is shown that a base metal of the icebreaker Lenin reactor vessel is of high quality as by an initial value of critical temperature of embrittlement, so by its radiation resistance. The weld metal possesses a sufficient radiation resistance but has an insufficient initial ductile-brittle transition temperature (approximately 63 Deg C). It is necessary to note that the final stage of operation for nuclear steam-generating plant should be carried out at the coolant temperature as high as possible [ru

  9. Hydraulic nuts (HydraNuts) for reactor vessel tensioning

    International Nuclear Information System (INIS)

    Greenwell, Steve

    2008-01-01

    The paper will present how the introduction of hydraulic nuts - HydraNuts, has reduced critical path times, dose exposure for workers and improved working safety conditions around the reactor vessel during tensioning or de-tensioning operations. It will focus upon detailing the advantages realized by utilities that have introduced the technology and providing examples of the improvements made to the process as well as discussing the engineering design change packages required to make the conversion to the new system. HydraNuts replace the traditional mechanical nut/stud tensioning equipment, combining the two functions into a single system, designed for easy installation and operation by one individual. The primary components of the HydraNut can be assembled without the need for external crane or hoist support and are designed so that each sub assembly can be fitted separately. Once all HydraNuts are fitted to the Rx vessel studs and are sitting on the main Rx vessel head flange, then a system of flexible hydraulic hoses is connected to them, forming a closed loop hydraulic harness, which will allow for simultaneous pressurization of all HydraNuts. Hydraulic pressure is obtained by the use of a hydraulic pumping unit and the resultant load generated in each HydraNut is transferred to the stud and main flange closure is obtained. While maintaining hydraulic pressure, a locking ring is rotated into place on the HydraNut assembly that will support the tensioned load mechanically when the hydraulic pressure is released from the hose harness assembly. The hose harness is removed and the HydraNut is now functioning as a mechanical nut retaining the tensioned load. The HydraNut system for Rx vessel applications was first introduced into a plant in the U.S. in October 2006 and based upon the benefits realized subsequent projects are under way within the Asian and U.S. operating fleet. (author)

  10. Advanced toroidal facility vaccuum vessel stress analyses

    International Nuclear Information System (INIS)

    Hammonds, C.J.; Mayhall, J.A.

    1987-01-01

    The complex geometry of the Advance Toroidal Facility (ATF) vacuum vessel required special analysis techniques in investigating the structural behavior of the design. The response of a large-scale finite element model was found for transportation and operational loading. Several computer codes and systems, including the National Magnetic Fusion Energy Computer Center Cray machines, were implemented in accomplishing these analyses. The work combined complex methods that taxed the limits of both the codes and the computer systems involved. Using MSC/NASTRAN cyclic-symmetry solutions permitted using only 1/12 of the vessel geometry to mathematically analyze the entire vessel. This allowed the greater detail and accuracy demanded by the complex geometry of the vessel. Critical buckling-pressure analyses were performed with the same model. The development, results, and problems encountered in performing these analyses are described. 5 refs., 3 figs

  11. Effect of radiation damage on operating safety of steel pressure vessels of nuclear reactors

    International Nuclear Information System (INIS)

    Vacek, M.; Havel, S.; Stoces, B.; Brumovsky, M.

    1980-01-01

    The effects are assessed of the environment upon mechanical properties of steel used generally for pressure vessels of light water nuclear reactors. Changes caused by radiation affect the reliability of vessels. Deterioration of steel properties is mainly due to neutron radiation. The article deals with factors bearing upon damage and with methods allowing to evaluate the reliability of vessels and predict their service life. Operating reliability of vessels is very unfavourably affected by planned and accidental reactor transients. (author)

  12. Variable flaw shape analysis for a reactor vessel under pressurized thermal shock loading

    International Nuclear Information System (INIS)

    Yang, C.Y.; Bamford, W.H.

    1984-01-01

    A study has been conducted to characterize the response of semi-elliptic surface flaws to thermal shock conditions which can result from safety injection actuation in nuclear reactor vessels. A methodology was developed to predict the behavior of a flaw during sample pressurized thermal shock events. The effects of a number of key variables on the flaw propagation were studied, including fracture toughness of the material and its gradient through the thickness, irradiation effects, effects of warm prestressing, and effects of the stainless steel cladding. The results of these studies show that under thermal shock loading conditions the flaw always tends to elongate along the vessel inside surface from the initial aspect ratio. However, the flaw shape always remains finite rather than becoming continuously long, as has often been assumed in earlier analyses. The final shape and size of the flaws were found to be rather strongly dependent on the effects of warm prestressing and the distribution of neutron flux. The improved methodology results in a more accurate and more realistic treatment of flaw shape changes during thermal shock events and provides the potential for quantifying additional margins for reactor vessel integrity analyses

  13. German boiler and pressure vessel codes and standards: materials, manufacture, testing, equipment, erection and operation

    International Nuclear Information System (INIS)

    Steffen, H.P.

    1987-01-01

    The methods by which the safety objectives on the operation of steam boilers and pressure vessels in Germany can be reached are set out in Technical Rules which are compiled and established in technical committees. Typical applications are described in the Technical Rules. A chart shows how the laws, provisions and Technical Rules for the sections 'steam boiler plant' and 'pressure vessels' are interlinked. This chapter concentrates on legal aspects, materials, manufacture, testing, erection and operation of boilers and pressure vessels in Germany. (U.K.)

  14. Improvement of daily load-following operation for boiling water reactors

    International Nuclear Information System (INIS)

    Kiguchi, Takashi; Kurihara, Kunitoshi; Sakurai, Mikio; Joge, Toshio; Asami, Kazuo.

    1980-01-01

    Recently, with the increase of the proportion of nuclear power generation to the total amount of power generation of electric power systems, the needs of daily load-following operation of nuclear power stations have heightened, accordingly the study on the method of daily load-following operation has been carried out for BWRs. In this study, by the combined use of the flow rate control of core coolnat being operated easily and the operation of control rods, the BWR system with the daily load-following performance of 100% power output in daytime and 50% power output at night was the target of development. For the purpose, the change of core characteristics during load-following was grasped analytically, and the range of load change was investigated. At the same time, as the first stage of developing operation control and monitoring system, the reactor output-adjusting device which makes generator output automatically follow the target load change pattern by the flow rate control of core coolnat, and the equipment for monitoring core performance on line were developed. The analysis of the method of daily load-following operation in present-day BWRs, the study on the improvement of load-following operation performance, the reactor output-adjusting device are described. (Kako, I.)

  15. Full vessel CFD analysis on thermal-hydraulic characteristics of CPR1000 PWR

    International Nuclear Information System (INIS)

    Chao Yanmeng; Yang Lixin; Zhang Mingqian

    2014-01-01

    To obtain flow distributions and thermal-hydraulic properties in a full vessel PWR under limited computation ability and time, a full vessel simulation model of CPR1000 was built based on two simplification methods. One simplified the inner geometry of the control rod guide tubes using equivalent flow area. Another substituted the core by a porous domain to maintain the pressure drop and temperature rise. After the computation, global and localized flow distributions, hydraulic loads of some main assemblies were obtained, as well as other thermal-hydraulic properties. The results indicate the flow distribution in the full vessel is asymmetrical. Therefore it is essential to use the full vessel model to simulate. The calculated thermal-hydraulic characteristics agree well with the operation statistics, providing the reference data for the reactor safety operation. (authors)

  16. Seismic transient analysis of a containment vessel with penetrations

    International Nuclear Information System (INIS)

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

    1979-12-01

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

  17. Mitigating the Long term Operating Extreme Load through Active Control

    International Nuclear Information System (INIS)

    Koukoura, Christina; Natarajan, Anand

    2014-01-01

    The parameters influencing the long term extreme operating design loads are identified through the implementation of a Design of Experiment (DOE) method. A function between the identified critical factors and the ultimate out-of-plane loads on the blade is determined. Variations in the initial blade azimuth location are shown to affect the extreme blade load magnitude during operation in normal turbulence wind input. The simultaneously controlled operation of generator torque variation and pitch variation at low blade pitch angles is detected to be responsible for very high loads acting on the blades. Through gain scheduling of the controller (modifications of the proportional Kp and the integral K gains) the extreme loads are mitigated, ensuring minimum instantaneous variations in the power production for operation above rated wind speed. The response of the blade load is examined for different values of the integral gain as resulting in rotor speed error and the rate of change of rotor speed. Based on the results a new load case for the simulation of extreme loads during normal operation is also presented

  18. Scenario based optimization of a container vessel with respect to its projected operating conditions

    Directory of Open Access Journals (Sweden)

    Wagner Jonas

    2014-06-01

    Full Text Available In this paper the scenario based optimization of the bulbous bow of the KRISO Container Ship (KCS is presented. The optimization of the parametrically modeled vessel is based on a statistically developed operational profile generated from noon-to-noon reports of a comparable 3600 TEU container vessel and specific development functions representing the growth of global economy during the vessels service time. In order to consider uncertainties, statistical fluctuations are added. An analysis of these data lead to a number of most probable upcoming operating conditions (OC the vessel will stay in the future. According to their respective likeliness an objective function for the evaluation of the optimal design variant of the vessel is derived and implemented within the parametrical optimization workbench FRIENDSHIP Framework. In the following this evaluation is done with respect to vessel’s calculated effective power based on the usage of potential flow code. The evaluation shows, that the usage of scenarios within the optimization process has a strong influence on the hull form.

  19. Remote maintenance of in-vessel components in Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Loesser, G.D.; Heitzenroeder, P.; Kungl, D.; Dylla, H.F.; Cerdan, G.

    1990-01-01

    The Tokamak Fusion Test Reactor (TFTR) will generate a total of 3 x 10 21 neutrons during its planned D-T operational period. A maintenance manipulator has been designed and tested to minimize personnel radiation during in-vessel maintenance activities. Its functions include visual inspection, first-wall tile replacement, cleaning, diagnostics calibrations and leak detection. To meet these objectives, the TFTR maintenance manipulator is required to be operable in the TFTR high vacuum environment, typically -8 torr, ( -6 Pa). Geometrically, the manipulator must extend 180 0 in either direction around the torus to assure complete coverage of the vessel first-wall. The manipulator consists of a movable carriage, and movable articulated link sections which are driven by electrical actuators. The boom has vertical load capacity of 455 kg and lateral load capacity of 46 kg. The boom can either be fitted with a general inspection arm or dextrous slave arms. The general inspection arm is designed to hold the leak detector and an inspection camera; it is capable of rotation along two axes and has a linkage system which permits motion normal to the vacuum vessel wall. All systems except the dextrous slave arms are operable in a vacuum. (author)

  20. Vessel Segmentation in Retinal Images Using Multi-scale Line Operator and K-Means Clustering.

    Science.gov (United States)

    Saffarzadeh, Vahid Mohammadi; Osareh, Alireza; Shadgar, Bita

    2014-04-01

    Detecting blood vessels is a vital task in retinal image analysis. The task is more challenging with the presence of bright and dark lesions in retinal images. Here, a method is proposed to detect vessels in both normal and abnormal retinal fundus images based on their linear features. First, the negative impact of bright lesions is reduced by using K-means segmentation in a perceptive space. Then, a multi-scale line operator is utilized to detect vessels while ignoring some of the dark lesions, which have intensity structures different from the line-shaped vessels in the retina. The proposed algorithm is tested on two publicly available STARE and DRIVE databases. The performance of the method is measured by calculating the area under the receiver operating characteristic curve and the segmentation accuracy. The proposed method achieves 0.9483 and 0.9387 localization accuracy against STARE and DRIVE respectively.

  1. Engineering analysis of ITER In-Vessel Viewing System guide tube

    Energy Technology Data Exchange (ETDEWEB)

    Casal, Natalia, E-mail: natalia.casal@iter.org [ITER Organization, Route de Vinon sur Verdon, St Paul-lez-Durance (France); Bates, Philip [Fusion for Energy, Barcelona (Spain); Bede, Ottó [Oxford Technologies Ltd., Abingdon (United Kingdom); Damiani, Carlo; Dubus, Gregory [Fusion for Energy, Barcelona (Spain); Omran, Hassan [Oxford Technologies Ltd., Abingdon (United Kingdom); Palmer, Jim [ITER Organization, Route de Vinon sur Verdon, St Paul-lez-Durance (France); Puiu, Adrian [Fusion for Energy, Barcelona (Spain); Reichle, Roger; Suárez, Alejandro; Walker, Christopher; Walsh, Michael [ITER Organization, Route de Vinon sur Verdon, St Paul-lez-Durance (France)

    2015-10-15

    Highlights: • Conceptual design of IVVS Loads action on IVVS Dominant loads. • Seismic and baking conditions. • No active cooling needed for IVVS. • IVVS requires additional support points to avoid excessive deformation. - Abstract: The In Vessel Viewing System (IVVS) will be one of the essential machine diagnostic systems at ITER to provide information about the status of in-vessel and plasma facing components and to evaluate the dust inside the Vacuum Vessel. The current design consists of six scanning probes and their deployment systems, which are placed in dedicated ports at the divertor level. These units are located in resident guiding tubes 10 m long, which allow the IVVS probes to go from their storage location to the scanning position by means of a simple straight translation. Moreover, each resident tube is supported inside the corresponding Vacuum Vessel and Cryostat port extensions, which are part of the primary confinement barrier. As the Vacuum Vessel and the Cryostat will move with respect to each other during operation (especially during baking) and during incidents and accidents (disruptions, vertical displacement events, seismic events), the structural integrity of the resident tube and the surrounding vacuum boundaries would be compromised if the required flexibility and supports are not appropriately assured. This paper focuses on the integration of the present design of the IVVS into the Vacuum Vessel and Cryostat environment. It presents the adopted strategy to withstand all the main interfacing loads without damaging the confinement barriers and the corresponding analysis supporting it.

  2. Evaluation of SMART load follow operation capability

    International Nuclear Information System (INIS)

    Song, Jae Seung; Zee, Sung Quun

    1998-07-01

    The daily load follow operation capability of SMART, which is 330 MWth integral reactor with boron free operation concept, was evaluated. In the boron free operation core the axial offset(AO) can not be limited or limited by wider range than the typical PWR that uses soluble boron. Current SMART conceptual core design does not have the limit for the axial offset and the operation is limited by three dimensional local power peak. In this report, the capability of load follow operation is evaluated for 14-2-6-2 daily load follow by 50% power reduction that is very typical in the large commercial power reactors. The cycle length of SMART is limited by three dimensional local power peak and the nominal axial power distribution is top skewed at the end of cycle. When the power returns to 100%, control rod should be withdrawn for the compensation of reactivity decrease due to the xenon buildup. The control rod withdrawal results in the increase of three dimensional local power peak. To solve this problem, an operation strategy for the control of local peak. To solve this problem, an operation strategy for the control of local peak was established and it was shown that the strategy is effective in controlling the local peak less than target value of the three dimensional local power peak. (author). 7 refs., 5 tabs., 13 figs

  3. Umbilical Cable Recovery Load Analysis

    Institute of Scientific and Technical Information of China (English)

    YAN Shu-wang; JIA Zhao-lin; FENG Xiao-wei; LI Shi-tao

    2013-01-01

    Umbilical cable is a kind of integrated subsea cable widely used in the exploration and exploitation of oil and gas field.The severe ocean environment makes great challenges to umbilical maintenance and repair work.Damaged umbilical is usually recovered for the regular operation of the offshore production system.Analysis on cables in essence is a two-point boundary problem.The tension load at the mudline must be known first,and then the recovery load and recovery angle on the vessel can be solved by use of catenary equation.The recovery analysis also involves umbilicalsoil interaction and becomes more complicated.Calculation methods for recovery load of the exposed and buried umbilical are established and the relationship between the position of touch down point and the recovery load as well as the recovery angle and recovery load are analyzed.The analysis results provide a theoretical reference for offshore on-deck operation.

  4. Changes in the vessels following aorto-coronary bypass operation

    International Nuclear Information System (INIS)

    Goebel, N.; Pfluger, N.; Speiser, K.; Turina, M.; Rothlin, M.; Zurich Univ.; Zurich Univ.

    1983-01-01

    In a prospective study (238 men, mean age 53 years) the changes of the native vessels were studied 3 months after a-c-bypass operation and 5 months after preop. angiography. Progression was defined as increase of stenoses of at least 20% or new total occlusion. Progression was significantly more frequent in vessels with than without bypass and was located proximally to the anastomoses in most cases, less frequently at the anastomoses and very rarely distally to the anastomoses. Proximal progression was significantly more frequent with patent than with occluded bypasses. Stenoses at the anastomoses were significantly more frequent with occluded than with patent bypasses. Stenoses of higher degrees hat a stonger tendency for progression than slighter stenoses. Regression was rare and nearly always caused by surgery. (orig.) [de

  5. Design, Analysis and R&D of the EAST In-Vessel Components

    Science.gov (United States)

    Yao, Damao; Bao, Liman; Li, Jiangang; Song, Yuntao; Chen, Wenge; Du, Shijun; Hu, Qingsheng; Wei, Jing; Xie, Han; Liu, Xufeng; Cao, Lei; Zhou, Zibo; Chen, Junling; Mao, Xinqiao; Wang, Shengming; Zhu, Ning; Weng, Peide; Wan, Yuanxi

    2008-06-01

    In-vessel components are important parts of the EAST superconducting tokamak. They include the plasma facing components, passive plates, cryo-pumps, in-vessel coils, etc. The structural design, analysis and related R&D have been completed. The divertor is designed in an up-down symmetric configuration to accommodate both double null and single null plasma operation. Passive plates are used for plasma movement control. In-vessel coils are used for the active control of plasma vertical movements. Each cryo-pump can provide an approximately 45 m3/s pumping rate at a pressure of 10-1 Pa for particle exhaust. Analysis shows that, when a plasma current of 1 MA disrupts in 3 ms, the EM loads caused by the eddy current and the halo current in a vertical displacement event (VDE) will not generate an unacceptable stress on the divertor structure. The bolted divertor thermal structure with an active cooling system can sustain a load of 2 MW/m2 up to a 60 s operation if the plasma facing surface temperature is limited to 1500 °C. Thermal testing and structural optimization testing were conducted to demonstrate the analysis results.

  6. Simulation of In-Vessel Corium Retention through External Reactor Vessel Cooling for SMART using SIMPLE

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jin-Sung; Son, Donggun; Park, Rae-Joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Thermal load analysis from the corium pool to the outer reactor vessel in the lower plenum of the reactor vessel is necessary to evaluate the effect of the IVR-ERVC during a severe accident for SMART. A computational code called SIMPLE (Sever Invessel Melt Progression in Lower plenum Environment) has been developed for analyze transient behavior of molten corium in the lower plenum, interaction between corium and coolant, and heat-up and ablation of reactor vessel wall. In this study, heat load analysis of the reactor vessel for SMART has been conducted using the SIMPLE. Transient behavior of the molten corium in the lower plenum and IVR-ERVC for SMART has been simulated using SIMPLE. Heat flux from the corium pool to the outer reactor vessel is concentrated in metallic layer by the focusing effect. As a result, metallic layer shows higher temperature than the oxidic layer. Also, vessel wall of metallic layer has been ablated by the high in-vessel temperature. Ex-vessel temperature of the metallic layer was maintained 390 K and vessel thickness was maintained 14 cm. It means that the reactor vessel integrity is maintained by the IVR-ERVC.

  7. Electromagnetic loads and structural response of the CIT [Compact Ignition Tokamak] vacuum vessel to plasma disruptions

    International Nuclear Information System (INIS)

    Salem, S.L.; Listvinsky, G.; Lee, M.Y.; Bailey, C.

    1987-01-01

    Studies of the electromagnetic loads produced by a variety of plasma disruptions, and the resulting structural effects on the compact Ignition Tokamak (CIT) vacuum vessel (VV), have been performed to help optimize the VV design. A series of stationary and moving plasmas, with disruption rates from 0.7--10.0 MA/ms, have been analyzed using the EMPRES code to compute eddy currents and electromagnetic pressures, and the NASTRAN code to evaluate the structural response of the vacuum vessel. Key factors contributing to the magnitude of EM forces and resulting stresses on the vessel have been found to include disruption rate, and direction and synchronization of plasma motion with the onset of plasma current decay. As a result of these analyses, a number of design changes have been made, and design margins for the present 1.75 meter design have been improved over the original CIT configuration. 1 ref., 10 figs., 4 tabs

  8. Mitigating the Long term Operating Extreme Load through Active Control

    DEFF Research Database (Denmark)

    Koukoura, Christina; Natarajan, Anand

    2014-01-01

    blade azimuth location are shown to affect the extreme blade load magnitude during operation in normal turbulence wind input. The simultaneously controlled operation of generator torque variation and pitch variation at low blade pitch angles is detected to be responsible for very high loads acting...... on the blades. Through gain scheduling of the controller (modifications of the proportional Kp and the integral Ki gains) the extreme loads are mitigated, ensuring minimum instantaneous variations in the power production for operation above rated wind speed. The response of the blade load is examined...

  9. Dynamic analysis of crack growth and arrest in a pressure vessel subjected to thermal and pressure loading

    International Nuclear Information System (INIS)

    Brickstad, B.

    1984-01-01

    Predictions of crack arrest behaviour are performed for a cracked reactor pressure vessel under both thermal and pressure loading. The object is to compare static and dynamic calculations. The dynamic calculations are made using an explicit finite element technique where crack growth is simulated by gradual nodal release. Three different load cases and the effect of different velocity dependence on the crack propagation toughness are studied. It is found that for the analysed cases the static analysis is slightly conservative, thus justifying its use for these problems. (orig.)

  10. 50 CFR 216.46 - U.S. citizens on foreign flag vessels operating under the International Dolphin Conservation...

    Science.gov (United States)

    2010-10-01

    ... 50 Wildlife and Fisheries 7 2010-10-01 2010-10-01 false U.S. citizens on foreign flag vessels operating under the International Dolphin Conservation Program. 216.46 Section 216.46 Wildlife and Fisheries....46 U.S. citizens on foreign flag vessels operating under the International Dolphin Conservation...

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  12. Load Management in District Heating Operation

    OpenAIRE

    Li, Hongwei; Wang, Stephen Jia

    2015-01-01

    Smooth operation of district heating system will avoid installation of expensive peak heat boilers, improve plant partial load performance, increase the system redundancy for further network expansion and improve its resilience to ensuresecurity of supply during severe heating seasons. The peak heating load can be reduced through building demand side management. The building thermal mass can be used to shift the heating supply under the circumstance withoutjeopardizing the consumer thermal co...

  13. Pressure vessels for reactors made from structural steel with limited tensile strength

    International Nuclear Information System (INIS)

    Machatti, H.

    1973-01-01

    The reactor pressure vessel is prestressed in several directions with prestressing elements fabricated of steel with a high yielding point. This design allows a substantial reduction of wall thickness or an increase of the inner diameter at equal wall thickness. The prestress of the prestressing elements is designed to achieve a maximum stress release of the vessel walls at normal operating conditions and to fully utilize the maximum load of the vessel walls. For safety reasons the cross section of the prestressing elements is constructed in a way that strain is always 20 % lower the yield point. (P.K.)

  14. Dynamic buckling and nonlinear response of FBR main vessels under earthquake loading

    International Nuclear Information System (INIS)

    Hagiwara, Yutaka; Kawamoto, Yoji; Nakagawa, Masaki; Akiyama, Hiroshi.

    1991-01-01

    Pseudo-dynamic tests of cylindrical shells under high temperature were performed in order to study elasto-plastic shear-bending buckling and the nonlinear response of FBR main vessels under earthquake loading. The test results showed a response reduction effect due to pre-buckling plasticity, and a large seismic margin due to post-buckling energy absorption of the cylinders. A simple expression of the response reduction effect was proposed, as a contribution to the safe and effective seismic design of FBRs. Two methods for seismic margin evaluation were also proposed, and it was shown that appropriate seismic margins can be ensured, when the response reduction effect is incorporated into the seismic design. (author)

  15. Aging impact on the safety and operability of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Pennell, W.E.

    1992-01-01

    Irradiation embrittlement causes a loss of reactor vessel material fracture toughness as nuclear plants age. Fracture mechanics based regulatory requirements limit the permissible level of irradiation embrittlement such that essential fracture prevention margins are maintained throughout the plant operating life. This paper reviews the regulatory requirements and the underlying fracture mechanics technology. Issues identified with that technology are identified and research programs implemented to resolve the issues are described. Where possible, an assessment is given of the anticipated impact on the research program output will have on the reactor vessel fracture-margin assessment process

  16. Loading/unloading buoy. Laste/lossebye

    Energy Technology Data Exchange (ETDEWEB)

    Breivik, K.; Smedal, A.; Syvertsen, K.

    1994-07-04

    The invention relates to a buoy for use in loading or unloading of a flowable medium, especially oil. The buoy is at its lower end arranged for connection to at least one transfer line and further being arranged to be introduced into a submerged downwardly open receiving space in a floating vessel. The buoy forms in operation a transfer connection between the transfer line and a tube system on the vessel. The buoy comprises an outer buoyancy member arranged for releasable locking to the receiving space of the vessel by means of a locking mechanism arranged therein, and centrally in the outer member a rotatably mounted member which forms a passage for medium and at its ends is arranged for connection to the transfer line and the tube system on the vessel, respectively. The buoy at its upper end is connected to a means for hoisting and introducing the buoy into the receiving space of the vessel. 8 figs.

  17. Prestressed reactor vessel for nuclear power plants

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.

    1982-01-01

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

  18. 46 CFR 117.204 - Survival craft-vessels operating on coastwise routes.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Survival craft-vessels operating on coastwise routes... PASSENGERS LIFESAVING EQUIPMENT AND ARRANGEMENTS Number and Type of Survival Craft § 117.204 Survival craft... allowed, the following survival craft requirements apply when not engaged in an overnight voyage: (1...

  19. Flexible Composite-Material Pressure Vessel

    Science.gov (United States)

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

    2003-01-01

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

  20. Conjugate heat transfer analysis for in-vessel retention with external reactor vessel cooling

    International Nuclear Information System (INIS)

    Park, Jong-Woon; Bae, Jae-ho; Song, Hyuk-Jin

    2016-01-01

    Highlights: • A conjugate heat transfer analysis method is applied for in-vessel corium retention. • 3D heat diffusion has a formidable effect in alleviating focusing heat load from metallic layer. • The focusing heat load is decreased by about 2.5 times on the external surface. - Abstract: A conjugate heat transfer analysis method for the thermal integrity of a reactor vessel under external reactor vessel cooling conditions is developed to resolve light metal layer focusing effect issue for in-vessel retention. The method calculates steady-state three-dimensional temperature distribution of a reactor vessel using coupled conjugate heat transfer between in-vessel three-layered stratified corium (metallic pool, oxide pool and heavy metal and polar-angle dependent boiling heat transfer at the outer surface of a reactor vessel). The three-layer corium heat transfer model is utilizing lumped-parameter thermal-resistance circuit method. For the ex-vessel boiling boundary conditions, nucleate, transition and film boiling are considered. The thermal integrity of a reactor vessel is addressed in terms of heat flux at the outer-most nodes of the vessel and remaining thickness profile. The vessel three-dimensional heat conduction is validated against a commercial code. It is found that even though the internal heat flux from the metal layer goes far beyond critical heat flux (CHF) the heat flux from the outermost nodes of the vessel may be maintained below CHF due to massive vessel heat diffusion. The heat diffusion throughout the vessel is more pronounced for relatively low heat generation rate in an oxide pool. Parametric calculations are performed considering thermal conditions such as peak heat flux from a light metal layer, heat generation in an oxide pool and external boiling conditions. The major finding is that the most crucial factor for success of in-vessel retention is not the mass of the molten light metal above the oxide pool but the heat generation rate

  1. Gigacycle fatigue behaviour of austenitic stainless steels used for mercury target vessels

    International Nuclear Information System (INIS)

    Naoe, Takashi; Xiong, Zhihong; Futakawa, Masatoshi

    2016-01-01

    A mercury enclosure vessel for the pulsed spallation neutron source manufactured from a type 316L austenitic stainless steel, a so-called target vessel, suffers the cyclic loading caused by the proton beam induced pressure waves. A design criteria of the JSNS target vessel which is defined based on the irradiation damage is 2500 h at 1 MW with a repetition rate of 25 Hz, that is, the target vessel suffers approximately 10 9 cyclic loading while in operation. Furthermore, strain rate of the beam window of the target vessel reaches 50 s −1 at the maximum, which is much higher than that of the conventional fatigue. Gigacycle fatigue strength up to 10 9 cycles for solution annealed 316L (SA) and cold-worked 316L (CW) were investigated through the ultrasonic fatigue tests. Fatigue tests were performed under room temperature and 250 °C which is the maximum temperature evaluated at the beam window in order to investigate the effect of temperature on fatigue strength of SA and CW 316L. The results showed that the fatigue strength at 250 °C is clearly reduced in comparison with room temperature, regardless of cold work level. In addition, residual strength and microhardness of the fatigue tested specimen were measured to investigate the change in mechanical properties by cyclic loading. Cyclic hardening was observed in both the SA and CW 316L, and cyclic softening was observed in the initial stage of cyclic loading in CW 316L. Furthermore, abrupt temperature rising just before fatigue failure was observed regardless of testing conditions.

  2. Structural analysis of the JT-60SA cryostat vessel body

    Energy Technology Data Exchange (ETDEWEB)

    Botija, José, E-mail: jose.botija@ciemat.es [Association EURATOM – CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Alonso, Javier; Fernández, Pilar; Medrano, Mercedes; Ramos, Francisco; Rincon, Esther; Soleto, Alfonso [Association EURATOM – CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Davis, Sam; Di Pietro, Enrico; Tomarchio, Valerio [Fusion for Energy, JT-60SA European Home Team, 85748 Garching bei Munchen (Germany); Masaki, Kei; Sakasai, Akira; Shibama, Yusuke [JAEA – Japan Atomic Energy Agency, Naka Fusion Institute, Ibaraki 311-0193 (Japan)

    2013-10-15

    Highlights: ► Structural analysis to validate the JT-60SA cryostat vessel body design. ► Design code ASME 2007 “Boiler and Pressure Vessel Code. Section VIII”. ► First buckling mode: load multiplier of 10.644, higher than the minimum factor 4.7. ► Elastic and elastic–plastic stress analysis meets ASME against plastic collapse. ► Bolted fasteners have been analyzed showing small gaps closed by strong welding. -- Abstract: The JT-60SA cryostat is a stainless steel vacuum vessel (14 m diameter, 16 m height) which encloses the Tokamak providing the vacuum environment (10{sup −3} Pa) necessary to limit the transmission of thermal loads to the components at cryogenic temperature. It must withstand both external atmospheric pressure during normal operation and internal overpressure in case of an accident. The paper summarizes the structural analyses performed in order to validate the JT-60SA cryostat vessel body design. It comprises several analyses: a buckling analysis to demonstrate stability under the external pressure; an elastic and an elastic–plastic stress analysis according to ASME VIII rules, to evaluate resistance to plastic collapse including localized stress concentrations; and, finally, a detailed analysis with bolted fasteners in order to evaluate the behavior of the flanges, assuring the integrity of the vacuum sealing welds of the cryostat vessel body.

  3. ITER vacuum vessel dynamic stress analysis of a disruption

    International Nuclear Information System (INIS)

    Riemer, B.W.; Conner, D.L.; Strickler, D.J.; Williamson, D.E.

    1994-01-01

    Dynamic stress analysis of the International Thermonuclear Experimental Reactor vacuum vessel loaded by disruption forces was performed. The deformation and stress results showed strong inertial effects when compared to static analyses. Maximum stress predicted dynamically was 300 MPa, but stress shown by static analysis from loads at the same point in time reached only 80 MPa. The analysis also provided a reaction load history in the vessel's supports which is essential in evaluating support design. The disruption forces were estimated by assuming a 25-MA plasma current decaying at 1 MA/ms while moving vertically. In addition to forces developed within the vessel, vertical loadings from the first wall/strong back assemblies and the divertor were applied to the vessel at their attachment points. The first 50 natural modes were also determined. The first mode's frequency was 6.0 Hz, and its shape is characterized by vertical displacement of the vessel inner leg. The predicted deformation of the vessel appeared similar to its first mode shape combined with radial contraction. Kinetic energy history from the analysis also correlated with the first mode frequency

  4. The procurement and testing of the stainless steel in-vessel panels of the Wendelstein 7-X Stellarator

    International Nuclear Information System (INIS)

    Peacock, A.; Girlinger, A.; Vorkoeper, A.; Boscary, J.; Greuner, H.; Hurd, F.; Mendelevitch, B.; Pirsch, H.; Stadler, R.; Zangl, G.

    2011-01-01

    320 In-vessel water cooled stainless steel panels, poloidal closure plates and pumping gap panels, covering an area of approximately 100 m 2 , are used in Wendelstein7-X to protect the plasma vessel. The panels are manufactured at Deggendorf, Germany by MAN Diesel and Turbo SE. The panels consist of a laser welded sandwich of stainless steel plates together with a labyrinth of cooling channels and have a complicated geometry to fit the plasma vessel of Wendelstein 7-X. The hydraulic and mechanical stability requirements whilst maintaining the tight tolerances for the shape of the components are very demanding. The panels are designed to operate at up to an average heat load of 100 kW/m 2 and a maximum heat load of 200 kW/m 2 with a water velocity of approximately 2 m s -1 . High heat flux testing of an un-cooled panel at a time averaged load of 200 kW/m 2 for 10 s were successfully performed to support the start up phase of Wendelstein 7-X operation. Extensive testing both during manufacture and after delivery to IPP-Garching demonstrates the suitability of the delivered panels for their purpose.

  5. Mooring system for a permanently moored storage vessel at an offshore site

    Energy Technology Data Exchange (ETDEWEB)

    Flory, J.F.

    1983-01-24

    A vessel, e.g. a storage vessel, is permanently moored by means such as a yoke pivoted on the forecastle of the vessel to a mooring leg, e.g. a riser or anchor chain, which is attached to a base located on the ocean floor. Mounted on the vessel are tension, exerting means e.g. counterweights, springs, winches, etc., operably connected with the mooring leg for applying tension e.g. by lifting the yoke. The top of the mooring leg is connected to the end of the yoke through a mooring swivel and gimbaled mooring table or a universal joint. A fluid swivel may be located above the mooring table or about a load-carrying shaft connected to the mooring leg.

  6. 77 FR 62247 - Dynamic Positioning Operations Guidance for Vessels Other Than Mobile Offshore Drilling Units...

    Science.gov (United States)

    2012-10-12

    ... Operations Guidance for Vessels Other Than Mobile Offshore Drilling Units Operating on the U.S. Outer... ``Mobile Offshore Drilling Unit Dynamic Positioning Guidance''. The notice recommended owners and operators of Mobile Offshore Drilling Units (MODUs) follow Marine Technology Society (MTS) Dynamic Positioning...

  7. 46 CFR 170.095 - Data submittal for a vessel equipped to lift.

    Science.gov (United States)

    2010-10-01

    ... vessel is engaged in lifting and is required to comply with subpart B of part 173 of this chapter: (a) A graph of maximum hook load versus maximum crane radius. (b) A table of crane radius versus the maximum... transverse moments at which the crane is to operate. ...

  8. The characterization of secondary lithium-ion battery degradation when operating complex, ultra-high power pulsed loads

    Science.gov (United States)

    Wong, Derek N.

    The US Navy is actively developing all electric fleets, raising serious questions about what is required of onboard power supplies in order to properly power the ship's electrical systems. This is especially relevant when choosing a viable power source to drive high power propulsion and electric weapon systems in addition to the conventional loads deployed aboard these types of vessels. Especially when high pulsed power loads are supplied, the issue of maintaining power quality becomes important and increasingly complex. Conventionally, a vessel's electrical power is generated using gas turbine or diesel driven motor-generator sets that are very inefficient when they are used outside of their most efficient load condition. What this means is that if the generator is not being utilized continuously at its most efficient load capacity, the quality of the output power may also be effected and fall outside of the acceptable power quality limits imposed through military standards. As a solution to this potential problem, the Navy has proposed using electrochemical storage devices since they are able to buffer conventional generators when the load is operating below the generator's most efficient power level or able to efficiently augment a generator when the load is operating in excess of the generator's most efficient power rating. Specifically, the US Navy is interested in using commercial off-the-shelf (COTS) lithium-ion batteries within an intelligently controlled energy storage module that could act as either a prime power supply for on-board pulsed power systems or as a backup generator to other shipboard power systems. Due to the unique load profile of high-rate pulsed power systems, the implementation of lithium-ion batteries within these complex systems requires them to be operated at very high rates and the effects these things have on cell degradation has been an area of focus. There is very little published research into the effects that high power transient

  9. Design and construction of Alborz tokamak vacuum vessel system

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  10. Assuring reliability of unconventional weld joint configurations in austenitic stainless steel pressure vessels through non-destructive examination

    International Nuclear Information System (INIS)

    Jayakumar, I.; Manimohan, M.; Chandrasekaran, G.V.; Abdul Majeeth, S.; Subrahmanyam, P.S.

    1996-01-01

    Design of weld configurations in engineering structures is based on NDE inspectability apart from other considerations. They are mostly standardised. This paper deals with the development of an effective NDE methodology for an unconventional weld joint configuration occurring in a critical pressure vessel with edge preparation orientations different from that normally encountered in fabrication of such vessels. It is K-type butt joint between a heavy load bearing member and a curved vessel wall resulting in an oblique fillet weld. The heavy load bearing functional requirement needs a high integrity fail safe joint during its operating life and the stringent quality level specified by customer was ensured at every stage of its workmanship through effective NDE relying on conventional methods as explained. (author)

  11. Studies on representative disruption scenarios, associated electromagnetic and heat loads and operation window in ITER

    International Nuclear Information System (INIS)

    Fujieda, Hirobumi; Shimada, Michiya; Kawano, Yasunori; Ohmori, Junji; Neyatani, Yuzuru; Sugihara, Masayoshi; Gribov, Yuri; Ioki, Kimihiro; Khayrutdinov, Rustan; Lukash, Victor

    2007-07-01

    The impacts of plasma disruptions on ITER have been investigated in detail to confirm the robustness of the design of the machine to the potential consequential loads. The loads include both electromagnetic (EM) and heat loads on the in-vessel components and the vacuum vessel (VV). Several representative disruption scenarios are specified based on newly derived physics guidelines for the shortest current quench time as well as the maximum product of halo current fraction and toroidal peaking factor arising from disruptions in ITER. Disruption simulations with the DINA code and EM load analyses with a 3D finite element method (FEM) code are performed for these scenarios. Some margins are confirmed in the EM load on in-vessel components due to induced eddy and halo currents for these representative scenarios. However, the margins are not very large. The heat load on various parts of the first wall due to the vertical movement and the thermal quench (TQ) is calculated with a 2D heat conduction code based on the database of heat deposition during disruptions and simulation results with the DINA code. It is found that the beryllium (Be) wall will not melt during the vertical movement. Significant melting is anticipated for the upper Be wall and tungsten divertor baffle due to the TQ after the vertical movement. However, its impact could be substantially mitigated by implementing a reliable detection system of the vertical movement and a mitigation system, e.g., massive noble gas injection (MGI). Some melting of the upper Be wall is anticipated at major disruptions (MD). At least several tens of unmitigated disruptions must be considered even if an advanced prediction/mitigation system is implemented. With these unmitigated disruptions, the loss of Be layer is expected to be within approx. = 30-100 μm/event out of 10 mm thick Be first wall. Various post processing programs of the results simulated with the DINA code, which are developed for the design work, are

  12. A generic approach for steel containment vessel success criteria for severe accident loads

    International Nuclear Information System (INIS)

    Sammataro, R.F.; Solonick, W.R.; Edwards, N.W.

    1993-01-01

    Safety has been defined as the foremost design criterion for the Heavy Water New Production Reactor (NPR-HWR) by the U.S. DOE, Office of New Production Reactors (NP). The DOE-NP issued the Deterministic Severe Accident Criteria (DSAC) concept to guide the design of the NPR-HWR containment for resistance to severe accidents. The DSAC concept provides for a generic approach for containment vessel success criteria to predict the threshold of containment failure under severe accident loads. This concept consists of two parts: (1) Problem Statements and (2) Success Criteria. The paper is limited to a discussion of a success criteria. These criteria define acceptable containment response measures and limits for each problem statement. The criteria are based on the 'best estimate' of failure with no conservatism. Rather, conservatism, if required, is to be provided in the problem statements prepared by the designer and/or the regulatory authorities. The success criteria are presented on a multi-tiered basis for static pressure and temperature loadings, dynamic loadings, and missiles that may impact the containment. Within the static pressure and temperature loadings and the dynamic loadings, the criteria are separated into elastic analysis success criteria and inelastic analysis success criteria. Each of these areas, in turn, defines limits on either the stress or strain measures as well as on measures for buckling and displacements. The rationale upon which these criteria are based is contained in referenced documents. Rigorous validation of the criteria by comparison with results from analytical or experimental programs and application of the criteria to a containment design remain as future tasks. (orig./HP)

  13. Load Management in District Heating Operation

    DEFF Research Database (Denmark)

    Li, Hongwei; Wang, Stephen Jia

    2015-01-01

    Smooth operation of district heating system will avoid installation of expensive peak heat boilers, improve plant partial load performance, increase the system redundancy for further network expansion and improve its resilience to ensure security of supply during severe heating seasons. The peak...... heating load can be reduced through building demand side management. The building thermal mass can be used to shift the heating supply under the circumstance without jeopardizing the consumer thermal comfort. In this paper, the multi-agent framework is applied to a simplified building dynamic model...

  14. Investigation of impulsively loaded pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Brown, N.; Cornwell, R.; Hanner, D.; Leichter, H.; Mohr, P.

    1963-10-15

    Explosion containment vessels for containing from 2,000 to 3,000 five ton nuclear explosions are considered. Analysis methods appear adequate and lowest weights using the most advanced materials available in the next five years are projected.None of these materials can be fabricated today and all require extensive development. Present material technology limits the choice of materials and defines the weight. The addition of safety factors and fixtures (nozzles, etc.) will add to this weight considerably, and may well radically alter the vessel response. Improvements in the strength weight ratios of metals and glasses over those considered in this report do not appear reasonable at this time. Winding schemes to utilize the high strength of steel wires and somehow maintain a reasonable thickness appear to offer the most promise. A `ductile` beryllium would of course offer vast improvement, but no indications that this is being developed have appeared and all presently known beryllium is much too brittle.

  15. Stress concentration factors for integral and pad reinforced nozzles in spherical pressure vessels subjected to radial load and moment

    International Nuclear Information System (INIS)

    Soliman, S.F.; Gill, S.S.

    1979-01-01

    Charts are presented giving the elastic stress concentration factors in spherical pressure vessels with pad and integral reinforcement for radial branches subjected to radial load and moment. The effect of all the geometrical parameters is discussed, including the limitations of thin shell theory on the validity of the results. (author)

  16. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    International Nuclear Information System (INIS)

    Lee, K. H.; Woo, H. K.; Im, K. H.; Cho, S. Y.; Kim, J. B.

    2000-01-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10 -6 ∼10 -7 Pa, to produce clean plasma with low impurity containments. For this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 .deg. C, 350 .deg. C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses

  17. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Woo, H. K. [Chungnam National Univ., Taejon (Korea, Republic of); Im, K. H.; Cho, S. Y. [korea Basic Science Institute, Taejon (Korea, Republic of); Kim, J. B. [Hyundai Heavy Industries Co., Ltd., Ulsan (Korea, Republic of)

    2000-07-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10{sup -6}{approx}10{sup -7}Pa, to produce clean plasma with low impurity containments. For this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 .deg. C, 350 .deg. C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses.

  18. Structural Analysis of the NCSX Vacuum Vessel

    International Nuclear Information System (INIS)

    Fred Dahlgren; Art Brooks; Paul Goranson; Mike Cole; Peter Titus

    2004-01-01

    The NCSX (National Compact Stellarator Experiment) vacuum vessel has a rather unique shape being very closely coupled topologically to the three-fold stellarator symmetry of the plasma it contains. This shape does not permit the use of the common forms of pressure vessel analysis and necessitates the reliance on finite element analysis. The current paper describes the NCSX vacuum vessel stress analysis including external pressure, thermal, and electro-magnetic loading from internal plasma disruptions and bakeout temperatures of up to 400 degrees centigrade. Buckling and dynamic loading conditions are also considered

  19. Assessment of LWR piping design loading based on plant operating experience

    International Nuclear Information System (INIS)

    Svensson, P.O.

    1980-08-01

    The objective of this study has been to: (1) identify current Light Water Reactor (LWR) piping design load parameters, (2) identify significant actual LWR piping loads from plant operating experience, (3) perform a comparison of these two sets of data and determine the significance of any differences, and (4) make an evaluation of the load representation in current LWR piping design practice, in view of plant operating experience with respect to piping behavior and response to loading

  20. Revisiting the reactor pressure vessel for long-time operation

    International Nuclear Information System (INIS)

    Lapena, J.; Serrano, M.; Diego, G. de; Hernandez Mayoral, M.

    2013-01-01

    The reactor pressure vessel (RPV) is one of the key components of nuclear power plants, especially for long time operation. It is a non-replaceable component, at least with current technology. the structural integrity of the vessel is evaluated within called monitoring programs where the degradation of the mechanical properties due to neutron irradiation is determined. From the first designs of the RPVs and monitoring programs in the years 60-70 currently still in force, there have been major advances in the understanding of radiation damage and methods of evaluation. Thus, it is recommended the use of forgings instead of plates in the construction of the RPVs in order to reduce the number of welds, more sensitive to neutron irradiation, and using starting materials with less content of impurities, particularly copper. To evaluate the embrittlement of RPVs the Master Curve methodology is currently used, through the testing of the charpy specimens from the surveillance capsules, to determine the fracture toughness. This article summarizes the last activities of CIEMAT into the European research projects LONGIIFE and PERFORM60, about the knowledge of radiation damage in materials with low copper content, traditionally considered less sensitive to irradiation, and the use of the Master Curve in advanced surveillance programs. The activities related to the problems associated with the use of large forging, such as the appearance of hydrogen flakes in the vessel of Doel 3, and its implications, are also presented. (Author)

  1. Dynamic analysis of scraper conveyor operation with external loads

    Directory of Open Access Journals (Sweden)

    Świder Jerzy

    2017-01-01

    Full Text Available A load to an armoured face conveyor (AFC during coal mining is changeable and very difficult or even impossible to be predicted. Changes of the load to the upper scraper chain affect the load of the driving motor and generate changes in a scraper chain tension. Impact of increasing the external load to the upper scraper chain on the operation of electric motors and on the scraper chain tension is presented. The developed numerical model of the Rybnik 850 conveyor enabled identifying the places of the scraper chain high tension or places of its loosening. An impact of changing frequency of driving motor voltage on AFC’s operational conditions was tested and analysed using the AFC’s numerical model. During tests, tension of the scraper chain on the discharge end and the return end was recorded. High tension of the scraper chain and its loosening during the changeable load were also recorded on upward and downward transportation of run-of-mine material.

  2. The impact of microwave stray radiation to in-vessel diagnostic components

    Energy Technology Data Exchange (ETDEWEB)

    Hirsch, M.; Laqua, H. P.; Hathiramani, D.; Baldzuhn, J.; Biedermann, C.; Cardella, A.; Erckmann, V.; König, R.; Köppen, M.; Zhang, D. [Max-Planck-Institut für Plasmaphysik, Teilinstitut Greifswald, EURATOM Association, D-17489 Greifswald (Germany); Oosterbeek, J.; Brand, H. von der; Parquay, S. [Technische Universiteit Eindhoven, department Technische Natuurkunde, working group for Plasma Physics and Radiation Technology, Den Doelch 2, 5612 AZ Eindhoven (Netherlands); Jimenez, R. [Centro de Investigationes Energeticas, Medioambientales y Technológicas, Association EURATOM/CIEMAT, Avenida Complutense 22, Madrid 28040 (Spain); Collaboration: W7-X Teasm

    2014-08-21

    Microwave stray radiation resulting from unabsorbed multiple reflected ECRH / ECCD beams may cause severe heating of microwave absorbing in-vessel components such as gaskets, bellows, windows, ceramics and cable insulations. In view of long-pulse operation of WENDELSTEIN-7X the MIcrowave STray RAdiation Launch facility, MISTRAL, allows to test in-vessel components in the environment of isotropic 140 GHz microwave radiation at power load of up to 50 kW/m{sup 2} over 30 min. The results show that both, sufficient microwave shielding measures and cooling of all components are mandatory. If shielding/cooling measures of in-vessel diagnostic components are not efficient enough, the level of stray radiation may be (locally) reduced by dedicated absorbing ceramic coatings on cooled structures.

  3. Disruptions, loads, and dynamic response of ITER

    International Nuclear Information System (INIS)

    Nelson, B.; Riemer, B.; Sayer, R.; Strickler, D.; Barabaschi, P.; Ioki, K.; Johnson, G.; Shimizu, K.; Williamson, D.

    1995-01-01

    Plasma disruptions and the resulting electromagnetic loads are critical to the design of the vacuum vessel and in-vessel components of the International Thermonuclear Experimental Reactor (ITER). This paper describes the status of plasma disruption simulations and related analysis, including the dynamic response of the vacuum vessel and in-vessel components, stresses and deflections in the vacuum vessel, and reaction loads in the support structures

  4. Preliminary assessment of the effects of biaxial loading on reactor pressure vessel structural-integrity-assessment technology

    International Nuclear Information System (INIS)

    Pennell, W.E.; Bass, B.R.; Bryson, J.W.; Dickson, T.L.; McAfee, W.J.; Merkle, J.G.

    1996-01-01

    Effects of biaxial loading on shallow-flaw fracture toughness were studied to determine potential impact on structural integrity assessment of a reactor pressure vessel (RPV) under pressurized thermal shock (PTS) transient loading and pressure-temperature (PT) loading produced by reactor heatup and cooldown transients. Biaxial shallow-flaw fracture-toughness tests results were also used to determine the parameter controlling fracture in the transition temperature range, and to develop a related dual-parameter fracture-toughness correlation. Shallow-flaw and biaxial loading effects were found to reduce the conditional probability of crack initiation by a factor of nine when the shallow-flaw fracture-toughness K Jc data set, with biaxial-loading effects adjustments, was substituted in place of ASME Code K Ic data set in PTS analyses. Biaxial loading was found to reduce the shallow-flaw fracture toughness of RPV steel such that the lower-bound curve was located between ASME K Ic and K IR curves. This is relevant to future development of P-T curve analysis procedures. Fracture in shallow-flaw biaxial samples tested in the lower transition temperature range was shown to be strain controlled. A strain-based dual-parameter fracture-toughness correlation was developed and shown to be capable of predicting the effect of crack-tip constraint on fracture toughness for strain-controlled fracture

  5. Reactor vessel head permanent shield

    International Nuclear Information System (INIS)

    Hankinson, M.F.; Leduc, R.J.; Richard, J.W.; Malandra, L.J.

    1989-01-01

    A nuclear reactor is described comprising: a nuclear reactor pressure vessel closure head; control rod drive mechanisms (CRDMs) disposed within the closure head so as to project vertically above the closure head; cooling air baffle means surrounding the control rod drive mechanisms for defining cooling air paths relative to the control rod drive mechanisms; means defined within the periphery of the closure head for accommodating fastening means for securing the closure head to its associated pressure vessel; lifting lugs fixedly secured to the closure head for facilitating lifting and lowering movements of the closure head relative to the pressure vessel; lift rods respectively operatively associated with the plurality of lifting lugs for transmitting load forces, developed during the lifting and lowering movements of the closure head, to the lifting lugs; upstanding radiation shield means interposed between the cooling air baffle means and the periphery of the enclosure head of shielding maintenance personnel operatively working upon the closure head fastening means from the effects of radiation which may emanate from the control rod drive mechanisms and the cooling air baffle means; and connecting systems respectively associated with each one of the lifting lugs and each one of the lifting rods for connecting each one of the lifting rods to a respective one of each one of the lifting lugs, and for simultaneously connecting a lower end portion of the upstanding radiation shield means to each one of the respective lifting lugs

  6. Dynamic load effects on gate valve operability

    International Nuclear Information System (INIS)

    Steele, R. Jr.; MacDonald, P.E.; Arendts, J.G.

    1986-01-01

    The Idaho National Engineering Laboratory (INEL) participated in an internationally sponsored seismic research program conducted at the decommissioned Heissdampfreaktor (HDR) located in the Federal Republic of Germany. An existing piping system was modified by installation of an 8-in., naturally aged, motor-operated gate valve from a US nuclear power plant and a piping support system of US design. Six other piping support systems of varying flexibility from stiff to flexible were also installed at various times during the tests. Additional valve loadings included internal hydraulic loads and, during one block of tests, elevated temperature. The operability and integrity of the aged gate valve and the dynamic response of the various piping support system were measured during 25 representative seismic events

  7. The procurement and testing of the stainless steel in-vessel panels of the Wendelstein 7-X Stellarator

    Energy Technology Data Exchange (ETDEWEB)

    Peacock, A., E-mail: alan.peacock@ipp.mpg.de [European Commission c/o Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Girlinger, A. [MAN Diesel and Turbo SE D-94469 Deggendorf (Germany); Vorkoeper, A. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 17491 Greifswald (Germany); Boscary, J.; Greuner, H. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Hurd, F. [European Commission c/o Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Mendelevitch, B.; Pirsch, H.; Stadler, R.; Zangl, G. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany)

    2011-10-15

    320 In-vessel water cooled stainless steel panels, poloidal closure plates and pumping gap panels, covering an area of approximately 100 m{sup 2}, are used in Wendelstein7-X to protect the plasma vessel. The panels are manufactured at Deggendorf, Germany by MAN Diesel and Turbo SE. The panels consist of a laser welded sandwich of stainless steel plates together with a labyrinth of cooling channels and have a complicated geometry to fit the plasma vessel of Wendelstein 7-X. The hydraulic and mechanical stability requirements whilst maintaining the tight tolerances for the shape of the components are very demanding. The panels are designed to operate at up to an average heat load of 100 kW/m{sup 2} and a maximum heat load of 200 kW/m{sup 2} with a water velocity of approximately 2 m s{sup -1}. High heat flux testing of an un-cooled panel at a time averaged load of 200 kW/m{sup 2} for 10 s were successfully performed to support the start up phase of Wendelstein 7-X operation. Extensive testing both during manufacture and after delivery to IPP-Garching demonstrates the suitability of the delivered panels for their purpose.

  8. 49 CFR 176.78 - Use of power-operated industrial trucks on board vessels.

    Science.gov (United States)

    2010-10-01

    ... operation may expose the operator to danger from a falling object, the truck must be equipped with a driver... used to handle small objects or unstable loads must be equipped with a load backrest extension having... must be permanently connected to an exhaust duct leading to the open deck and terminate in a gooseneck...

  9. A locking mechanism for securing a loading buoy to a vessel. Lsemekanisme for fastgjring av en lastebye til et farty

    Energy Technology Data Exchange (ETDEWEB)

    Breivik, K.; Smedal, A.

    1994-07-04

    The invention relates to a locking mechanism for securing a loading/unloading buoy on a vessel. The buoy is of the type to be introduced into a submerged downwardly open receiving space in the vessel, and to be fastened in a releasable manner in the receiving space. The mechanism comprises hydraulically actuated locking elements, mounted about horizontal axes at the sides of the receiving space, to pivot between the locking and releasing positions, the buoy having a peripheral collar having a downwards facing abutment edge for engagement with the locking elements in the locking position thereof. 6 figs.

  10. 77 FR 11995 - Passenger Vessel Operator Financial Responsibility Requirements for Non-Performance of...

    Science.gov (United States)

    2012-02-28

    ... Vessel Operator Financial Responsibility Requirements for Non-Performance of Transportation AGENCY..., 2011, the Commission issued its Notice of Proposed Rulemaking (NPRM) to update its financial... cost of financial responsibility coverage because of the use of alternative coverage options. However...

  11. Applying the TOC Project Management to Operation and Maintenance Scheduling of a Research Vessel

    Science.gov (United States)

    Manti, M. Firdausi; Fujimoto, Hideo; Chen, Lian-Yi

    Marine research vessels and their systems are major assets in the marine resources development. Since the running costs for the ship are very high, it is necessary to reduce the total cost by an efficient scheduling for operation and maintenance. To reduce project period and make it efficient, we applied TOC project management method that is a project management approach developed by Dr. Eli Goldratt. It challenges traditional approaches to project management. It will become the most important improvement in the project management since the development of PERT and critical path methodologies. As a case study, we presented the marine geology research project for the purpose of operations in addition to repair on the repairing dock projects for maintenance of vessels.

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  13. Performance of a 2-megawatt high voltage test load

    International Nuclear Information System (INIS)

    Horan, D.; Kustom, R.; Ferguson, M.

    1995-01-01

    A high-power, water-cooled resistive load which simulates the electrical load characteristics of a high-power klystron, capable of 2 megawatts dissipation at 95 kV DC, was built and installed at the Advanced Photon Source for use in load-testing high voltage power supplies. During this testing, the test load has logged approximately 35 hours of operation at power levels in excess of one mezawatt. Slight variations in the resistance of the load during operation indicate that leakage currents in the cooling water may be a significant factor affecting the performance of the load. Sufficient performance data have been collected to indicate that leakage current through the deionized (DI) water coolant shunts roughly 15 percent of the full-load current around the load resistor elements. The leakage current could cause deterioration of internal components of the load. The load pressure vessel was disassembled and inspected internally for any signs of significant wear and distress. Results of this inspection and possible modifications for improved performance will be discussed

  14. Reduction of Dynamic Loads in Mine Lifting Installations

    Science.gov (United States)

    Kuznetsov, N. K.; Eliseev, S. V.; Perelygina, A. Yu

    2018-01-01

    Article is devoted to a problem of decrease in the dynamic loadings arising in transitional operating modes of the mine lifting installations leading to heavy oscillating motions of lifting vessels and decrease in efficiency and reliability of work. The known methods and means of decrease in dynamic loadings and oscillating motions of the similar equipment are analysed. It is shown that an approach based on the concept of the inverse problems of dynamics can be effective method of the solution of this problem. The article describes the design model of a one-ended lifting installation in the form of a two-mass oscillation system, in which the inertial elements are the mass of the lifting vessel and the reduced mass of the engine, reducer, drum and pulley. The simplified mathematical model of this system and results of an efficiency research of an active way of reduction of dynamic loadings of lifting installation on the basis of the concept of the inverse problems of dynamics are given.

  15. Modelling Vessel Traffic Service to understand resilience in everyday operations

    International Nuclear Information System (INIS)

    Praetorius, Gesa; Hollnagel, Erik; Dahlman, Joakim

    2015-01-01

    Vessel Traffic Service (VTS) is a service to promote traffic fluency and safety in the entrance to ports. This article's purpose has been to explore everyday operations of the VTS system to gain insights in how it contributes to safe and efficient traffic movements. Interviews, focus groups and an observation have been conducted to collect data about everyday operations, as well as to grasp how the VTS system adapts to changing operational conditions. The results show that work within the VTS domain is highly complex and that the two systems modelled realise their services vastly differently, which in turn affects the systems' ability to monitor, respond and anticipate. This is of great importance to consider whenever changes are planned and implemented within the VTS domain. Only if everyday operations are properly analysed and understood, it can be estimated how alterations to technology and organisation will affect the overall system performance

  16. Option of operating speed for vessels under low-carbon economy

    Directory of Open Access Journals (Sweden)

    Gang Li

    2013-03-01

    Full Text Available Purpose: To find out ships' optimum operating speed under low-carbon economy. Approach: First, it analyzes the relations between ship’s carbon emission and the operating speed, gets the optimum speed under which the entire fleet emit minimum carbon, then establishes the relations between the ship owner’s profit and the speed, extracts the speed under which the ship owner can gain the maximum profit and founds out it’s different from the speed under which the entire fleet emit minimum carbon. Findings: The government must take effective measures to make the ship owner slowdown and reduce emission. Originality: It first works out a balance point between the decrease of carbon emission brought by a lower operating speed and the increase of that caused by more vessels putting into service in a mathematical method.

  17. ITER cryostat main chamber and vacuum vessel pressure suppression system design

    International Nuclear Information System (INIS)

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

    1999-03-01

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

  18. ITER cryostat main chamber and vacuum vessel pressure suppression system design

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  19. Control of power distribution and the safety of a WWER-1000 reactor during operation in a load-following regime

    International Nuclear Information System (INIS)

    Filipchuk, E.V.; Dunaev, V.G.; Luk'yanets, I.A.; Potapenko, P.T.; Timokhin, E.S.; Voznesenskij, V.A.; Mitin, V.I.

    1984-01-01

    An important task in ensuring the operational safety of a nuclear power plant with a vessel-type water-cooled, water-moderated reactor is the development of algorithms for controlling power distribution in base-load and load-following regimes. The control strategy must ensure that the form of power distribution is maintained within the permissible limits in the case of situations presenting a nuclear hazard and of xenon fluctuations when there are variations in power output. The task of controlling power distribution can be broken down into an axial problem and a radial-azimuthal problem. For maintenance of the form of axial power distribution an algorithm of optimum fast-action control is proposed; for maintenance of the radial-azimuthal form, an algorithm of optimum (i.e. best) approximation to the given distribution of control. The two control problems are solved by using a single computerized procedure based on mathematical programming. The algorithms are designed for use in computerized control. The paper presents the results of algorithm simulation. The authors propose that in the first stage, activation of the control systems should involve the operator, who can view the sequence of actions on a display screen. (author)

  20. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K.H. [Chungnam National University Graduate School, Taejeon (Korea); Im, K.H.; Cho, S.Y. [Korea Basic Science Institute, Taejeon (Korea); Kim, J.B. [Hyundai Heavy Industries Co., Ltd. (Korea); Woo, H.K. [Chungnam National University, Taejeon (Korea)

    2000-11-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10{sup -6} {approx} 10{sup -7} Pa, to produce clean plasma with low impurity containments. for this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 deg.C, 350 deg.C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses. (author). 9 refs., 11 figs., 1 tab.

  1. Effects of dynamic loading of motor-operated valve actuators

    International Nuclear Information System (INIS)

    Damerell, P.S.; Daubresse, S.; Wolfe, K.J.; Dogan, T.; Gleeson, J.

    1994-01-01

    Experience has shown that valves with rising, nonrotating stems that are operated using electro-motor driven actuators can be susceptible to changes in output thrust at a constant torque switch setting as a result of changes in stem load time history. This effect is a concern because tests on these types of valves to verify thrust achieved at torque switch trip are often performed in situ under load conditions different from the required performance conditions. As part of a motor-operated valve research program being carried out by the Electric Power Research Institute, tests of typical electric motor actuators used with nuclear services valves have been performed. The test results show that changes in output thrust with load time history occur o varying degrees on different stem and stem nut combinations. When the effect exists, there is generally an increase in thrust at torque switch trip when load is developed rapidly from low initial loads, compared to when load is developed slowly. The effect is mainly a result of changes in the coefficient of friction at the stem-stem nut interface. The coefficient of friction is temporarily reduced under rapid loading conditions from low initial load, leading to increased thrust. The root cause is hypothesized to be a open-quotes squeeze-filmclose quotes effect, whereby mixed-mode lubrication (hydrodynamic plus boundary) temporarily replaces boundary lubrication. This paper describes the results of tests performed to better understand the phenomenon

  2. Homemade battery-operated multi-barreled muzzle-loading gun.

    Science.gov (United States)

    Ramiah, R; Thirunavukkarasu, G

    2003-11-01

    In a recent shootout by a terrorist group against a law enforcement agency, some unusual firearms were seized. On examination, these firearms were found to be homemade, battery-operated, multi-barreled muzzle-loading guns, analogous to a repeater. Reference to battery-operated firearms is rather scanty in the literature. Hence, the unique design features, electrical circuit, and the operation system of these unusual guns are described.

  3. Behavior of an operating floor subjected to heavy load drops

    International Nuclear Information System (INIS)

    Levin, H.A.; Chugh, S.C.; Goodyear, W.D.; Hall, W.J.; Tracy, T.J.

    1984-01-01

    A structural evaluation of the behavior of a Boiling Water Reactor operating floor subjected to postulated heavy load impact is presented relative to the evaluation of nuclear industry requirements for addressing the safe handling of heavy loads. A methodology for evaluation of important modes of structural response is presented along with a specification of performance limits that are consistent with desired functional requirements of the operating floor system. Options for improving undesirable behavior are discussed based upon an understanding of the physical phenomena involved, ultimately factoring this into a modification program. Reference is made to recent test experience that is applicable to the improvement of structural performance in modes of high shear ductility. Other relevant information is referenced that is applicable to the overall structural evaluation of nuclear power plant heavy load handling operations

  4. Containment vessel design and practice

    International Nuclear Information System (INIS)

    Bangash, Y.

    1983-01-01

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

  5. Estimation of inelastic behavior for a tapered nozzle in vessel due to thermal transient load using stress redistribution locus method

    International Nuclear Information System (INIS)

    Kobayashi, Ken-ichi; Yamada, Jun-ichi

    2010-01-01

    Simplified inelastic design procedures for elevated temperature components have been required to reduce simulation cost and to shorten a period of time for developing new projects. Stress redistribution locus (SRL) method has been proposed to provide a reasonable estimate employing both the elastic FEM analysis and a unique hyperbolic curve: ε tilde={1/σ tilde + (κ - 1)σ tilde}/κ, where ε tilde and σ tilde show dimensionless strain and stress normalized by corresponding elastic ones, respectively. This method is based on a fact that stress distribution in well deformed or high temperature components would change with deformation or time, and that the relation between the dimensionless stress and strain traces a kind of the elastic follow-up locus in spite of the constitutive equation of material and loading modes. In this paper, FEM analyses incorporating plasticity and creep in were performed for a tapered nozzle in reactor vessel under some thermal transient loads through the nozzle thickness. The normalized stress and strain was compared with the proposed SRL curve. Calculation results revealed that a critical point in the tapered nozzle due to the thermal transient load depended on a descending rate of temperature from the higher temperature in the operation cycle. Since the inelastic behavior in the nozzle resulted in a restricted area, the relationship between the normalized stress and strain was depicted inside the proposed SRL curve: Coefficient κ of the SRL in analyses is greater than the proposed one, and the present criterion guarantees robust structures for complicated components involving inelastic deformation. (author)

  6. The effects of load-sensitive behavior on the operability margins of motor-operated gate valves

    International Nuclear Information System (INIS)

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

    1993-01-01

    Testing of motor-operated gate valves at various loads has produced a phenomenon we call load-sensitive behavior. This phenomenon has a significant effect on the accuracy of the methods used (and proposed) in the nuclear industry for determining that these valves can perform their design basis function. A valve subjected to tests with low flow and pressure loadings may achieve a stem thrust (at seating) analytically determined to be adequate for design basis flows and pressures, but this is no guarantee that the valve will achieve the same stem thrust when actually subjected to those design basis loads. This is because the friction at the interface between the stem and the stem nut is higher in tests with higher flow and pressure loadings, and this loss to friction is outside the control of the motor-operator's torque switch. This paper identifies a tentative method for determining, a stable, useful value for the stem/stem-nut coefficient of friction, one that can possibly be extrapolated and used in calculations to accurately estimate the design basis thrust requirements of these valves

  7. Composite Overwrap Pressure Vessels: Mechanics and Stress Rupture Lifting Philosophy

    Science.gov (United States)

    Thesken, John C.; Murthy, Pappu L. N.; Phoenix, S. L.

    2009-01-01

    The NASA Engineering and Safety Center (NESC) has been conducting an independent technical assessment to address safety concerns related to the known stress rupture failure mode of filament wound pressure vessels in use on Shuttle and the International Space Station. The Shuttle s Kevlar-49 (DuPont) fiber overwrapped tanks are of particular concern due to their long usage and the poorly understood stress rupture process in Kevlar-49 filaments. Existing long term data show that the rupture process is a function of stress, temperature and time. However due to the presence of load sharing liners and the complex manufacturing procedures, the state of actual fiber stress in flight hardware and test articles is not clearly known. Indeed nonconservative life predictions have been made where stress rupture data and lifing procedures have ignored the contribution of the liner in favor of applied pressure as the controlling load parameter. With the aid of analytical and finite element results, this paper examines the fundamental mechanical response of composite overwrapped pressure vessels including the influence of elastic plastic liners and degraded/creeping overwrap properties. Graphical methods are presented describing the non-linear relationship of applied pressure to Kevlar-49 fiber stress/strain during manufacturing, operations and burst loadings. These are applied to experimental measurements made on a variety of vessel systems to demonstrate the correct calibration of fiber stress as a function of pressure. Applying this analysis to the actual qualification burst data for Shuttle flight hardware revealed that the nominal fiber stress at burst was in some cases 23 percent lower than what had previously been used to predict stress rupture life. These results motivate a detailed discussion of the appropriate stress rupture lifing philosophy for COPVs including the correct transference of stress rupture life data between dissimilar vessels and test articles.

  8. Composite Overwrap Pressure Vessels: Mechanics and Stress Rupture Lifing Philosophy

    Science.gov (United States)

    Thesken, John C.; Murthy, Pappu L. N.; Phoenix, Leigh

    2007-01-01

    The NASA Engineering and Safety Center (NESC) has been conducting an independent technical assessment to address safety concerns related to the known stress rupture failure mode of filament wound pressure vessels in use on Shuttle and the International Space Station. The Shuttle's Kevlar-49 fiber overwrapped tanks are of particular concern due to their long usage and the poorly understood stress rupture process in Kevlar-49 filaments. Existing long term data show that the rupture process is a function of stress, temperature and time. However due to the presence of load sharing liners and the complex manufacturing procedures, the state of actual fiber stress in flight hardware and test articles is not clearly known. Indeed non-conservative life predictions have been made where stress rupture data and lifing procedures have ignored the contribution of the liner in favor of applied pressure as the controlling load parameter. With the aid of analytical and finite element results, this paper examines the fundamental mechanical response of composite overwrapped pressure vessels including the influence of elastic-plastic liners and degraded/creeping overwrap properties. Graphical methods are presented describing the non-linear relationship of applied pressure to Kevlar-49 fiber stress/strain during manufacturing, operations and burst loadings. These are applied to experimental measurements made on a variety of vessel systems to demonstrate the correct calibration of fiber stress as a function of pressure. Applying this analysis to the actual qualification burst data for Shuttle flight hardware revealed that the nominal fiber stress at burst was in some cases 23% lower than what had previously been used to predict stress rupture life. These results motivate a detailed discussion of the appropriate stress rupture lifing philosophy for COPVs including the correct transference of stress rupture life data between dissimilar vessels and test articles.

  9. Adoption of in-vessel retention concept for VVER-440/V213 reactors in Central European Countries

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, Peter, E-mail: peter.matejovic@ivstt.sk [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Barnak, Miroslav; Bachraty, Milan; Vranka, Lubomir [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Berky, Robert [Integrita a Bezpecnost Ocelovych Konstrukcii, Rybnicna 40, 831 07 Bratislava (Slovakia)

    2017-04-01

    Highlights: • Design of in-vessel retention concept for VVER-440/V213 reactors. • Thermal loads acting on the inner reactor surface. • Structural response of reactor pressure vessel. • External reactor vessel cooling. - Abstract: An in-vessel retention (IVR) concept was proposed for standard VVER-440/V213 reactors equipped with confinement made of reinforced concrete and bubbler condenser pressure suppression system. This IVR concept is based on simple modifications of existing plant technology and thus it was attractive for plant operators in Central European Countries. Contrary to the solution that was adopted before at Loviisa NPP in Finland (two units of VVER-440/V213 reactor with steel confinement equipped with ice condenser), the coolant access to the reactor pressure vessel from flooded cavity is enabled via closable hole installed in the centre of thermal shield of the reactor lower head instead of lowering this massive structure in the case of severe accident. As a consequence, the crucial point of this IVR concept is narrow gap between torispherical lower head and thermal and biological shield. Here the highest thermal flux is expected in the case of severe accident. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width for coolant flow are of primarily importance. In this contribution the attention is paid especially to the analytical support with emphasis to the following points: 1) {sup ∗}Estimation of thermal loads acting on the inner reactor surface; 2) {sup ∗}Estimation of structural response of reactor pressure vessel (RPV) with emphasis on the deformation of outer reactor surface and its impact on the annular gap between RPV wall and thermal/biological shield; 3) {sup ∗}Analysis of external reactor vessel cooling. For this purpose the ASTEC code was used for performing analysis of core degradation scenarios, the ANSYS code for structural analysis of reactor vessel

  10. Features of systems for operational control of WWER vessel metal, used in the USSR

    International Nuclear Information System (INIS)

    Yurchenko, Yu.F.

    1987-01-01

    The report descrides key features of an improved system developed to serve for monitoring the soundness of the metal material of the operating high-pressure reactor vessels in nuclear power generation plants in the Soviet Union. The most important feature is that an external monitoring subsystem is incorporated in the system. The subsystem has the advantage of ensuring the following: high defect detectability due to the absense of austenite lining on the outer surface of the reactor vessel; implementation of monitoring work without removing in-pile structures in parallel with preventive maintenance work during annual partial fuel replacement; and application of other monitoring techniques, such as accoustic emission, in future. Another feature is that radiography by iridium-192 and cobalt-90 is employed to support the external monitoring of the metal material of the nozzle component. An optical periscope is incorporated to permit detailed visual inspection of the lining surface of the inner face of a reactor vessel. Data on the coordinates of defects are displayed on a TV screen and recorded and reproduced by a video recorder. The system also uses an 'echo method' for ultrasonic monitoring and a high sensitive 'tandem method' for detecting vertically oriented defects. The entire system can be operated by remote control. (Nogami, K.)

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

    International Nuclear Information System (INIS)

    Neidel, Andreas

    2016-01-01

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

  12. Operational Strategy of CBPs for load balancing of Operators in Advanced Main Control Room

    International Nuclear Information System (INIS)

    Kim, Seunghwan; Kim, Yochan; Jung, Wondea

    2014-01-01

    With the using of a computer-based control room in an APR1400 (Advanced Pressurized Reactor-1400), the operators' behaviors in the main control room had changed. However, though the working environment of operators has been changed a great deal, digitalized interfaces can also change the cognitive tasks or activities of operators. First, a shift supervisor (SS) can confirm/check the conduction of the procedures and the execution of actions of board operators (BOs) while confirming directly the operation variables without relying on the BOs. Second, all operators added to their work the use of a new CBP and Soft Controls, increasing their procedural workload. New operational control strategies of CBPs are necessary for load balancing of operator's task load in APR1400. In this paper, we compared the workloads of operators in an APR1400 who work with two different usages of the CBP. They are SS oriented usage and SS-BO collaborative usage. In this research, we evaluated the workloads of operators in an advanced main control room by the COCOA method. Two types of CBP usages were defined and the effects of these usages on the workloads were investigated. The obtained results showed that the workloads between operators in a control room can be balanced according to the CBP usages by assigning control authority to the operators

  13. Operational Strategy of CBPs for load balancing of Operators in Advanced Main Control Room

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seunghwan; Kim, Yochan; Jung, Wondea [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    With the using of a computer-based control room in an APR1400 (Advanced Pressurized Reactor-1400), the operators' behaviors in the main control room had changed. However, though the working environment of operators has been changed a great deal, digitalized interfaces can also change the cognitive tasks or activities of operators. First, a shift supervisor (SS) can confirm/check the conduction of the procedures and the execution of actions of board operators (BOs) while confirming directly the operation variables without relying on the BOs. Second, all operators added to their work the use of a new CBP and Soft Controls, increasing their procedural workload. New operational control strategies of CBPs are necessary for load balancing of operator's task load in APR1400. In this paper, we compared the workloads of operators in an APR1400 who work with two different usages of the CBP. They are SS oriented usage and SS-BO collaborative usage. In this research, we evaluated the workloads of operators in an advanced main control room by the COCOA method. Two types of CBP usages were defined and the effects of these usages on the workloads were investigated. The obtained results showed that the workloads between operators in a control room can be balanced according to the CBP usages by assigning control authority to the operators.

  14. Temperature field and thermal stress analysis of the HT-7U vacuum vessel

    International Nuclear Information System (INIS)

    Song Yuntao; Yao Damao; Wu Songtao; Weng Peide

    2000-01-01

    The HT-7U vacuum vessel is an all-metal-welded double-wall interconnected with toroidal and poloidal stiffening ribs. The channels formed between the ribs and walls are filled with boride water as a nuclear shielding. On the vessel surface facing the plasma are installed cable-based Ohmic heaters. Prior to plasma operation the vessel is to be baked out and discharge cleaned at about 250 degree C. During baking out the non-uniformity of temperature distribution on the vacuum vessel will bring about serious thermal stress that can damage the vessel. In order to determine and optimize the design of the HT-7U vacuum vessel, a three-dimensional finite element model was performed to analyse its temperature field and thermal stress. the maximal thermal stress appeared on the round of lower vertical port and maximal deformation located just on the region between the upper vertical port and the horizontal port. The results show that the reinforced structure has a good capability of withstanding the thermal loads

  15. PWR vessel flaw distribution development

    International Nuclear Information System (INIS)

    Rosinski, S.T.; Kennedy, E.L.; Foulds, J.R.; Kinsman, K.M.

    1990-01-01

    This paper reports on PWR pressure vessels which operate under NRC rules and regulatory guides intended to prevent failure of the vessels. Plants failing to meet the operating criteria specified under these rules and regulations are required to analytically demonstrate fitness for service in order to continue operation. The initial flaw size or distribution of initial vessel flaws is a key input to the required vessel integrity analyses. However, the flaw distribution assumed in the development of the NRC Regulations and recommended for the plant specific analyses is potentially over-conservative. This is because the distribution is based on the limited amount of vessel inspection data available at the time the criteria were being developed and does not take full advantage of the more recent and reliable domestic vessel inspection results. The U.S. Department of Energy is funding an effort through Sandia National Laboratories to investigate the possibility of developing a new flaw distribution based on the increased amount and improved reliability of domestic vessel inspection data. Results of Phase I of the program indicate that state-of-the-art NDE systems' capabilities are sufficient for development of a new flaw distribution that could ultimately provide life extension benefits over the presently required operating practice

  16. Replacement of a vessel head, an operation which today gets easily into its stride

    International Nuclear Information System (INIS)

    Mardon, P.; Chaumont, J.C.; Lambiotte, P.

    1995-01-01

    In 1992, one year after the detection of a leak in a vessel head of the Electricite de France (EDF) Bugey 4 reactor, the head was replaced by the Framatome-Jeumont Industrie Group. Today, this group, which has developed new methods and new tools to optimize the cost, the time-delay and the dosimetry of this kind of intervention, has performed 11 additional replacements, two of which on 1300 MWe power units. This paper describes step by step the successive operations required for a complete vessel head replacement, including the testing of safety systems before starting up the reactor. (J.S.). 7 photos

  17. Manufacture of EAST VS In-Vessel Coil

    International Nuclear Information System (INIS)

    Long, Feng; Wu, Yu; Du, Shijun; Jin, Huan; Yu, Min; Han, Qiyang; Wan, Jiansheng; Liu, Bin; Qiao, Jingchun; Liu, Xiaochuan; Li, Chang; Cai, Denggang; Tong, Yunhua

    2013-01-01

    Highlights: • ITER like Stainless Steel Mineral Insulation Conductor (SSMIC) used for EAST Tokamak VS In-Vessel Coil manufacture first time. • Research on SSMIC fabrication was introduced in detail. • Two sets totally four single-turn VS coils were manufactured and installed in place symmetrically above and below the mid-plane in the vacuum vessel of EAST. • The manufacture and inspection of the EAST VS coil especially the joint for the SSMIC connection was described in detail. • The insulation resistances of all the VS coils have no significant reduction after endurance test. -- Abstract: In the ongoing latest update round of EAST (Experimental Advanced Superconducting Tokamak), two sets of two single-turn Vertical Stabilization (VS) coils were manufactured and installed symmetrically above and below the mid-plane in the vacuum vessel of EAST. The Stainless Steel Mineral Insulated Conductor (SSMIC) developed for ITER In-Vessel Coils (IVCs) in Institute of Plasma Physics, Chinese Academy of Science (ASIPP) was used for the EAST VS coils manufacture. Each turn poloidal field VS coil includes three internal joints in the vacuum vessel. The middle joint connects two pieces of conductor which together form an R2.3 m arc segment inside the vacuum vessel. The other two joints connect the arc segment with the two feeders near the port along the toroidal direction to bear lower electromagnetic loads during operation. Main processes and tests include material performances checking, conductor fabrication, joint connection and testing, coil forming, insulation performances measurement were described herein

  18. Transient pressure measurements at part load operating condition of ...

    Indian Academy of Sciences (India)

    Rahul Goyal

    2 Water Power Laboratory, Department of Energy and Process Engineering, Norwegian ... The present paper focuses on the investigation of a high head model Francis turbine operating at 50% load. ..... cascade of bandpass filters in MATLAB to eliminate the .... simulation of unsteady operation of hydroelectric systems.

  19. Constrained model predictive control for load-following operation of APR reactors

    International Nuclear Information System (INIS)

    Kim, Jae Hwan; Lee, Sim Won; Kim, Ju Hyun; Na, Man Gyun; Yu, Keuk Jong; Kim, Han Gon

    2012-01-01

    The load-following operation of APR+ reactor is needed to control the power effectively using the control rods and to restrain the reactivity control from using the boric acid for flexibility of plant operation. Usually, the reason why the disproportion of axial flux distribution occurs during load-following operation is xenon-induced oscillation. The xenon has a very high absorption cross-section and makes the impact on the reactor delayed by the iodine precursor. The power maneuvering using automatically load-following operation has advantage in terms of safety and economic operation of the reactor, so the controller has to be designed efficiently. Therefore, an advanced control method that meets the conditions such as automatic control, flexibility, safety, and convenience is necessary to load-following operation of APR+ reactor. In this paper, the constrained model predictive control (MPC) method is applied to design APR reactor's automatic load-following controller for the integrated thermal power level and axial shape index (ASI) control. Some controllers use only the current tracking command, but MPC considers future commands in addition to the current tracking command. So, MPC can achieve better tracking performance than others. Furthermore, an MPC is to used in many industrial process control systems. The basic concept of the MPC is to solve an optimization problem for a finite future time interval at present time and to implement the first optimal control input as the current control input. The KISPAC-1D code, which models the APR+ nuclear power plants, is interfaced to the proposed controller to verify the tracking performance of the reactor power level and ASI. It is known that the proposed controller exhibits very fast tracking responses

  20. A bankruptcy problem approach to load-shedding in multiagent-based microgrid operation.

    Science.gov (United States)

    Kim, Hak-Man; Kinoshita, Tetsuo; Lim, Yujin; Kim, Tai-Hoon

    2010-01-01

    A microgrid is composed of distributed power generation systems (DGs), distributed energy storage devices (DSs), and loads. To maintain a specific frequency in the islanded mode as an important requirement, the control of DGs' output and charge action of DSs are used in supply surplus conditions and load-shedding and discharge action of DSs are used in supply shortage conditions. Recently, multiagent systems for autonomous microgrid operation have been studied. Especially, load-shedding, which is intentional reduction of electricity use, is a critical problem in islanded microgrid operation based on the multiagent system. Therefore, effective schemes for load-shedding are required. Meanwhile, the bankruptcy problem deals with dividing short resources among multiple agents. In order to solve the bankruptcy problem, division rules, such as the constrained equal awards rule (CEA), the constrained equal losses rule (CEL), and the random arrival rule (RA), have been used. In this paper, we approach load-shedding as a bankruptcy problem. We compare load-shedding results by above-mentioned rules in islanded microgrid operation based on wireless sensor network (WSN) as the communication link for an agent's interactions.

  1. PHM Enabled Autonomous Propellant Loading Operations

    Science.gov (United States)

    Walker, Mark; Figueroa, Fernando

    2017-01-01

    The utility of Prognostics and Health Management (PHM) software capability applied to Autonomous Operations (AO) remains an active research area within aerospace applications. The ability to gain insight into which assets and subsystems are functioning properly, along with the derivation of confident predictions concerning future ability, reliability, and availability, are important enablers for making sound mission planning decisions. When coupled with software that fully supports mission planning and execution, an integrated solution can be developed that leverages state assessment and estimation for the purposes of delivering autonomous operations. The authors have been applying this integrated, model-based approach to the autonomous loading of cryogenic spacecraft propellants at Kennedy Space Center.

  2. Modeling and measurement of the motion of the DIII-D vacuum vessel during vertical instabilities

    International Nuclear Information System (INIS)

    Reis, E.; Blevins, R.D.; Jensen, T.H.; Luxon, J.L.; Petersen, P.I.; Strait, E.J.

    1991-11-01

    The motions of the D3-D vacuum vessel during vertical instabilities of elongated plasmas have been measured and studied over the past five years. The currents flowing in the vessel wall and the plasma scrapeoff layer were also measured and correlated to a physics model. These results provide a time history load distribution on the vessel which were input to a dynamic analysis for correlation to the measured motions. The structural model of the vessel using the loads developed from the measured vessel currents showed that the calculated displacement history correlated well with the measured values. The dynamic analysis provides a good estimate of the stresses and the maximum allowable deflection of the vessel. In addition, the vessel motions produce acoustic emissions at 21 Hertz that are sufficiently loud to be felt as well as heard by the D3-D operators. Time history measurements of the sounds were correlated to the vessel displacements. An analytical model of an oscillating sphere provided a reasonable correlation to the amplitude of the measured sounds. The correlation of the theoretical and measured vessel currents, the dynamic measurements and analysis, and the acoustic measurements and analysis show that: (1) The physics model can predict vessel forces for selected values of plasma resistivity. The model also predicts poloidal and toroidal wall currents which agree with measured values; (2) The force-time history from the above model, used in conjunction with an axisymmetric structural model of the vessel, predicts vessel motions which agree well with measured values; (3) The above results, input to a simple acoustic model predicts the magnitude of sounds emitted from the vessel during disruptions which agree with acoustic measurements; (4) Correlation of measured vessel motions with structural analysis shows that a maximum vertical motion of the vessel up to 0.24 in will not overstress the vessel or its supports. 11 refs., 10 figs., 1 tab

  3. Operation and Equivalent Loads of Wind Turbines in Large Wind Farms

    Science.gov (United States)

    Andersen, Soren Juhl; Sorensen, Jens Norkaer; Mikkelsen, Robert Flemming

    2017-11-01

    Wind farms continue to grow in size and as the technology matures, the design of wind farms move towards including dynamic effects besides merely annual power production estimates. The unsteady operation of wind turbines in large wind farms has been modelled with EllipSys3D(Michelsen, 1992, and Sørensen, 1995) for a number of different scenarios using a fully coupled large eddy simulations(LES) and aero-elastic framework. The turbines are represented in the flow fields using the actuator line method(Sørensen and Shen, 2002), where the aerodynamic forces and deflections are derived from an aero-elastic code, Flex5(Øye, 1996). The simulations constitute a database of full turbine operation in terms of both production and loads for various wind speeds, turbulence intensities, and turbine spacings. The operating conditions are examined in terms of averaged power production and thrust force, as well as 10min equivalent flapwise bending, yaw, and tilt moment loads. The analyses focus on how the performance and loads change throughout a given farm as well as comparing how various input parameters affect the operation and loads of the wind turbines during different scenarios. COMWIND(Grant 2104-09- 067216/DSF), Nordic Consortium on Optimization and Control of Wind Farms, Eurotech Greentech Wind project, Winds2Loads, and CCA LES. Ressources Granted on SNIC and JESS. The Vestas NM80 turbine has been used.

  4. Design features of the KSTAR in-vessel control coils

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-06-15

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

  5. Engineering to Control Noise, Loading, and Optimal Operating Points

    International Nuclear Information System (INIS)

    Mitchell R. Swartz

    2000-01-01

    Successful engineering of low-energy nuclear systems requires control of noise, loading, and optimum operating point (OOP) manifolds. The latter result from the biphasic system response of low-energy nuclear reaction (LENR)/cold fusion systems, and their ash production rate, to input electrical power. Knowledge of the optimal operating point manifold can improve the reproducibility and efficacy of these systems in several ways. Improved control of noise, loading, and peak production rates is available through the study, and use, of OOP manifolds. Engineering of systems toward the OOP-manifold drive-point peak may, with inclusion of geometric factors, permit more accurate uniform determinations of the calibrated activity of these materials/systems

  6. Thermal–mechanical stress analysis of pressurized water reactor pressure vessel with/without a preexisting crack under grid load following conditions

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov; Soppet, William K.; Majumdar, Saurin; Natesan, Krishnamurti

    2016-12-15

    Highlights: • Use of intermittent renewable-energy source in power grid is becoming a trend. • Gird load-following can leads to variable power demand from Nuclear power plant. • Reactor components can be stressed differently under gird load-following mode. • Estimation of stress–strain state under grid load-following condition is essential. - Abstract: In this paper, we present thermal–mechanical stress analysis of a pressurized water reactor pressure vessel and its hot-leg and cold-leg nozzles. Results are presented from thermal and thermal–mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting crack in the reactor nozzle (axial crack in hot leg nozzle). From the model results it is found that the stress–strain states are significantly higher in case of presence of crack than without crack. The stress–strain state under grid load following condition are more realistic compared to the stress–strain state estimated assuming simplified transients.

  7. Thermal–mechanical stress analysis of pressurized water reactor pressure vessel with/without a preexisting crack under grid load following conditions

    International Nuclear Information System (INIS)

    Mohanty, Subhasish; Soppet, William K.; Majumdar, Saurin; Natesan, Krishnamurti

    2016-01-01

    Highlights: • Use of intermittent renewable-energy source in power grid is becoming a trend. • Gird load-following can leads to variable power demand from Nuclear power plant. • Reactor components can be stressed differently under gird load-following mode. • Estimation of stress–strain state under grid load-following condition is essential. - Abstract: In this paper, we present thermal–mechanical stress analysis of a pressurized water reactor pressure vessel and its hot-leg and cold-leg nozzles. Results are presented from thermal and thermal–mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting crack in the reactor nozzle (axial crack in hot leg nozzle). From the model results it is found that the stress–strain states are significantly higher in case of presence of crack than without crack. The stress–strain state under grid load following condition are more realistic compared to the stress–strain state estimated assuming simplified transients.

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  9. NET in-vessel vehicle system

    International Nuclear Information System (INIS)

    Jones, H.

    1991-02-01

    The CFFTP/Spar In-vessel Vehicle System concept for in-vessel remote maintenance of the NET/ITER machine is described. It comprises a curved deployable boom, a vehicle which can travel on the boom and an end effector or work unit mounted on the vehicle. The stowed boom, vehicle, and work unit are inserted via the equatorial access port of the torus. Following insertion the boom is deployed and locked in place. The vehicle may then travel along the boom to transport the work unit to any desired location. A novel feature of the concept is the deployable boom. When fully deployed, it closely resembles a conventional curved truss structure in configuration and characteristics. However, the joints of the truss structure are hinged so that it can fold into a compact package, of less than 20% of deployed volume for storage, transportation and insertion into the torus. A full-scale 2-metre long section of this boom was produced for demonstration purposes. As part of the concept definition the work unit for divertor handling was studied to demonstrate that large payloads could be manipulated within the confines of the torus using the in-vessel vehicle system. Principal advantages of the IVVS are its high load capacity and rigidity, low weight and stowed volume, simplicity of control and operation, and its relatively high speed of transportation

  10. 19 CFR 4.97 - Salvage vessels.

    Science.gov (United States)

    2010-04-01

    ... United States and Great Britain ‘concerning reciprocal rights for United States and Canada in the... meaning of this statute. (e) A Mexican vessel may engage in a salvage operation on a Mexican vessel in any territorial waters of the United States in which Mexican vessels are permitted to conduct such operations by...

  11. A dynamic simulation to study NET in-vessel handling operations

    International Nuclear Information System (INIS)

    Fung, P.T.F.

    1989-01-01

    The inspection, maintenance and repair of the Next European Torus (NET) fusion machine will require the extensive use of remote handling equipment to minimise the human exposure to the high radiation environment. The use of efficient manipulators will reduce the NET downtime by reducing the preparation time for entry into the controlled area and by performing the task with reasonable area and by performing the task with reasonable dexterity and speed, consistent with safety. A high fidelity simulation is a valuable tool to assist in the manipulator design, operations, trajectory planning, parameter optimisation and system verification. A manipulator simulation package called ASAD was originally developed by Spar for space manipulator applications. It is now being adapted to simulate the in-Vessel HandlingUnit for the NET program. This terestrial version of ASAD has been name ASAD - T. Spar, through the services of the Canadian Fusion Fuels Technology Project, is under contract to the NET program for the performance of this activity. This paper describes the capabillities and underlying assumptions of ASAD - T, aling with description of the simulation development of the NET in-vessel manipulator. (author). 4 refs.; 7 figs

  12. Nonlinear analysis of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

    Connor, J.J.

    1975-01-01

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

  13. Benchmarking of multigroup neutron cross sections libraries on neutron transmission through WWER-440 vessel

    International Nuclear Information System (INIS)

    Ilieva, K.; Belousov, S.; Apostolov, T.

    1998-01-01

    The verification of calculated neutron fluence onto the WWER-440/230 pressure vessel is very topical task in particular referring that some of this type of reactors have been operated the major part of its design lifetime. Since the induced activity from the neutron irradiation onto the elements is a simple response of neutron flux the neutron fluence verification usually is done using the measured activity of radionuclides produced during reactor operation. Calculational and experimental results of 54 Mn induced activity of scraps from inner wall of Unit 1 reactor pressure vessel after 18th cycle and detectors irradiated behind the vessel during the 18th cycle of Unit 1 at Kozloduy NPP as well as neutron flux attenuation through the WWER-440/230 pressure vessel are presented. Neutron cross sections libraries generated on the base of ENDF/B-IV and ENDF/B-VI have been used in the calculations. The comparative analysis of evaluated activities and attenuation coefficient demonstrates the better reliability of the neutron fluence calculations by the libraries based on ENDF/B-VI than by ones on ENDF/B-IV. The extreme rarity of data for the activity of scraps from the WWER-440 reactor vessel and its combination with the data for the detectors irradiated behind the vessel makes them especially attractive for verification of calculational methods of neutron fluence onto the WWER-440 vessel with dummy cassettes loading. (author)

  14. Mark III Containment vessel/annulus concrete design

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  15. Rod behaviour under base load, load follow and frequency control operation: CYRANO 2 code predictions versus experimental results

    International Nuclear Information System (INIS)

    Gautier, B.; Raybaud, A.

    1984-01-01

    The French PWR reactors are now currently operating under load follow and frequency control. In order to demonstrate that these operating conditions were not able to increase the fuel failure rate, fuel rod behaviour calculations have been performed by E.D.F. with CYRANO 2 code. In parallel with these theoretical calculations, code predictions have been compared to experimental results. The paper presents some of the comparisons performed on 17x17 fuel irradiated in FESSENHEIM 2 up to 30 GWd/tU under base load operation and in the CAP reactor under load follow and frequency control conditions. It is shown that experimental results can be predicted with a reasonable accuracy by CYRANO 2 code. The experimental work was carried out under joint R and D programs by EDF, FRAGEMA, CEA, and WESTINGHOUSE (CAP program by French partners only). (author)

  16. A Bankruptcy Problem Approach to Load-shedding in Multiagent-based Microgrid Operation

    Directory of Open Access Journals (Sweden)

    Yujin Lim

    2010-09-01

    Full Text Available A microgrid is composed of distributed power generation systems (DGs, distributed energy storage devices (DSs, and loads. To maintain a specific frequency in the islanded mode as an important requirement,  the control of DGs’ output and charge action of DSs are used in supply surplus conditions and load-shedding and discharge action of DSs are used in supply shortage conditions. Recently, multiagent systems for autonomous microgrid operation have been studied. Especially, load-shedding, which is intentional reduction of electricity use, is a critical problem in islanded microgrid operation based on the multiagent system. Therefore, effective schemes for load-shedding are required. Meanwhile, the bankruptcy problem deals with dividing short resources among multiple agents. In order to solve the bankruptcy problem, division rules, such as the constrained equal awards rule (CEA, the constrained equal losses rule (CEL, and the random arrival rule (RA, have been used. In this paper, we approach load-shedding as a bankruptcy problem. We compare load-shedding results by above-mentioned rules in islanded microgrid operation based on wireless sensor network (WSN as the communication link for an agent’s interactions.

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

    International Nuclear Information System (INIS)

    Watson, P.S.

    1990-01-01

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

  18. Development of a load cell for mechanical testing in hydrogen

    International Nuclear Information System (INIS)

    McCabe, L.P.

    1982-01-01

    Mechanical testing in hydrogen environments is performed on materials to determine hydrogen compatibility. Many tests are performed on small test samples in pressure vessels where monitoring of actual sample load is difficult. A method was developed to monitor small samples by placing inside the vessel a miniature load cell which is capable of measuring loads of less than 100 lbs. The load cell monitors load by means of a Wheatstone Bridge circuit composed of four strain gages. Two of the gages are mounted on a stainless steel stub which becomes part of the vessel load string; the others are wired outside the pressure vessel. Previously, load cells have been short-lived because of hydrogen diffusion into the epoxy-phenolic adhesive used to attach the strain gages to the stub. The use of a flame-sprayed ceramic, however, rather than an organic epoxy to mount the strain gages appears to produce a load cell resistant to the hydrogen test environment

  19. The role of pressure vessel embrittlement in the long term operation of nuclear power plants

    International Nuclear Information System (INIS)

    Ballesteros, A.; Ahlstrand, R.; Bruynooghe, C.; Estorff, U. von; Debarberis, L.

    2012-01-01

    Highlights: ► Relevant open scientific issues for the long term operation of RPVs are discussed (flux effect, late blooming phases, etc.). ► Several European and American research programmes dealing with these open issues are reviewed. ► A method for consolidation and preservation of knowledge in this field is presented. - Abstract: The lack of new build of plants over the last twenty years has resulted in a switch within the industry from design, construction and development of new systems to the strengthening of safety systems and to the life extension, or long term operation (LTO), of existing reactors. The most relevant component of any nuclear power plan (NPP) is the reactor pressure vessel (RPV). This is because currently the RPV is still considered irreplaceable or prohibitively expensive to replace. This means, that if it degrades sufficiently, it could be the operational life limiting feature of the NPP. A RPV operational life of 60 years is being considered frequently by many utilities in their plant life management programmes. Areas of improvement facing long term operation are the reduction of uncertainties in the embrittlement parameters of irradiated vessels, and the development of embrittlement trend curves at high fluence levels, where surveillance data are scarce. Different techniques can be used to upgrade the surveillance programmes, as the use of miniature or reconstituted specimens and the application of best estimate assessment tools (e.g. Master Curve). Several relevant international research projects are on-going or have been proposed to clarify the material condition of long operated vessels. Knowledge management is a complementary tool, but not for it less important. The general context for LTO of RPVs is presented in this paper. Starting with a review of relevant embrittlement issues still open, followed by presenting the different techniques and tools that can be used to support LTO, and summarising the scopes of relevant European

  20. Expected load spectra of prototype Francis turbines in low-load operation using numerical simulations and site measurements

    Science.gov (United States)

    Eichhorn, M.; Taruffi, A.; Bauer, C.

    2017-04-01

    The operators of hydropower plants are forced to extend the existing operating ranges of their hydraulic machines to remain competitive on the energy market due to the rising amount of wind and solar power. Faster response times and a higher flexibility towards part- and low-load conditions enable a better electric grid control and assure therefore an economic operation of the power plant. The occurring disadvantage is a higher dynamic excitation of affected machine components, especially Francis turbine runners, due to pressure pulsations induced by unsteady flow phenomena (e.g. draft tube vortex ropes). Therefore, fatigue analysis becomes more important even in the design phase of the hydraulic machines to evaluate the static and dynamic load in different operating conditions and to reduce maintenance costs. An approach including a one-way coupled fluid-structure interaction has been already developed using unsteady CFD simulations and transient FEM computations. This is now applied on two Francis turbines with different specific speeds and power ranges, to obtain the load spectra of both machines. The results are compared to strain gauge measurements on the according Francis turbines to validate the overall procedure.

  1. Laguna Verde annulus pressurization loads evaluation

    International Nuclear Information System (INIS)

    Castaneda, M. A.; Cruz, M. A.; Cardenas, J. B.; Vargas, A.; Cruz, H. J.; Mercado, J. J.

    2010-10-01

    Annulus pressurization, jet impingement, pipe whip restraint and jet thrust are phenomena related to postulated pipe ruptures. A postulated pipe rupture at the weld between recirculation, or feedwater piping and a reactor nozzle safe end, will lead to a high flow rate of flashing water/steam mixture into the annulus between the reactor pressure vessel and the biological shield wall. The total effect of the vessel and pipe inventory blowdown from the break being postulated must be accounted for in the evaluation. A recirculation line break will give rise to an angular dependent short term pressure differential around the vessel, followed by a longer term pressure buildup in the annulus. A recirculation line postulated rupture may not produce worst case conditions and reference to time intervals for only the recirculation break should be treated superficially. A postulated rupture of the feedwater piping may produce the extreme case for determining: 1) the shield wall and reactor vessel to pedestal interactions, 2) loading on the reactor vessel internals, or 3) responses for the balance of piping attached to the vessel. Recently it was identified a potential issue regarding the criteria used to determine which cases were evaluated for Annulus Pressurization (A P) loads for new loads plants. The original A P loads methodology in the late 1970 and early 1980 years separated the mass/energy release calculation from the structural response calculation based on the implicit assumption that the maximum overall mass/energy release will result in maximizing the structural response and corresponding stresses on the reactor pressure vessel, internals, and containment structures. This process did not consider the dynamic response in the primary and secondary safety related structures, components and equipment. Consequently, the A P loads used as input for design adequacy evaluations of Nuclear Steam Supply System safety related components for new loads plants might have

  2. Capacity of Prestressed Concrete Containment Vessels with Prestressing Loss

    International Nuclear Information System (INIS)

    SMITH, JEFFREY A.

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  4. Containment vessel drain system

    Science.gov (United States)

    Harris, Scott G.

    2018-01-30

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

  5. Application of a newly built semi-submersible vessel for transportation of a tension leg platform

    Science.gov (United States)

    Zhang, Dagang; Sun, Weiying; Fan, Zhixia

    2012-09-01

    Transportation of tension leg platform (TLP) structures for a long distance has always been associated with the use of a heavy semi-transport vessel. The requirements of this type of vessel are always special, and their availability is limited. To prepare for the future development of South China Sea deepwater projects, the China Offshore Oil Engineering Corporation has recently built a heavy lift transport vessel-Hai Yang Shi You 278. This semi-submersible vessel has a displacement capacity of 50k DWT, and a breath of 42 meters. Understanding the vessel's applicability and preparing it for use in future deepwater projects are becoming imminent needs. This paper reviews the current critical issues associated with TLP transportation and performs detailed analysis of the designed TLP during load-out and transportation. The newly built COOEC transportation vessel HYSY 278 was applied to dry transport of the TLP structure from the COOEC fabrication yard in Qingdao to an oil field in South China Sea. The entire process included the load-out of the TLP structure from the landsite of the fabrication yard, the offloading and float-on of the platform from the vessel, the dry transport of the TLP over a long distance, and the final offloading of the platform. Both hydrodynamic and structure analysis were performed to evaluate the behavior of the transport vessel and TLP structure. Special attention was paid to critical areas associated with the use of this new vessel, along with any potential limitations. The results demonstrate that HYSY 278 can effectively be used for transporting the structure with proper arrangement and well-prepared operation. The procedure and details were presented on the basis of the study results. Special attention was also given to discussion on future use based on the results from the analysis.

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

    International Nuclear Information System (INIS)

    Hannah, I.W.

    1978-01-01

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

  7. Wave Loads on Ships Sailing in Restricted Water Depth

    DEFF Research Database (Denmark)

    Vidic-Perunovic, Jelena; Jensen, Jørgen Juncher

    2003-01-01

    depth for a container vessel. The results show that if the water depth is less than two times the draft of the vessel, the wave-induced bending moment becomes significant larger than in deep water with the same sea state description. The peak in the frequency response function for the wave bending......The wave-induced bending moment in ships is the most important sea load parameter for ships larger than 100m in length. Hence, any rational ship design procedure must include a reasonable accurate determination of this load and a large amount of various hydrodynamic formulations have been published......, ranging from semi-empirical formulas to three-dimensional non-linear procedures. A review of the state-of-the art can be found in ISSC.VI.1 (2000). These procedures must be combined with operational and sea state information to predict the probability distribution of the maximum wave-induced bending...

  8. Thermal information regarding the cooldown and operation of liquid argon calorimeters

    International Nuclear Information System (INIS)

    Rucinski, R.A.; Cooper, W.E.; Dixon, K.D.; Krempetz, K.J.; Mulholland, G.T.; Primdahl, K.; Urbin, J.B.

    1993-07-01

    Three liquid argon calorimeters were cooled down and operated as part of the D-Zero detector at Fermi National Accelerator laboratory. The largest vessel contains 248 metric tons of uranium and copper plates and 19 kL (5000 gal.) of liquid argon. The other two vessels are mirror images, each containing 185 metric tons of uranium and stainless steel plates and 12.1 kL (3200 gal.) of liquid argon. The cool down was accomplished by convection heat transfer between boiling liquid nitrogen filled finned heat exchangers and argon gas inside the vessels. Information regarding the general internal geometry of the calorimeters, cool down, operation, and steady state heat loads will be presented

  9. 46 CFR 122.340 - Vessels carrying vehicles.

    Science.gov (United States)

    2010-10-01

    ... vehicles freely in the event of fire or other disaster. The decks, where necessary, must be distinctly... their emergency brakes set when the vessel is underway, and that the motors are not started until the... loading ramp must have its wheels securely blocked, while the vessel is being navigated. (c) The master...

  10. Models for ductile crack initiation and tearing resistance under mode 1 loading in pressure vessel steels

    International Nuclear Information System (INIS)

    Jones, M.R.

    1988-06-01

    Micromechanistic models are presented which aim to predict plane strain ductile initiation toughness, tearing resistance and notched bar fracture strains in pressure vessel steels under monotonically increasing tensile (mode 1) loading. The models for initiation toughness and tearing resistance recognize that ductile fracture proceeds by the growth and linkage of voids with the crack-tip. The models are shown to predict the trend of initiation toughness with inclusion spacing/size ratio and can bound the available experimental data. The model for crack growth can reproduce the tearing resistance of a pressure vessel steel up to and just beyond crack growth initiation. The fracture strains of notched bars pulled in tension are shown to correspond to the achievement of a critical volume fraction of voids. This criterion is combined with the true stress - true strain history of a material point ahead of a blunting crack-tip to predict the initiation toughness. An attempt was made to predict the fracture strains of notched tensile bars by adopting a model which predicts the onset of a shear localization phenomenon. Fracture strains of the correct order are computed only if a ''secondary'' void nucleation event at carbide precipitates is taken into account. (author)

  11. Design loads, loading combinations and structural acceptance criteria for BWR containments in the United States

    International Nuclear Information System (INIS)

    Edwards, N.W.

    1979-01-01

    The definition of loads, loading combinations, and structural acceptance criteria used for the design and evaluation of BWR containments in the Unites States has become much more comprehensive over the past decade. The Mark I pressure suppression containment vessels were designed for a static design pressure, a design temperature, dead load and static equivalent earthquake. The current Mark III containments are being designed to accommodate many more loads such as safety relief valve discharge loads, and suppression pool hydrodynamic loadings associated with the steam condensation phenomena as well as pressure and temperature transients for a range of pipe break sizes. Consistent with the more comprehensive definition of loads and loading combinations, the ASME Code presently establishes structural acceptance criteria with different margins of safety by the definition of Service Level Assignments A, B, C and D. Acting in a responsible manner, United States utilities are currently evaluating and modifying existing containment vessels to account for the more detailed load definition and structural acceptance criteria. (orig.)

  12. Numerical studies of large penetrations and closures for containment vessels subjected to loadings beyond the design basis

    International Nuclear Information System (INIS)

    Kulak, R.F.; Hsieh, B.J.; Kennedy, J.M.; Ash, J.E.; McLennan, G.A.

    1984-01-01

    Numerical simulations of the macro-deformations of the sealing surfaces (gasketed junctures) of a PWR steel containment vessel's equipment hatch and a BWR Mk II containment vessel head have been performed. Results for the equipment hatch juncture indicate that the rotations of the hatch cover and penetration sleeve must be accounted for when performing leakage analysis because they can effect the compression of the gasket even though the gasket is in a pressure-seated configuration. Results from a leakage analysis indicated that excessive leakage can occur if the surface roughness is high and/or the compression set is high. Results for the Mk II head show that both the temperature and pressure loadings must be taken into account to obtain realistic responses. The temperature difference between the flanges and bolts has the important net effect of keeping the gasketed juncture closed, that is in metal-to-metal contact. Due to the high accident temperature, the gasket itself was found to achieve 100% compression set and thus could not perform its sealing function within the juncture

  13. Heat loads on Tore Supra ICRF Launchers Plasma Facing Components

    International Nuclear Information System (INIS)

    Bremond, S.; Colas, L.; Beaumont, B.; Chantant, M.; Goniche, M.; Mitteau, R.

    2005-01-01

    Understanding the heat loads on Ion Cyclotron Range of Frequency (ICRF) launchers plasma-facing components is a crucial task both for operating present tokamaks and for designing ITER ICRF launchers as these loads may limit the RF power coupling capability. Tore Supra facility is particularly well suited to take this issue. Parametric studies have been performed which enables to get an overall detailed picture of the different heat loads on several areas, pointing to different mechanisms at the origin of the heat power fluxes. It is found that the most critical items for Tore-Supra operation are localized heat loads on the Faraday screen top left corner and vertical edges. Warming up close to maximum temperature limit originally set for protection of the plasma-facing components is found of high power pulses, but no erosion was observed after detailed inspection of the launcher in Tore-Supra vessel. Yet, the associated heat loads could be limiting for Tore-Supra operation in the future, and some dedicated work is under progress to improve the understanding of these power fluxes, pointing out the importance of getting a better knowledge of particle flows in the scrape of layer

  14. Deep-water subsea lifting operations

    Energy Technology Data Exchange (ETDEWEB)

    Nestegaard, Arne; Boee, Tormod

    2010-07-01

    Significant costs are related to marine operations in the installation phase of deep water subsea field developments. In order to establish safe operational criteria and procedures for the installation, detailed planning is necessary, including numerical modelling and analysis of the environmental conditions and hydrodynamic loads on the installed object as well as the installation equipment. This paper presents recommendations for modelling and analysis of deep water subsea lifting operations developed for the new DNV RP-H103 [1]. During installation of subsea structures, the highest dynamic forces are most often encountered in the splash zone. Recommendations for estimation of maximum forces will be presented. For small structures and tools, installation through the moon pool of a small installation vessel is often preferred. Calculation methods for loading on structures installed through a moon pool will be presented. During intervention or installation in deep water a significant amplification of amplitude and forces can be experienced when the frequency range of vertical crane tip motion coincides with the natural vertical oscillation of the lift wire and load. Vertical resonance may reduce the operability of the operation. Simplified calculation methods for such operations are presented. (Author)

  15. Proof testing of an explosion containment vessel

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

  16. Operational aspects of the Calder Hall and Chapelcross pressure vessel ultrasonic inspections

    International Nuclear Information System (INIS)

    Bithell, S.J.; Howard, S.R.

    1993-01-01

    As a consequence of the NII's assessment of the Calder Hall and Chapelcross Long Term Safety Review, BNFplc were required to demonstrate the integrity of the Reactor Pressure Vessels through a programme of volumetric seam weld inspection. Existing equipment proved to be inadequate and necessitated the design and manufacture of a remote power manipulator and ultrasonic scanning package. Calder Hall Operations Department and Sellafield Technical Department, working closely with contract staff, completed the first stage of this technically demanding task within 14 months of the project's initiation, resulting in the first deployment of ''REDIMAN'' in March 1991. The design of the new equipment, and the technical and operational difficulties which were overcome by the Inspection Team are outlined. (author)

  17. Containment vessel bottom head transport and lifting technique

    International Nuclear Information System (INIS)

    Zheng Donghong; Tian Shiyong; Hu Dequan; Xiao Hongtao

    2013-01-01

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

  18. 46 CFR 185.340 - Vessels carrying vehicles.

    Science.gov (United States)

    2010-10-01

    ... and away from the vehicles freely in the event of fire or other disaster. The decks, where necessary... motors turned off and their emergency brakes set when the vessel is underway, and that the motors are not... vehicles or next to a loading ramp must have its wheels securely blocked, while the vessel is being...

  19. Increase of cyclic durability of pressure vessels

    International Nuclear Information System (INIS)

    Vorona, V.A.; Zvezdin, Yu.I.

    1980-01-01

    The durability of multilayer pressure vessels under cyclic loading is compared with single-layer vessels. The relative conditional durability is calculated taking into account the assumption on the consequent destruction of layers and viewing a vessel wall as an indefinite plate. It is established that the durability is mainly determined by the number of layers and to a lesser degree depends on the relative size of the defect for the given layer thickness. The advantage of the multilayer vessels is the possibility of selecting layer materials so that to exclude the effect of agressive corrosion media on the strength [ru

  20. Contribution of the different erosion processes to material release from the vessel walls of fusion devices during plasma operation

    International Nuclear Information System (INIS)

    Behrisch, R.

    2002-01-01

    In high temperature plasma experiments several processes contribute to erosion and loss of material from the vessel walls. This material may enter the plasma edge and the central plasma where it acts as impurities. It will finally be re-deposited at other wall areas. These erosion processes are: evaporation due to heating of wall areas. At very high power deposition evaporation may become very large, which has been named ''blooming''. Large evaporation and melting at some areas of the vessel wall surface may occur during heat pulses, as observed in plasma devices during plasma disruptions. At tips on the vessel walls and/or hot spots on the plasma exposed solid surfaces electrical arcs between the plasma and the vessel wall may ignite. They cause the release of ions, atoms and small metal droplets, or of carbon dust particles. Finally, atoms from the vessel walls are removed by physical and chemical sputtering caused by the bombardment of the vessel walls with ions as well as energetic neutral hydrogen atoms from the boundary plasma. All these processes have been, and are, observed in today's plasma experiments. Evaporation can in principle be controlled by very effective cooling of the wall tiles, arcing is reduced by very stable plasma operation, and sputtering by ions can be reduced by operating with a cold plasma in front of the vessel walls. However, sputtering by energetic neutrals, which impinge on all areas of the vessel walls, is likely to be the most critical process because ions lost from the plasma recycle as neutrals or have to be refuelled by neutrals leading to the charge exchange processes in the plasma. In order to quantify the wall erosion, ''materials factors'' (MF) have been introduced in the following for the different erosion processes. (orig.)

  1. Method and system for connecting a loading buoy to a floating vessel. Fremgangsmte og system for tilkopling av en lastebye til et flytende farty

    Energy Technology Data Exchange (ETDEWEB)

    Breivik, K.; Kleppest, H.; Smedal, A.

    1994-07-04

    The invention deals with a method and a system for connecting a submerged loading/unloading buoy to a submerged receiving space in a floating vessel, for transfer of a medium, especially oil, to or from the vessel, wherein the buoy is anchored to the sea bed and is connected to a transfer line for medium. According to one variant of the method, a sink line is lowered from the vessel through the receiving space, an auxiliary buoy being attached to the sink line end, possible via an additional line, and the auxiliary buoy being caused to come to the water surface. A suitably marked pick-up line, which is connected to the buoy, is taken up and connected to the sink line, whereafter the vessel by a positioning means is moved into position above the submerged buoy and said lines are pulled up through the receiving space, so that the buoy is hoisted up and moved to a locking position therein, whereafter the buoy is locked in place in the receiving space. The vessel is provided with a hoisting means to hoist up said lines and the buoy, and also with a service shaft connecting the receiving space to the deck of the vessel. 9 figs.

  2. On the prediction of the reactor vessel integrity under severe accident loadings (RPVSA)

    Energy Technology Data Exchange (ETDEWEB)

    Krieg, R. E-mail: maeule@irs.fzk.de; Devos, J.; Caroli, C.; Solomos, G.; Ennis, P.J.; Kalkhof, D

    2001-11-01

    In order to allow more reliable predictions on the lower head response under core melt-down conditions, the temperature distribution has been analysed including the natural convection in the corium pool. Furthermore, the mechanical models and the failure criteria have been improved based on the RUPTHER and FASTHER experiments where typical temperature gradients are simulated. Lower head local melting as well as corium crust development has been addressed in the CORVIS experiments studying the contact between an alumina/iron thermite and a thick steel plate. The upper head loading by corium impact due to a postulated in-vessel steam explosion has been investigated by the BERDA experiments. Similarity rules were considered such that the results can be directly converted to reactor conditions. Based on these investigations admissible steam explosion energy releases are determined which the upper head can carry. If these limits are not exceeded the reactor containment cannot be endangered by broken head fragments. To provide the necessary basic data, mechanical material tests have been performed.

  3. Emergency Load Shedding Strategy Based on Sensitivity Analysis of Relay Operation Margin against Cascading Events

    DEFF Research Database (Denmark)

    Liu, Zhou; Chen, Zhe; Sun, Haishun Sun

    2012-01-01

    the runtime emergent states of related system component. Based on sensitivity analysis between the relay operation margin and power system state variables, an optimal load shedding strategy is applied to adjust the emergent states timely before the unwanted relay operation. Load dynamics is also taken...... into account to compensate load shedding amount calculation. And the multi-agent technology is applied for the whole strategy implementation. A test system is built in real time digital simulator (RTDS) and has demonstrated the effectiveness of the proposed strategy.......In order to prevent long term voltage instability and induced cascading events, a load shedding strategy based on the sensitivity of relay operation margin to load powers is discussed and proposed in this paper. The operation margin of critical impedance backup relay is defined to identify...

  4. Nuclear reactor vessel decontamination systems

    International Nuclear Information System (INIS)

    McGuire, P. J.

    1985-01-01

    There is disclosed in the present application, a decontamination system for reactor vessels. The system is operatable without entry by personnel into the contaminated vessel before the decontamination operation is carried out and comprises an assembly which is introduced into the vertical cylindrical vessel of the typical boiling water reactor through the open top. The assembly includes a circular track which is centered by guideways permanently installed in the reactor vessel and the track guides opposed pairs of nozzles through which water under very high pressure is directed at the wall for progressively cutting and sweeping a tenacious radioactive coating as the nozzles are driven around the track in close proximity to the vessel wall. The whole assembly is hoisted to a level above the top of the vessel by a crane, outboard slides on the assembly brought into engagement with the permanent guideways and the assembly progressively lowered in the vessel as the decontamination operation progresses. The assembly also includes a low pressure nozzle which forms a spray umbrella above the high pressure nozzles to contain radioactive particles dislodged during the decontamination

  5. Embrittlement of the nuclear icebreaker Lenin reactor pressure vessel materials reconstruction

    International Nuclear Information System (INIS)

    Krasikov, E.A.; Nikolaenko, V.A.

    2008-01-01

    Paper deals with the results of the efforts to examine the radiation damage of the Lenin nuclear-powered ice-breaker decommissioned reactor pressure vessel on the basis of which one has determined the peculiar features of the metal radiation embrittlement. Under 10 10 -10 11 s -1 cm -2 low density neutron flux irradiation one notes the most intensive embrittlement of the metal. Then, as the noxious element content in the metal matrix grows smaller the embrittlement reduces up to the change of sign as to the normal curve plotted at the neutron flux density exceeding 10 13 s -1 cm -2 . One assumes that as a result of the low density neutron flux irradiation the reactor pressure vessel edge spaces at some operation stages may be damaged more severely in contrast to these near the reactor core. The neutron irradiation density is the factor affecting the reactor vessel material embrittlement, that is why, it is important to study the damage mechanism of the materials of the power reactor vessels under design characterized by the low radiation load. The mentioned is important, as well, to evaluate the efficiency of the efforts undertaken to mitigate the effect of the neutron radiation on the reactor vessel [ru

  6. Repairing method for shroud in reactor pressure vessel

    International Nuclear Information System (INIS)

    Watanabe, Yusuke.

    1996-01-01

    The present invention provides a method of repairing a shroud disposed in a pressure vessel of a BWR type reactor. Namely, a baffle plate is disposed on the outer surface of the lower portion of the shroud supported by a shroud support of the pressure vessel. The baffle plate is connected with a lug for securing a shroud head bolt disposed on the outer surface of an upper portion of the shroud by reinforcing members. With such a constitution, when crackings are caused in the shroud, the development of the crackings can be prevented without losing the function of securing the shroud head bolt. Further, if a material having thermal expansion coefficient lower than that of austenite stainless steel is used for the material of the reinforcing member, clamping load to be applied upon attaching the auxiliary member can be reduced. As a result, operation for the attachment is facilitated. (I.S.)

  7. 46 CFR 4.03-40 - Public vessels.

    Science.gov (United States)

    2010-10-01

    ... INVESTIGATIONS Definitions § 4.03-40 Public vessels. Public vessel means a vessel that— (a) Is owned, or demise... Department (except a vessel operated by the Coast Guard or Saint Lawrence Seaway Development Corporation...

  8. Fracture risk assessment for the pressurized water reactor pressure vessel under pressurized thermal shock events

    International Nuclear Information System (INIS)

    Chou, Hsoung-Wei; Huang, Chin-Cheng

    2016-01-01

    Highlight: • The PTS loading conditions consistent with the USNRC's new PTS rule are applied as the loading condition for a Taiwan domestic PWR. • The state-of-the-art PFM technique is employed to analyze a reactor pressure vessel. • Novel flaw model and embrittlement correlation are considered in the study. • The RT-based regression formula of NUREG-1874 was also utilized to evaluate the failure risks of RPV. • For slightly embrittled RPV, the SO-1 type PTSs play more important role than other types of PTS. - Abstract: The fracture risk of the pressurized water reactor pressure vessel of a Taiwan domestic nuclear power plant has been evaluated according to the technical basis of the U.S.NRC's new pressurized thermal shock (PTS) screening criteria. The ORNL's FAVOR code and the PNNL's flaw models were employed to perform the probabilistic fracture mechanics analysis associated with plant specific parameters of the domestic reactor pressure vessel. Meanwhile, the PTS thermal hydraulic and probabilistic risk assessment data analyzed from a similar nuclear power plant in the United States for establishing the new PTS rule were applied as the loading conditions. Besides, an RT-based regression formula derived by the U.S.NRC was also utilized to verify the through-wall cracking frequencies. It is found that the through-wall cracking of the analyzed reactor pressure vessel only occurs during the PTS events resulted from the stuck-open primary safety relief valves that later reclose, but with only an insignificant failure risk. The results indicate that the Taiwan domestic PWR pressure vessel has sufficient structural margin for the PTS attack until either the current license expiration dates or during the proposed extended operation periods.

  9. Rapid Operational Access and Maneuver Support (ROAMS) Platform for Improved Military Logistics Lines of Communication and Operational Vessel Routing

    Science.gov (United States)

    2017-06-01

    discussed in detail. Toolbox: AdH. AdH is a finite element engine capable of solving the 2D and three-dimensional (3D) shallow water equations, the...sites will be the shortest that exists on the mesh. However, the algorithm neither guarantees that the found path will satisfy all navigation...between resolution and computation time. Penalty Function: Draft. The draft constraint ensures that the vessel operates only in sufficiently deep

  10. Intraoperative angiography in reconstructive vessel operations in the lower parts of the body

    International Nuclear Information System (INIS)

    Zehle, A.; Weinhold, C.H.; Hauger, W.

    1981-01-01

    The intraoperative angiography offers decisive advantages in reconstructive vessel operations, because this technique permits a direct and immediate examination of the obtained results and which thus can directly influence the technical and tactic management. Therefore this method allows in the most favourable case to improve prognosis. The technical realization and the procedure are facilitated by the combination of screening method and simple documentation, which is presented here. (orig./MG) [de

  11. 2XIIB vacuum vessel: a unique design

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  12. Electromagnetic forces on a metallic Tokamak vacuum vessel following a disruptive instability

    International Nuclear Information System (INIS)

    Eckhartt, D.

    1979-04-01

    During a 'hard' disruptive instability of a Tokamak plasma the current-carrying plasma is lost within a very short time, typically few milliseconds. If the plasma is contained in a metallic vacuum vessel, electric currents are set up in the vessel following the disappearance of the plasma current. These vessel currents together with the magnetic fields intersecting the vessel generate electromagnetic forces which appear as mechanical loads on the vessel. In the following note it is assumed that the vacuum vessel is surrounded by an 'outer equivalent' or 'flux-conserving' shell having a characteristic time of magnetic field penetration which is long compared to the time of existence of the vessel currents. This property defines the distribution of vessel current densities (and hence the load distribution) without referring to the exact mechanism or time sequence of events by which the plasma current is lost. Numerical examples of the electromagnetic force distribution from this model refer to parameters of the JET-device with the simplifying assumption of circular cross-sections for plasma current, vacuum vessel, and outer equivalent shell. (orig.)

  13. Evaluation on radioactive waste disposal amount of Kori Unit 1 reactor vessel considering cutting and packaging methods

    International Nuclear Information System (INIS)

    Choi, Yu Jong; Lee, Seong Cheol; Kim, Chang Lak

    2016-01-01

    Decommissioning of nuclear power plants has become a big issue in South Korea as some of the nuclear power plants in operation including Kori unit 1 and Wolsung unit 1 are getting old. Recently, Wolsung unit 1 received permission to continue operation while Kori unit 1 will shut down permanently in June 2017. With the consideration of segmentation method and disposal containers, this paper evaluated final disposal amount of radioactive waste generated from decommissioning of the reactor pressure vessel in Kori unit 1 which will be decommissioned as the first in South Korea. The evaluation results indicated that the final disposal amount from the top and bottom heads of the reactor pressure vessel with hemisphere shape decreased as they were cut in smaller more effectively than the cylindrical part of the reactor pressure vessel. It was also investigated that 200 L and 320 L radioactive waste disposal containers used in Kyung-Ju disposal facility had low payload efficiency because of loading weight limitation

  14. The NINO [No Inspector, No Operator system] cask-loading safeguards system

    International Nuclear Information System (INIS)

    Fiarman, S.

    1987-01-01

    It is, in general difficult to determine by means of camera-surveillance techniques what is loaded into spent-fuel casks being prepared for shipment from light-water reactors to other reactors, reprocessing facilities, or long-term storage. Furthermore, the expected high frequency of cask loadings in the coming years would place too great a burden on the IAEA and Euratom inspectorates if each had to be observed by an inspector. For the case of shipment to other reactors and reprocessing facilities, the casks are soon opened and, in principle, their contents could be ascertained by direct inspection. In the case of long-term-storage facilities, the casks would stay sealed for years, thereby requiring the IAEA to know positively how many spent-fuel assemblies were loaded at the reactor and to have a continuity of knowledge of the cask's contents. It has been proposed instead that the facility operator place the cask seal on the cask within the field of view of a surveillance system linked to the cask seal. This solution, however, may not provide enough credibility for acceptance by the safeguards community. This paper presents an alternative to both inspector presence at cask loading and operator assistance in applying seals; this alternative is called the No Inspector, No Operator system (NINO)

  15. The TPX vacuum vessel and in-vessel components

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  16. Heavy-Section Steel Technology Program intermediate-scale pressure vessel tests

    International Nuclear Information System (INIS)

    Bryan, R.H.; Merkle, J.G.; Smith, G.C.; Whitman, G.D.

    1977-01-01

    The tests of intermediate-size vessels with sharp flaws permitted the comparison of experimentally observed behavior with analytical predictions of the behavior of flawed pressure vessels. Fracture strains estimated by linear elastic fracture mechanics (LEFM) were accurate in the cases in which the flaws resided in regions of high transverse restraint and the fracture toughness was sufficiently low for unstable fracture to occur prior to yielding through the vessel wall. When both of these conditions were not present, unstable fracture did occur, always preceded by stable crack growth; and the cylinders with flaws initially less than halfway through the wall attained gross yield prior to burst. Predictions of failure pressure of the vessels with flawed nozzles, based upon LEFM estimates of failure strain, were very conservative. LEFM calculations of critical load were based upon small-specimen fracture toughness test data. Whenever gross yielding preceded failure, the actual strains achieved were considerably greater than the estimated strains at failure based on LEFM. In such cases the strength of the vessel may be no longer dependent upon plane-strain fracture toughness but upon the capacity of the cracked section to carry the imposed load stably in the plastic range. Stable crack growth, which has not been predictable quantitatively, is an important factor in elastic-plastic analysis of strength. The ability of the flawed vessels to attain gross yield in unflawed sections has important qualitative implications on pressure vessel safety margins. The gross yield condition occurs in light-water-reactor pressure vessels at about 2 x design pressure. The intermediate vessel tests that demonstrated a capacity for exceeding this load confirm that the presumed margin of safety is not diminished by the presence of flaws of substantial size, provided that material properties are adequate

  17. Fabrication of High Temperature and High Pressure Vessel for the Fuel Test

    International Nuclear Information System (INIS)

    Park, Kook Nam; Lee, Jong Min; Sim, Bong Shick; Shon, Jae Min; Ahn, Seung Ho; Yoo, Seong Yeon

    2007-01-01

    The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR and CANDU nuclear power plants has been developed and installed in HANARO, KAERI. It is consisted of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS which is located inside the pool is divided into 3-parts; they are in-pool pipes, IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The localization of the IVA is achieved by manufacturing through local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique of the instrument lines has been checked for its functionality and yield. A IVA has been manufactured by local technique and will be finally tested under out of the high temperature and high pressure test

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

    Directory of Open Access Journals (Sweden)

    Li Liangchao

    2017-01-01

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

  19. Study on fatigue analysis for operational load histories

    International Nuclear Information System (INIS)

    Wilhelm, Paul; Rudolph, Juergen; Steinmann, Paul

    2013-01-01

    Some laboratories performed fatigue tests in dissolved oxygen water at elevated temperature to better understand the influence of a long hold-time within cyclic loading. Also, the combined effect of complex waveform and surface finish was examined. The data show a less severe influence compared to the prediction model from Argonne National Laboratory; an increase in fatigue life was noticed and attributed to different effects. To evaluate an operational load history with this experimental data an algorithm is developed, which finds hold-times and the examined complex waveform in a stress-time series. All those cycles, which are either geometrically comparable to the complex loading signal or containing a hold period, are evaluated with the test results and not with the formula from Argonne National Laboratory. The reduction of the cumulative usage factor is calculated. Based on this discussion a realistic test condition is derived for further research activities.

  20. Design of High Temperature Reactor Vessel Using ANSYS Software

    International Nuclear Information System (INIS)

    Bandriyana; Kasmudin

    2003-01-01

    Design calculation and evaluation of material strength for high temperature reactor vessel based on the design of HTR-10 high temperature reactor vessel were carried out by using the ANSYS 5.4 software. ANSYS software was applied to calculate the combined load from thermal and pressure load. Evaluation of material strength was performed by calculate and determine the distribution of temperature, stress and strain in the thickness direction of vessel, and compared with its material strength for designed. The calculation was based on the inner wall temperature of vessel of 600 o C and the outer temperature of 500 and 600 o C. Result of calculation gave the maximum stress for outer temperature of 600 o C was 288 N/ mm 2 and strain of 0.000187. For outer temperature of 500 o C the maximum stress was 576 N/ mm 2 and strain of 0.003. Based on the analysis result, the material of steel SA 516-70 with limited stress for design of 308 N/ mm 2 can be used for vessel material with outer wall temperature of 600 o C

  1. Evaluation of structural reliability for vacuum vessel under external pressure and electromagnetic force

    International Nuclear Information System (INIS)

    Minato, Akio

    1983-08-01

    Static and dynamic structural analyses of the vacuum vessel for a Swimming Pool Type Tokamak Reactor (SPTR) have been conducted under the external pressure (hydraulic and atmospheric pressure) during normal operation or the electromagnetic force due to plasma disruption. The reactor structural design is based on the concept that the adjacent modules of the vacuum vessel are not connected mechanically with bolts in the torus inboard region each other, so as to save the required space for inserting the remote handling machine for tightenning and untightenning bolts in the region and to simplify the repair and maintenance of the reactor. The structural analyses of the vacuum vessel have been carried out under the external pressure and the electromagnetic force and the structural reliability against the static and dynamic loads is estimated. The several configurations of the lip seal between the modules, which is required to make a plasma vacuum boundary, have been proposed and the structural strength under the forced displacements due to the deformation of the vacuum vessel is also estimated. (author)

  2. Application of load follow operation to equilibrium cycle of OPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyeongju; Choe, Jiwon; Lee, Deokjung [Interdisciplinary School of Green Energy, Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2012-10-15

    All nuclear power plants in Korea are operated at a base load, that is 100% rated power, and do not rely largely on power tracking control except for startup, shutdown, and some minor problem occurrences. However, if the electricity from nuclear power plants exceeds 50% of total electricity generation according to national energy plan, load follow operation is necessary to efficiently use the electrical energy. But it is very difficult to control the axial power distribution and reactor core reactivity at the same time as needed because of variations in nuclear system parameters. In 1990s, an advanced reactor control algorithm, Mode-K, was developed which uses regulation banks, boron control, and a heavy-worth bank (H-bank). The regulation banks and boron control are used for core reactivity control and the H-bank is used for the control of axial power shape. In this study, reactor core simulations with HELIOS/MASTER code system using Mode-K strategy are applied to the daily load follow operation in equilibrium cycle of OPR1000.

  3. Warm pre-stress experiments on highly irradiated reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Landron, C.; Ait-Bachir, M.; Moinereau, D.; Molinie, E.; Garbay, E.

    2015-01-01

    In the aim to justify in-service integrity of reactor pressure vessel beyond 40 years, experimental warm pre-stress (WPS) tests were performed on irradiated materials representative of RPV steels corresponding to 40 operating years. Different types of WPS loading path have been considered to cover typical postulated accidental transients. These results confirmed the beneficial effect of WPS on the cleavage fracture resistance of the irradiated materials. No fracture occurred during the cooling phase of the loading path and the fracture toughness values are higher than that measured with conventional isothermal tests. The analyses of the experiments, conducted using either simplified engineering models or more refined fracture models based on local approach to cleavage fracture, are in agreement with the experimental results. (authors)

  4. Increasing the flexibility of base-load generating units in operation on fossil fuel

    Energy Technology Data Exchange (ETDEWEB)

    Girshfel' d, V Ya; Khanaev, V A; Volkova, E D; Gorelov, V A; Gershenkroi, M L

    1979-01-01

    Increasing the flexibility of base-load generating units operating on fossil fuel by modifying them is a necessary measure. The highest economic effect is attained with modification of gas- and oil-fired generating units in the Western United Power Systems of the European part of the SPSS. On the basis of available experience, 150- and 200-MW units can be extensively used to regulate the power in the European part of the SPSS through putting them into reserve for the hours of the load dip at night. The change under favorable conditions of 150- and 200-MW units operating on coal to a district-heating operating mode does not reduce the possibilities for flexible operation of these units because it is possible greatly to unload the turbines while the minimum load level of the pulverized fuel fired boiler is retained through transferring a part of the heat load to the desuperheater. It is necessary to accumulate and analyze experience with operation of generating units (especially of supercritical units) with regular shutdowns and starts of groups of units and to solve the problems of modification of generating units, with differentiation with respect to types of fuel and to the united power supply system.

  5. The vacuum vessel for the FTU device: design constraints and stress analysis

    International Nuclear Information System (INIS)

    Andreani, R.; Cecchini, A.; Gasparotto, M.; Lovisetto, L.; Migliori, S.; Pizzuto, A.

    1984-01-01

    The FTU vacuum vessel must withstand large electromagnetic loads due to the interactions between the eddy currents in the vessel and high magnetic fields of the machine, the atmospheric pressure and the severe thermal loads due to plasma losses and RF power not coupled to the plasma. In order to minimise the stresses on the vacuum chamber, an optimization of the wall thickness has been performed and, in order to assess the feasibility of the vessel, an extensive three dimensional finite element stress analysis has been developed. The main results obtained are illustrated. (author)

  6. Fatigue evaluation in reactor vessel components

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel; Miranda, Carlos A. de J.

    1994-01-01

    This paper presents a sequence of increasing complexity forms of evaluating fatigue damage of nuclear pressure vessel components caused by cycling loadings. Examples are included in order to illustrate such procedures. (author)

  7. Identification of critical equipment and determination of operational limits in helium refrigerators under pulsed heat load

    Science.gov (United States)

    Dutta, Rohan; Ghosh, Parthasarathi; Chowdhury, Kanchan

    2014-01-01

    Large-scale helium refrigerators are subjected to pulsed heat load from tokamaks. As these plants are designed for constant heat loads, operation under such varying load may lead to instability in plants thereby tripping the operation of different equipment. To understand the behavior of the plant subjected to pulsed heat load, an existing plant of 120 W at 4.2 K and another large-scale plant of 18 kW at 4.2 K have been analyzed using a commercial process simulator Aspen Hysys®. A similar heat load characteristic has been applied in both quasi steady state and dynamic analysis to determine critical stages and equipment of these plants from operational point of view. It has been found that the coldest part of both the cycles consisting JT-stage and its preceding reverse Brayton stage are the most affected stages of the cycles. Further analysis of the above stages and constituting equipment revealed limits of operation with respect to variation of return stream flow rate resulted from such heat load variations. The observations on the outcome of the analysis can be used for devising techniques for steady operation of the plants subjected to pulsed heat load.

  8. The nature of operating flight loads and their effect on propulsion system structures

    Science.gov (United States)

    Dickenson, K. H.; Martin, R. L.

    1981-01-01

    Past diagnostics studies revealed the primary causes of performance deterioration of high by-pass turbofan engines to be flight loads, erosion, and thermal distortion. The various types of airplane loads that are imposed on the engine throughout the lifetime of an airplane are examined. These include flight loads from gusts and maneuvers and ground loads from takeoff, landing, and taxi conditions. Clarification is made in definitions of the airframer's limit and ultimate design loads and the engine manufacturer's operating design loads. Finally, the influence of these loads on the propulsion system structures is discussed.

  9. Scaled Testing to Evaluate Pulse Jet Mixer Performance in Waste Treatment Plant Mixing Vessels

    International Nuclear Information System (INIS)

    Fort, James A.; Meyer, Perry A.; Bamberger, Judith A.; Enderlin, Carl W.; Scott, Paul A.; Minette, Michael J.; Gauglitz, Phillip A.

    2010-01-01

    The Waste Treatment and Immobilization Plant (WTP) at Hanford is being designed and built to pre-treat and vitrify the waste in Hanford's 177 underground waste storage tanks. Numerous process vessels will hold waste at various stages in the WTP. These vessels have pulse jet mixer (PJM) systems. A test program was developed to evaluate the adequacy of mixing system designs in the solids-containing vessels in the WTP. The program focused mainly on non-cohesive solids behavior. Specifically, the program addressed the effectiveness of the mixing systems to suspend settled solids off the vessel bottom, and distribute the solids vertically. Experiments were conducted at three scales using various particulate simulants. A range of solids loadings and operational parameters were evaluated, including jet velocity, pulse volume, and duty cycle. In place of actual PJMs, the tests used direct injection from tubes with suction at the top of the tank fluid. This gave better control over the discharge duration and duty cycle and simplified the facility requirements. The mixing system configurations represented in testing varied from 4 to 12 PJMs with various jet nozzle sizes. In this way the results collected could be applied to the broad range of WTP vessels with varying geometrical configurations and planned operating conditions. Data for 'just-suspended velocity', solids cloud height, and solids concentration vertical profile were collected, analyzed, and correlated. The correlations were successfully benchmarked against previous large-scale test results, then applied to the WTP vessels using reasonable assumptions of anticipated waste properties to evaluate adequacy of the existing mixing system designs.

  10. The effect of a self-balancing through wall residual stress distribution on the extension of a through-wall crack in a pressure vessel

    International Nuclear Information System (INIS)

    Smith, E.

    1993-01-01

    Leak-before-break arguments for pressurized components involve a comparison of the critical size of crack that will grow unstably under accident loadings and the critical leakage crack size for normal operation loadings. The paper is concerned with the former crack size and particularly with regard to the effect of residual stresses on the critical unstable crack size. Results from an analysis of a simple simulation model are used to provide underpinning for the view, expressed by Green and Knowles at the 1992 American Society of Mechanical Engineers Pressure Vessel and Piping Conference, that self-balancing through-wall residual stresses have little overall effect on the extension of a through-wall crack in a pressure vessel

  11. Expert System Models for Forecasting Forklifts Engagement in a Warehouse Loading Operation: A Case Study

    Directory of Open Access Journals (Sweden)

    Dejan Mirčetić

    2016-08-01

    Full Text Available The paper focuses on the problem of forklifts engagement in warehouse loading operations. Two expert system (ES models are created using several machine learning (ML models. Models try to mimic expert decisions while determining the forklifts engagement in the loading operation. Different ML models are evaluated and adaptive neuro fuzzy inference system (ANFIS and classification and regression trees (CART are chosen as the ones which have shown best results for the research purpose. As a case study, a central warehouse of a beverage company was used. In a beverage distribution chain, the proper engagement of forklifts in a loading operation is crucial for maintaining the defined customer service level. The created ES models represent a new approach for the rationalization of the forklifts usage, particularly for solving the problem of the forklifts engagement incargo loading. They are simple, easy to understand, reliable, and practically applicable tool for deciding on the engagement of the forklifts in a loading operation.

  12. Optimal control of load-following operations in a pressurized water reactor

    International Nuclear Information System (INIS)

    Zhao Fuyu; Zhou Dawei

    2000-01-01

    According to the optimal control theory, the problem of load-following operation in a pressurized water reactor is formulated as a nonlinear-quadratic optimal control problem. One-dimensional core model is adopted. A successful optimization algorithm DDPSR is proposed to solving the obtained problem. The research results show that the DDPSR can converge with a long time interval and needs very small iteration number and computing time, and the practical reactor can be fairly operated in an optimal load-following manner and axial offset satisfies the required value from beginning to end. Control characters of boron concentration are discussed specially

  13. Development of Fast-Running Simulation Methodology Using Neural Networks for Load Follow Operation

    International Nuclear Information System (INIS)

    Seong, Seung-Hwan; Park, Heui-Youn; Kim, Dong-Hoon; Suh, Yong-Suk; Hur, Seop; Koo, In-Soo; Lee, Un-Chul; Jang, Jin-Wook; Shin, Yong-Chul

    2002-01-01

    A new fast-running analytic model has been developed for analyzing the load follow operation. The new model was based on the neural network theory, which has the capability of modeling the input/output relationships of a nonlinear system. The new model is made up of two error back-propagation neural networks and procedures to calculate core parameters, such as the distributions and density of xenon in a quasi-steady-state core like load follow operation. One neural network is designed to retrieve the axial offset of power distribution, and the other is for reactivity corresponding to a given core condition. The training data sets for learning the neural networks in the new model are generated with a three-dimensional nodal code and, also, the measured data of the first-day test of load follow operation. Using the new model, the simulation results of the 5-day load follow test in a pressurized water reactor show a good agreement between the simulation data and the actual measured data. Required computing time for simulating a load follow operation is comparable to that of a fast-running lumped model. Moreover, the new model does not require additional engineering factors to compensate for the difference between the actual measurements and analysis results because the neural network has the inherent learning capability of neural networks to new situations

  14. JT-60SA vacuum vessel manufacturing and assembly

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-08-15

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

  15. Recalculation of loads on LOFT penetrations 1A, 2A, 3E, 3F, 5A, 5B, 5C, 5D, 5E, 5F, 7A, 9A, 11C, 17A, 17B, 20A, 20B, 20C, 21A

    International Nuclear Information System (INIS)

    McFadden, D.F.

    1978-01-01

    The loads on the piping nozzles penetrating the LOFT containment vessel are presented. Since the design and construction of the containment vessel, piping systems that penetrate the nozzles have been changed. Consequently, the moments and forces on the containment vessel are different than those stated in LOFT specification S-1. Two combined load cases were tabulated, deadweight plus thermal expansion plus operating basis earthquake and deadweight plus safe shutdown earthquake

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  17. Efficient approach to simulate EM loads on massive structures in ITER machine

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, A. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Andreeva, Z.; Belov, A.; Belyakov, V.; Filatov, O. [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation); Gribov, Yu.; Ioki, K. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Kukhtin, V.; Labusov, A.; Lamzin, E.; Lyublin, B.; Malkov, A.; Mazul, I. [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation); Rozov, V.; Sugihara, M. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Sychevsky, S., E-mail: sytch@sintez.niiefa.spb.su [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation)

    2013-10-15

    Highlights: ► A modelling technique to predict EM loads in ITER conducting structures is presented. ► The technique provides low computational cost and parallel computations. ► Detailed models were built for the system “vacuum vessel, cryostat, thermal shields”. ► EM loads on massive in-vessel structures were simulated with the use of local models. ► A flexible combination of models enables desired accuracy of load distributions. -- Abstract: Operation of the ITER machine is associated with high electromagnetic (EM) loads. An essential contributor to EM loads is eddy currents induced in passive conductive structures. Reasoning from the ITER construction, a modelling technique has been developed and applied in computations to efficiently predict anticipated loads. The technique allows us to avoid building a global 3D finite-element (FE) model that requires meshing of the conducting structures and their vacuum environment into 3D solid elements that leads to high computational cost. The key features of the proposed technique are: (i) the use of an existing shell model for the system “vacuum vessel (VV), cryostat, and thermal shields (TS)” implementing the magnetic shell approach. A solution is obtained in terms of a single-component, in this case, vector electric potential taken within the conducting shells of the “VV + cryostat + TS” system. (ii) EM loads on in-vessel conducting structures are simulated with the use of local FE models. The local models use either the 3D solid body or shell approximations. Reasoning from the simulation efficiency, the local boundary conditions are put with respect to the total field or an external field. The use of an integral-differential formulation and special procedures ensures smooth and accurate simulated distributions of fields from current sources of any geometry. The local FE models have been developed and applied for EM analyses of a variety of the ITER components including the diagnostic systems

  18. Combining operational models and data into a dynamic vessel risk assessment tool for coastal regions

    Science.gov (United States)

    Fernandes, R.; Braunschweig, F.; Lourenço, F.; Neves, R.

    2016-02-01

    The technological evolution in terms of computational capacity, data acquisition systems, numerical modelling and operational oceanography is supplying opportunities for designing and building holistic approaches and complex tools for newer and more efficient management (planning, prevention and response) of coastal water pollution risk events. A combined methodology to dynamically estimate time and space variable individual vessel accident risk levels and shoreline contamination risk from ships has been developed, integrating numerical metocean forecasts and oil spill simulations with vessel tracking automatic identification systems (AIS). The risk rating combines the likelihood of an oil spill occurring from a vessel navigating in a study area - the Portuguese continental shelf - with the assessed consequences to the shoreline. The spill likelihood is based on dynamic marine weather conditions and statistical information from previous accidents. The shoreline consequences reflect the virtual spilled oil amount reaching shoreline and its environmental and socio-economic vulnerabilities. The oil reaching shoreline is quantified with an oil spill fate and behaviour model running multiple virtual spills from vessels along time, or as an alternative, a correction factor based on vessel distance from coast. Shoreline risks can be computed in real time or from previously obtained data. Results show the ability of the proposed methodology to estimate the risk properly sensitive to dynamic metocean conditions and to oil transport behaviour. The integration of meteo-oceanic + oil spill models with coastal vulnerability and AIS data in the quantification of risk enhances the maritime situational awareness and the decision support model, providing a more realistic approach in the assessment of shoreline impacts. The risk assessment from historical data can help finding typical risk patterns ("hot spots") or developing sensitivity analysis to specific conditions, whereas real

  19. Automatic Power Control for Daily Load-following Operation using Model Predictive Control Method

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Keuk Jong; Kim, Han Gon [KH, Daejeon (Korea, Republic of)

    2009-10-15

    Under the circumstances that nuclear power occupies more than 50%, nuclear power plants are required to be operated on load-following operation in order to make the effective management of electric grid system and enhanced responsiveness to rapid changes in power demand. Conventional reactors such as the OPR1000 and APR1400 have a regulating system that controls the average temperature of the reactor core relation to the reference temperature. This conventional method has the advantages of proven technology and ease of implementation. However, this method is unsuitable for controlling the axial power shape, particularly the load following operation. Accordingly, this paper reports on the development of a model predictive control method which is able to control the reactor power and the axial shape index. The purpose of this study is to analyze the behavior of nuclear reactor power and the axial power shape by using a model predictive control method when the power is increased and decreased for a daily load following operation. The study confirms that deviations in the axial shape index (ASI) are within the operating limit.

  20. Boundary element analysis of stress due to thermal shock loading or reactor pressure vessel nozzle; Napetostna analiza pri nestacionarni termicni obremenitvi cevnega prikljucka reaktorske tlacne posode z metodo robnih elementov

    Energy Technology Data Exchange (ETDEWEB)

    Kramberger, J; Potrc, I [Tehniska fakulteta, Maribor (Yugoslavia)

    1989-07-01

    Apart from being exposed to the primary loading of internal pressure and steady temperature field, the reactor pressure vessel is also subject to various thermal transients (thermal shocks). Theoretical and experimental stress analyses show that severe material stresses occur in the nozzle area of the pressure vessel which may lead to defects (cracks). It has been our aim to evaluate these stresses by the use of the Boundary Element method. (author)

  1. Autonomic adjusting of activity of cardio-vessel system of girls of the prepubescence period, engaged in dancing aerobic

    Directory of Open Access Journals (Sweden)

    Romanchuk А.P.

    2010-03-01

    Full Text Available The vegetative adjusting of activity of the cardio-vessel system of girls is investigational. In an experiment took part girls of 9-12 years old. A type of sport is a dancing aerobics. The indexes of general power of spectrum of variability of cardiac rhythm are presented, systole and diastole arteriotony. Motion of the adaptation re-erecting is appraised under influence of the physical loadings in the conditions of current and operative control. Certain change activity and tone of vegetative influences on the cardio-vessel system. They determine the features of the further adaptation re-erecting in an organism.

  2. An investigation into the RCCI engine operation under low load and its achievable operational range at different engine speeds

    International Nuclear Information System (INIS)

    Wang, Yifeng; Zhu, ZhongWen; Yao, Mingfa; Li, Tie; Zhang, Weijing; Zheng, Zunqing

    2016-01-01

    Highlights: • The response of allowable RCCI operating range to engine speed variation is studied. • The RCCI and diesel LTC engine operations at are compared at low engine load. • The potential of expanding RCCI operating range at low engine speed is explored. - Abstract: Reactivity controlled compression ignition (RCCI) is demonstrated as a promising combustion strategy to achieve high efficiency and clean combustion. However, less effort has been devoted to examine the achievable RCCI operational range over a wide range of engine speed. In addition, previous studies have found that superior EGR rate and high diesel/gasoline fuel ratio are required to ease the extension of the low-load operating range of RCCI regime. Even then, relatively high CO and HC (unburned hydrocarbon) emissions and the accompanying fuel con-sum ption penalty still remain a problem to be resolved. Therefore, in this work the potential of diesel-fueled LTC to achieve simultaneously low NOx and soot emissions while maintaining high thermal efficiency at low load (IMEP ≈0.23–0.26 MPa) is investigated and compared with the gasoline/diesel RCCI strategy. The results show that the diesel LTC operation can yield slightly higher soot and NOx emissions (soot: 0.002 g/kW h, NOx: 0.446 g/kW h), but CO and HC emissions as well as the fuel consumption are much lower than the RCCI strategy, implying the diesel LTC regime may be more suitable for low-load operations. In addition, the RCCI operational range at speeds ranging from 900 to 2500 r/min is determined, the results show that the maximum achievable load (IMEP) increases with an increase in speed, and a maximum IMEP of 1.2 MPa can be achieved at an engine speed of 2300 r/min. Ultra-low NOx and soot emissions (soot < 0.003 g/kW h, NOx < 0.4 g/kW h) can be achieved under the maximum loading conditions at each speed investigated. However, high levels of CO and HC emissions still remain a big problem to be solved. The lowest fuel consumption

  3. Study on operation conditions and an operation system of a nuclear powered submersible research vessel, 'report of working group on application of a very small nuclear reactor to an ocean research'

    International Nuclear Information System (INIS)

    Ura, Tamaki; Takamasa, Tomoji; Nishimura, Hajime

    2001-07-01

    JAERI has studied on design of a nuclear powered submersible research vessel, which will navigate under sea mainly in the Arctic Ocean, as a part of the design activity of advanced marine reactors. This report describes operation conditions and an operating system of the vessel, which were discussed by the specialists of hull design, sound positioning, ship motions and oceanography, etc. The design conditions on ship motions for submersible vessels were surveyed considering regulations in our country, and ship motions were evaluated in the cases of underwater and surface navigations taking account of observation activities in the Arctic Ocean. The effect of ship motions on the compact nuclear reactor SCR was assessed. A submarine transponder system and an on-ice communication buoy system were examined as a positioning and communication system, supposing the activity under ice. The interval between transponders or communication buoys was recommended as 130 km. Procedures to secure safety of nuclear powered submersible research vessel were discussed according to accidents on the hull or the nuclear reactor. These results were reflected to the concept of the nuclear powered submersible research vessel, and subjects to be settled in the next step were clarified. (author)

  4. Mass transfer experiments for the heat load during in-vessel retention of core melt

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hae Kyun; Chung, Bum Jin [Dept. of Nuclear Engineering, Kyung Hee University, Seoul (Korea, Republic of)

    2016-08-15

    We investigated the heat load imposed on the lower head of a reactor vessel by the natural convection of the oxide pool in a severe accident. Mass transfer experiments using a CuSO{sub 4}–H{sub 2}SO{sub 4} electroplating system were performed based on the analogy between heat and mass transfer. The Ra′{sub H} of 10{sup 14} order was achieved with a facility height of only 0.1 m. Three different volumetric heat sources were compared; two had identical configurations to those previously reported, and the other was designed by the authors. The measured Nu's of the lower head were about 30% lower than those previously reported. The measured angular heat flux ratios were similar to those reported in existing studies except for the peaks appearing near the top. The volumetric heat sources did not affect the Nu of the lower head but affected the Nu of the top plate by obstructing the rising flow from the bottom.

  5. Operation of the lithium pellet injector during D-T operations on TFTR

    International Nuclear Information System (INIS)

    Barnes, G.W.; Gernhardt, R.C.; Mansfield, D.

    1995-01-01

    The Lithium Pellet Injector is used as an operational tool to condition the vacuum vessel walls prior to and after plasma formation as well as to support diagnostics using the resulting neutral or spectral emissions. This paper addresses the injector operational issues that have been experienced during D-T experimental operations. Reliability enhancements are discussed, such as modification to the pellet magazine, gas purging and exhaust handling. Nuclear boundary line breaks for routine magazine wheel changeout are addressed. Procedures that were developed and utilized to ensure ALARA (As Low As Reasonably Achievable) during line breaks are discussed. Lithium loading of the pellet wheel and the special techniques that are required are described

  6. Pressure vessel integrity and weld inspection procedure

    International Nuclear Information System (INIS)

    Solomon, K.A.; Okrent, D.; Kastenberg, W.E.

    1975-01-01

    The primary objective of this paper is to develop a simple methodology which, when coupled with existing observations on pressure vessel behavior, provides an inter-relation between pressure vessel integrity, and the parameters of the in-service inspection program, including inspection sample size, frequency and efficiency. A modified Markov process is employed and a computer code was written to obtain numerical results. The Markov process mathematically describes the following physical events. In a nuclear reactor pressure vessel weld, some defects may exist prior to the zeroth inspection (i.e., prior to vessel operation). During the zeroth inspection and repair processes, some of these defects are removed. During the first cycle of vessel operation, the existing defects may grow and some new defects may be generated. Those defects that are found at the first (and succeeding) inspection interval and warrant repair, are repaired. The above process continues through several operating cycles to the end of vessel life. During any inspection, only a portion of the welds may be inspected, and with less than perfect efficiency

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  8. Reactor operational transient analysis

    International Nuclear Information System (INIS)

    Shin, W.K.; Chae, S.K.; Han, K.I.; Yang, K.S.; Chung, H. D.; Kim, H.G.; Moon, H.J.; Ryu, Y.H.

    1983-01-01

    To build up efficient capability of safety review and inspection for the nuclear power plants, four area of studies have performed as follows: 1) In order to search the most optimized operating method during load follow operating schemes, automatic control and normal control, are compared each other under the CAOC condition. The analysis performed by DDID code has shown that the reactor has to be controlled by the operator manually during load follow operation. 2) Through the sensitivity analysis by COBRA code, the operating parameters, such as coolant pressure, flow rate, inlet temperature, and power distribution are shown to be important to the determination of DNBR. Expecially, inlet temperature of primary coolant system is appeared as the most senstive parameter on DNBR. 3) FRAPCON code is adapted to study the sensitivity of several operational parameters on the mechanical properties of reactor fuel rod. 4) The calculations procedure which is required to be obtained the neutron fluence at the reactor vessel and the spectrum at the surveillance capsule is established. The results of computation are conpared with those of FSAR and SWRI report and proved its applicability to reactor surveillance program. (Author)

  9. Verification and application of MASTER for load follow operation

    International Nuclear Information System (INIS)

    Park, Yong Soo; Cho, Byung Oh; Lee, Chang Ho; Jung, Yil Sup; Park, Chan Oh

    1996-01-01

    The xenon dynamics module in the nuclear design code MASTER was verified through a simulation calculation. The simulation result shows that the xenon dynamics module in MASTER can trace and predict xenon behavior with accuracy under any core transient state and therefore can simulate load follow operations

  10. The probabilistic structural integrity assessment of reactor pressure vessels under pressurized thermal shock loading

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Mingya, E-mail: chenmingya@cgnpc.com.cn [Suzhou Nuclear Power Research Institute, 215004 Suzhou, Jiangsu Province (China); Lu, Feng; Wang, Rongshan; Yu, Weiwei [Suzhou Nuclear Power Research Institute, 215004 Suzhou, Jiangsu Province (China); Wang, Donghui [State Nuclear Power Plant Service Company, 200237 Shanghai (China); Zhang, Guodong; Xue, Fei [Suzhou Nuclear Power Research Institute, 215004 Suzhou, Jiangsu Province (China)

    2015-12-01

    Highlights: • The methodology and the case study of the FAVOR software were shown. • The over-conservative parameters in the DFM were shown. • The differences between the PFM and the DFM were discussed. • The limits in the current FAVOR were studied. - Abstract: The pressurized thermal shock (PTS) event poses a potentially significant challenge to the structural integrity of the reactor pressure vessel (RPV) during the long time operation (LTO). In the USA, the “screening criteria” for maximum allowable embrittlement of RPV material, which forms part of the USA regulations, is based on the probabilistic fracture mechanics (PFM). The FAVOR software developed by Oak Ridge National Laboratory (ORNL) is used to establish the regulation. As the technical basis of FAVOR is not the most widely-used and codified methodologies, such as the ASME and RCC-M codes, in most countries (with exception of the USA), proving RPV integrity under the PTS load is still based on the deterministic fracture mechanics (DFM). As the maximum nil-ductility-transition temperature (RT{sub NDT}) of the beltline material for the 54 French RPVs after 40 years operation is higher than the critical values in the IAEA-TECDOC-1627 and European NEA/CSNI/R(99)3 reports (while still obviously lower than the “screening criteria” of the USA), it may conclude that the RPV will not be able to run in the LTO based on the DFM. In the FAVOR, the newest developments of fracture mechanics are applied, such as the warm pre-stress (WPS) effect, more accurate estimation of the flaw information and less conservation of the toughness (such as the three-parameter Weibull distribution of the fracture toughness). In this paper, the FAVOR software is first applied to show both the methodology and the results of the PFM, and then the limits in the current FAVOR software (Version 6.1, which represents the baseline for re-assessing the regulation of 10 CFR 50.61), lack of the impact of the constraint effect

  11. Testing of Full Scale Flight Qualified Kevlar Composite Overwrapped Pressure Vessels

    Science.gov (United States)

    Greene, Nathanael; Saulsberry, Regor; Yoder, Tommy; Forsyth, Brad; Thesken, John; Phoenix, Leigh

    2007-01-01

    Many decades ago NASA identified a need for low-mass pressure vessels for carrying various fluids aboard rockets, spacecraft, and satellites. A pressure vessel design known as the composite overwrapped pressure vessel (COPV) was identified to provide a weight savings over traditional single-material pressure vessels typically made of metal and this technology has been in use for space flight applications since the 1970's. A typical vessel design consisted of a thin liner material, typically a metal, overwrapped with a continuous fiber yarn impregnated with epoxy. Most designs were such that the overwrapped fiber would carry a majority of load at normal operating pressures. The weight advantage for a COPV versus a traditional singlematerial pressure vessel contributed to widespread use of COPVs by NASA, the military, and industry. This technology is currently used for personal breathing supply storage, fuel storage for auto and mass transport vehicles and for various space flight and aircraft applications. The NASA Engineering and Safety Center (NESC) was recently asked to review the operation of Kevlar 2 and carbon COPVs to ensure they are safely operated on NASA space flight vehicles. A request was made to evaluate the life remaining on the Kevlar COPVs used on the Space Shuttle for helium and nitrogen storage. This paper provides a review of Kevlar COPV testing relevant to the NESC assessment. Also discussed are some key findings, observations, and recommendations that may be applicable to the COPV user community. Questions raised during the investigations have revealed the need for testing to better understand the stress rupture life and age life of COPVs. The focus of this paper is to describe burst testing of Kevlar COPVs that has been completed as a part of an the effort to evaluate the effects of ageing and shelf life on full scale COPVs. The test articles evaluated in this discussion had a diameter of 22 inches for S/N 014 and 40 inches for S/N 011. The

  12. Seismic analysis of a containment vessel

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  13. Device for the simultaneous operation of the closing valve of a vessel and the closing valve of a transport container

    International Nuclear Information System (INIS)

    Tellier, Claude; Surriray, Michel.

    1982-01-01

    This device includes mechanisms for unlatching the closing valve of the vessel and securing it to the closing valve of the transport container and other mechanisms for vertically raising the assembly of valves, pivoting it and bringing it into a vertical position in a bulge provided in the bottom of the transport container. For example the first containment is a nuclear reactor vessel and the transport container is used for carrying an item from the vessel to an external area (for instance, a defective pump to the repair area) and for the return transport operation [fr

  14. EDS V25 containment vessel explosive qualification test report.

    Energy Technology Data Exchange (ETDEWEB)

    Rudolphi, John Joseph

    2012-04-01

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

  15. The effect of thermal loads on buckling strength of cylindrical shells

    International Nuclear Information System (INIS)

    Kawamoto, Y.; Kodama, T.; Matsuura, S.

    1993-01-01

    Nuclear power plant components must be designed taking account of strong seismic loads in countries with frequent earthquakes like Japan. When designing such thin-walled shell components as a main vessel of a fast breeder reactor (FBR), one should consider the possibility that buckling might occur. In Japan, a series of buckling research has been conducted under contract with the Ministry of International Trade and Industry to develop the aseismic design method for a demonstration FBR. This study has been also done as a part of them. The problem of thermal loads on buckling strength is one of the important problems in the buckling research for FBR because axial temperature gradient is produced in a main vessel and the significant thermal stress is shown. Some studies on the effect of thermal loads on buckling strength were carried out (Brochard, 1987), (Nakamura, 1987), but its effect in the actual vessel has not been evaluated quantitatively. We have already reported the effect of thermal loads on buckling strength of a pool-type reactor vessel. (Kawamoto ,1989) In this paper, we focus on a loop-type reactor vessel and investigate the effect of thermal loads accompanying with axial temperature change near the sodium level. And the reduction of buckling strength due to the thermal loads is quantitatively evaluated

  16. A Bankruptcy Problem Approach to Load-shedding in Multiagent-based Microgrid Operation

    OpenAIRE

    Kim, Hak-Man; Kinoshita, Tetsuo; Lim, Yujin; Kim, Tai-Hoon

    2010-01-01

    A microgrid is composed of distributed power generation systems (DGs), distributed energy storage devices (DSs), and loads. To maintain a specific frequency in the islanded mode as an important requirement,  the control of DGs’ output and charge action of DSs are used in supply surplus conditions and load-shedding and discharge action of DSs are used in supply shortage conditions. Recently, multiagent systems for autonomous microgrid operation have been studied. Especially, load-shedding, whi...

  17. Reactor vessel supported by flexure member

    International Nuclear Information System (INIS)

    Crawford, J.D.; Pankow, B.

    1975-01-01

    A description is given of a reactor pressure vessel which is provided with vertical support means in the form of circumferentially spaced columns upon which the vessel is mounted. The columns are adapted to undergo flexure in order to accommodate the thermally induced displacements experienced by the vessel during operational transients

  18. 46 CFR 169.689 - Demand loads.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Demand loads. 169.689 Section 169.689 Shipping COAST... Electrical Electrical Installations on Vessels of 100 Gross Tons and Over § 169.689 Demand loads. Demand loads must meet § 111.60-7 of this chapter except that smaller demand loads for motor feeders are...

  19. Operation and control of a hybrid microgrid containing unbalanced and nonlinear loads

    Energy Technology Data Exchange (ETDEWEB)

    Shahnia, Farhad; Majumder, Ritwik; Ghosh, Arindam; Ledwich, Gerard; Zare, Firuz [School of Engineering, Queensland University of Technology, 2 George Street, Brisbane, Queensland 4001 (Australia)

    2010-08-15

    This paper shows how the power quality can be improved in a microgrid that is supplying a nonlinear and unbalanced load. The microgrid contains a hybrid combination of inertial and converter interfaced distributed generation units where a decentralized power sharing algorithm is used to control its power management. One of the distributed generators in the microgrid is used as a power quality compensator for the unbalanced and harmonic load. The current reference generation for power quality improvement takes into account the active and reactive power to be supplied by the micro-source which is connected to the compensator. Depending on the power requirement of the nonlinear load, the proposed control scheme can change modes of operation without any external communication interfaces. The compensator can operate in two modes depending on the entire power demand of the unbalanced nonlinear load. The proposed control scheme can even compensate system unbalance caused by the single-phase micro-sources and load changes. The efficacy of the proposed power quality improvement control and method in such a microgrid is validated through extensive simulation studies using PSCAD/EMTDC software with detailed dynamic models of the micro-sources and power electronic converters. (author)

  20. Development of advanced design features for KNGR reactor vessel and internals

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-01

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

  1. Development of advanced design features for KNGR reactor vessel and internals

    International Nuclear Information System (INIS)

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

    1995-12-01

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

  2. Coupled thermo-mechanical analysis of corium-loaded lower head of pressure vessel

    International Nuclear Information System (INIS)

    Mishra, J.; Balasubramaniyan, V.

    2016-01-01

    A severe accident in the pressurised water reactor may lead to the relocation of core materials to the lower head of Reactor Pressure Vessel (RPV). The core debris at the bottom of RPV forms a melt pool of corium due to decay heat. The understanding of behaviour of pressure vessel, characterised by failure mode and time to failure, in this scenario is one of the important steps in predicting the accident progression. The most predominant failure mode is multi-axial creep deformation of the vessel with a non-uniform temperature field. Towards this, a numerical analysis methodology is developed for the prediction of pressure vessel deformation during the severe accidents. The methodology involves 2-D finite element modelling under multi-physics environment, which account the creep phenomena using Norton-Bailey creep law with a typical damage model of RPV material. The validation of the methodology is carried out using the results from OLHF experiment carried out in Sandia National Laboratory (SNL), USA, within the framework of an OECD. (author)

  3. TPE upgrade for enhancing operational safety and improving in-vessel tritium inventory assessment in fusion nuclear environment

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, M., E-mail: Masashi.Shimada@inl.gov [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Taylor, C.N.; Moore-McAteer, L.; Pawelko, R.J. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Kolasinski, R.D.; Buchenauer, D.A. [Sandia National Laboratories, Hydrogen and Materials Science Department, Livermore, CA 94550 (United States); Cadwallader, L.C.; Merrill, B.J. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States)

    2016-11-01

    The Tritium Plasma Experiment (TPE) is a unique high-flux linear plasma device that can handle beryllium, tritium, and neutron-irradiated plasma facing materials, and is the only existing device dedicated to evaluate in-vessel tritium inventory in the nuclear environment for fusion safety. The electrical upgrade were recently carried out to enhance operational safety and to improve plasma performance. New DC power supplies and a new control center enable remote plasma operations from outside of the contamination area for tritium, minimizing the possible exposure risk with tritium and beryllium and eliminating heat stress issue. In November 2015, the TPE successfully achieved first deuterium plasma via remote operation after a significant three-year upgrade. Simple linear scaling estimate showed that the TPE is expected to achieve Γ{sub i}{sup max} of >1.0 × 10{sup 23} m{sup −2} s{sup −1} and q{sub heat} of >1 MW m{sup −2} with new power supplies. This upgrade not only improves operational safety of the worker, but also enhances plasma performance to better simulate extreme plasma-material conditions expected in ITER, FNSF, and DEMO for improving in-vessel tritium inventory assessment in fusion nuclear environment.

  4. Multiple shell pressure vessel

    International Nuclear Information System (INIS)

    Wedellsborg, B.W.

    1988-01-01

    A method is described of fabricating a pressure vessel comprising the steps of: attaching a first inner pressure vessel having means defining inlet and outlet openings to a top flange, placing a second inner pressure vessel, having means defining inlet and outlet opening, concentric with and spaced about the first inner pressure vessel and attaching the second inner pressure vessel to the top flange, placing an outer pressure vessel, having inlet and outlet openings, concentric with and spaced apart about the second inner pressure vessel and attaching the outer pressure vessel to the top flange, attaching a generally cylindrical inner inlet conduit and a generally cylindrical inner outlet conduit respectively to the inlet and outlet openings in the first inner pressure vessel, attaching a generally cylindrical outer inlet conduit and a generally cylindrical outer outlet conduit respectively to the inlet and outlet opening in the second inner pressure vessel, heating the assembled pressure vessel to a temperature above the melting point of a material selected from the group, lead, tin, antimony, bismuth, potassium, sodium, boron and mixtures thereof, filling the space between the first inner pressure vessel and the second inner pressure vessel with material selected from the group, filling the space between the second inner pressure vessel and the outer pressure vessel with material selected from the group, and pressurizing the material filling the spaces between the pressure vessels to a predetermined pressure, the step comprising: pressurizing the spaces to a pressure whereby the wall of the first inner pressure vessel is maintained in compression during steady state operation of the pressure vessel

  5. Stress and Thermal Analysis of the In-Vessel RMP Coils in HL-2M

    International Nuclear Information System (INIS)

    Cen Yishun; Li Qiang; Cai Lijun; Jiang Jiaming; Li Guangsheng; Liu Yi; Ding Yonghua

    2013-01-01

    A set of in-vessel resonant magnetic perturbation (RMP) coils for MHD instability suppression is proposed for the design of a HL-2M tokamak. Each coil is to be fed with a current of up to 5 kA, operated in a frequency range from DC to about 1 kHz. Stainless steel (SS) jacketed mineral insulated cables are proposed for the conductor of the coils. In-vessel coils must withstand large electromagnetic (EM) and thermal loads. The support, insulation and vacuum sealing in a very limited space are crucial issues for engineering design. Hence finite element calculations are performed to verify the design, optimize the support by minimizing stress caused by EM forces on the coil conductors and work out the temperature rise occurring on the coil in different working conditions, the corresponding thermal stress caused by the thermal expansion of materials is evaluated to be allowable. The techniques to develop the in-vessel RMP coils, such as support, insulation and cooling, are discussed

  6. A study on ex-vessel steam explosion for a flooded reactor cavity of reactor scale - 15216

    International Nuclear Information System (INIS)

    Song, S.; Yoon, E.; Kim, Y.; Cho, Y.

    2015-01-01

    A steam explosion can occur when a molten corium is mixed with a coolant, more volatile liquid. In severe accidents, corium can come into contact with coolant either when it flows to the bottom of the reactor vessel and encounters the reactor coolant, or when it breaches the reactor vessel and flows into the reactor containment. A steam explosion could then threaten the containment structures, such as the reactor vessel or the concrete walls/penetrations of the containment building. This study is to understand the shortcomings of the existing analysis code (TEXAS-V) and to estimate the steam explosion loads on reactor scale and assess the effect of variables, then we compared results and physical phenomena. Sensitivity study of major parameters for initial condition is performed. Variables related to melt corium such as corium temperature, falling velocity and diameter of melt are more important to the ex-vessel steam explosion load and the steam explosion loads are proportional to these variables related to melt corium. Coolant temperature on reactor cavity has a specific area to increase the steam explosion loads. These results will be used to evaluate the steam explosion loads using ROAAM (Risk Oriented Accident Analysis Methodology) and to develop the evaluation methodology of ex-vessel steam explosion. (authors)

  7. The Assembly and Test of Pressure Vessel for Irradiation

    International Nuclear Information System (INIS)

    Park, Kook Nam; Lee, Jong Min; Youn, Young Jung; June, Hyung Kil; Ahn, Sung Ho; Lee, Kee Hong; Kim, Young Ki; Kennedy, Timothy C.

    2009-01-01

    The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts: the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature

  8. The Assembly and Test of Pressure Vessel for Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kook Nam; Lee, Jong Min; Youn, Young Jung; June, Hyung Kil; Ahn, Sung Ho; Lee, Kee Hong; Kim, Young Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kennedy, Timothy C. [Oregon State University, Corvallis (United States)

    2009-02-15

    The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts: the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature.

  9. Approaches for accounting and prediction of fast neutron fluence on WWER pressure vessels and results of validation of calculational procedure

    International Nuclear Information System (INIS)

    Borodkin, P.G.; Khrennikov, N.N.; Ryabinin, Yu.A.; Adeev, V.A.

    2015-01-01

    A description is given of the universal procedure for calculation of fast neutron fluence (FNF) on WWER vessels. Approbation of the calculation procedure was carried out by comparing the calculation results for this procedure and measurements on the outer surface of the WWER-440 and WWER-1000 vessels. In addition, an estimation of the uncertainty of the settlement procedure was made in accordance with the requirements of regulatory documents. The developed procedure is applied at Kola NPP for independent fast neutron fluence estimates on the WWER-440 reactor vessels when planning core loads taking into account the introduction of new fuels. The results of the pilot operation of the procedure for calculating FNF at the Kola NPP were taken into account when improving the procedure and its application to the calculations of FNF on the WWER-1000 vessels [ru

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

    International Nuclear Information System (INIS)

    Baum, M.R.

    1989-04-01

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

  11. The development of an in-vessel cryopump system for the DIII-D tokamak

    International Nuclear Information System (INIS)

    Schaubel, K.M.; Baxi, C.B.; Campbell, G.L.; Laughon, G.J.; Mahdavi, M.A.; Makariou, C.C.; Smith, J.P.; Schaffer, M.J.; Menon, M.M.

    1993-07-01

    The design, testing and initial operation of the DIII-D advanced divertor cryocondensation pumping system is presented. The pump resides inside the tokamak plasma containment vessel where it provides particle exhaust pumping, and it is subjected to Joule heating and hot particle heat loads during each 10 second discharge. In addition, the pump must withstand plasma disruption induced electromagnetic forces and 400 degrees C bake-out temperatures. Cooling is accomplished by forced flow liquid helium with the two-phase helium exhaust passing through a reliquefier for thermal efficiency. A prototype pump was constructed to study surface temperature rise as a function of flow geometry, applied heat load, helium mass flow rate, and pump outlet conditions. Prototype testing led to the development of a special geometry which was demonstrated to enhance two-phase flow stability and overall heat transfer. During initial operation, deuterium pumping speeds of 32,000 L/s at 2 mTorr pressure were achieved with a helium flow rate of 5 g/s. This speed was maintained during 300 W, 8 s long test beat pulses which meets operational goals

  12. Dynamic fracture characterization of a pressure vessel steel

    International Nuclear Information System (INIS)

    Schmitt, W.; Boehme, W.; Klemm, W.; Memhard, D.; Winkler, S.

    1991-01-01

    Dynamic events are characterized by time and space-dependent stress and strain fields caused by wave or inertia effect. The dynamic effect at cracks may be originated from the rapid loading rate or impact loading of a structure containing a stationary crack or the time-dependent stress and strain fields of a propagating or arresting crack itself. Dynamic effects complicate the analysis of crack tip stress and strain fields, and usually considerable experimental effort and numerical technique are required. High loading rate influences the deformation and yield behavior and also the fracture toughness of materials. In order to know the propagation and arrest behavior of cracks, a heat of a German reactor pressure vessel steel was investigated, and the dynamic J-resistance curves were evaluated with large three-point bending specimens by impact loading, moreover, the crack propagation energy at large crack extension was determined with wide tension plates. The material tested was a ferritic pressure vessel steel, ASTM A 508 Cl 2. The dynamic J-resistance curves and numerical simulation and fractographic examination, and crack propagation energy are reported. (K.I.)

  13. ITER in-vessel system design and performance

    International Nuclear Information System (INIS)

    Parker, R.R.

    1999-01-01

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

  14. ITER in-vessel system design and performance

    International Nuclear Information System (INIS)

    Parker, R.R.

    2001-01-01

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

  15. Stress categorization in nozzle to pressure vessel connections finite elements models

    International Nuclear Information System (INIS)

    Albuquerque, Levi Barcelos de

    1999-01-01

    The ASME Boiler and Pressure Vessel Code, Section III , is the most important code for nuclear pressure vessels design. Its design criteria were developed to preclude the various pressure vessel failure modes throughout the so-called 'Design by Analysis', some of them by imposing stress limits. Thus, failure modes such as plastic collapse, excessive plastic deformation and incremental plastic deformation under cyclic loading (ratchetting) may be avoided by limiting the so-called primary and secondary stresses. At the time 'Design by Analysis' was developed (early 60's) the main tool for pressure vessel design was the shell discontinuity analysis, in which the results were given in membrane and bending stress distributions along shell sections. From that time, the Finite Element Method (FEM) has had a growing use in pressure vessels design. In this case, the stress results are neither normally separated in membrane and bending stress nor classified in primary and secondary stresses. This process of stress separation and classification in Finite Element (FE) results is what is called stress categorization. In order to perform the stress categorization to check results from FE models against the ASME Code stress limits, mainly from 3D solid FE models, several research works have been conducted. This work is included in this effort. First, a description of the ASME Code design criteria is presented. After that, a brief description of how the FEM can be used in pressure vessel design is showed. Several studies found in the literature on stress categorization for pressure vessel FE models are reviewed and commented. Then, the analyses done in this work are presented in which some typical nozzle to pressure vessel connections subjected to internal pressure and concentrated loads were modeled with solid finite elements. The results from linear elastic and limit load analyses are compared to each other and also with the results obtained by formulae for simple shell

  16. Load following operation of nuclear power plants for meeting power system requirements

    International Nuclear Information System (INIS)

    Isoda, Hachiro

    1987-01-01

    This paper describes a calculating program on the availability factors of nuclear, thermal and pumed storage hydro power stations and some calculated results for typical three load factors, 55 %, 60 % and 71 %, are provided when the share of the nuclea power station in the generation facilities is increased. The load following requirement of the nuclear power station is also provided. Load following requirement: If there is a 10 % pumped storage hydro power station, the nuclear power station enables to be operated with its rated output up to 30 % - 35 % of its share. Its daily load following operation for 40 % and 50 % nuclear power station needs every weekend and every day respectively. Availability factor: The availability factor of the nuclear power station manages to get 80 % (maximum availability factor of the nuclear power station in this study) up to 30 % share of it with 10 % pumpued storage hydro power station. When the nuclear power station shares 40 % and 50 %, its availability factor decreases down 1 % and 5 % respectively. (author)

  17. Lubricant coating of dowel for the ITER vacuum vessel gravity support

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B.Y. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Ahn, H.J., E-mail: hjahn@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Bak, J.S. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Choi, C.H.; Ioki, K. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Zauner, C. [KRP-Mechatec Engineering GbR, 85748 Garching b, Muenchen (Germany)

    2012-08-15

    The ITER vacuum vessel gravity supports located in the lower level shall sustain loads in radial, toroidal and vertical directions. The hinge type VVGS consists of two hinges, upper and lower blocks and dowels. In order to develop the design concept and verify the structural integrity of the hinge system, the design analysis has been performed in detail. Inclination of 15 Degree-Sign for the hinge based supporting system was introduced to provide centering force to make stable equilibrium state of the vacuum vessel. Due to this inclination the hinges are rotated by the radial expansion of the VV during operation and baking, respectively. If a dowel is seized in the hinge, the supporting system can be highly stressed due to the restrained displacement in the seized dowel. Therefore, solid lubricant coatings were suggested on dowels in order to avoid seizing in the sliding area. In this work, several sets of coupons were made with different coating materials to investigate the effect according to the selection of coating material. Also, a test facility was designed to cover the ITER relevant loading and boundary conditions, e.g. vacuum condition, temperature, contact pressure, cycles, etc. From those test results, the optimized coating method was found to avoid seizure of dowel in the ITER VVGS.

  18. Lower bounds of collapse loads in axisymmetrical vessels

    International Nuclear Information System (INIS)

    Fonseca Neto, J. de D.; Ebecken, N.F.F.

    1981-01-01

    The rigid-plastic limit analysis of shells of revolution subject to rotationally symmetric loadings, is presented. After assembling the finite elements, the limit analysis program is reduced to a simple application of the non-linear programming technique, where the sequential unconstrained minimization technique (SUMT) is utilized for the statically admissible approach. Lower bounds of the collapse loads are presented and compared with the results described in the literature. (Author) [pt

  19. Forging technology for large nuclear pressure vessel parts

    International Nuclear Information System (INIS)

    Kakimoto, Hideki; Ikegami, Tomonori

    2014-01-01

    The increasing output of nuclear power generation calls for larger vessels for next-generation nuclear power plants. A vessel with an increased diameter requires increased load for its forging, which can make it difficult to use a conventional solid die. In order to reduce the forging load, a rotary incremental forging method has been applied to hot forging. This method includes pressing and rotating a material in an incremental manner such that a target shape is obtained. This study aimed at improving the accuracy of numerical simulation for the rotary incremental forging to reduce the load when forging large vessels. This has enabled the temperature of the material and flow stress to be precisely predicted; an example of this is reported in the paper. Specifically, the heat transfer coefficient to be used for the numerical simulation had been determined experimentally from a small-scale hot-forging. The reduction of the flow stress associated with incremental forging, had been deduced from a compression test, and the value was applied to the numerical simulation. A preform was designed on the basis of the above simulation to perform a 1/1 size scale experiment. A precision of better than 5% has been confirmed for the shape prediction. (author)

  20. Design and performance of a 2-megawatt high voltage dc test load

    International Nuclear Information System (INIS)

    Horan, D.; Kustom, R.; Ferguson, M.

    1994-01-01

    A high-power water-cooled resistive load which simulates the electrical load characteristics of a high-power klystron, capable of a 2 MW dissipation at 95 kV DC, is designed and installed. The load utilizes wirewound resistor elements suspended inside G-11 insulated tubing contained within a single-wall 316 stainless steel pressure vessel with flanged elliptical heads. The vessel supplies a continuous flow of deionized water. Baffles fabricated from G-10 sheets support the tubing and promote water turbulence to maximize heat removal. A companion oil tank houses resistive filament and mod-anode power supply test loads, plus an electrical interlock system which provides protection from inadequate water flow, excessive oil temperature, and arcing in either the pressure vessel or oil tank. A secondary safety system consists of both hydrostatic and steam pressure relief valves on the pressure vessel. Power supply tests indicate the load simulates the electrical load characteristics of a high-power klystron to a degree sufficient to accurately performance-test the rf high voltage power supplies used at the Advanced Photon Source

  1. Pressure vessel design manual

    CERN Document Server

    Moss, Dennis R

    2013-01-01

    Pressure vessels are closed containers designed to hold gases or liquids at a pressure substantially different from the ambient pressure. They have a variety of applications in industry, including in oil refineries, nuclear reactors, vehicle airbrake reservoirs, and more. The pressure differential with such vessels is dangerous, and due to the risk of accident and fatality around their use, the design, manufacture, operation and inspection of pressure vessels is regulated by engineering authorities and guided by legal codes and standards. Pressure Vessel Design Manual is a solutions-focused guide to the many problems and technical challenges involved in the design of pressure vessels to match stringent standards and codes. It brings together otherwise scattered information and explanations into one easy-to-use resource to minimize research and take readers from problem to solution in the most direct manner possible. * Covers almost all problems that a working pressure vessel designer can expect to face, with ...

  2. LANL Robotic Vessel Scanning

    Energy Technology Data Exchange (ETDEWEB)

    Webber, Nels W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-25

    Los Alamos National Laboratory in J-1 DARHT Operations Group uses 6ft spherical vessels to contain hazardous materials produced in a hydrodynamic experiment. These contaminated vessels must be analyzed by means of a worker entering the vessel to locate, measure, and document every penetration mark on the vessel. If the worker can be replaced by a highly automated robotic system with a high precision scanner, it will eliminate the risks to the worker and provide management with an accurate 3D model of the vessel presenting the existing damage with the flexibility to manipulate the model for better and more in-depth assessment.The project was successful in meeting the primary goal of installing an automated system which scanned a 6ft vessel with an elapsed time of 45 minutes. This robotic system reduces the total time for the original scope of work by 75 minutes and results in excellent data accumulation and transmission to the 3D model imaging program.

  3. An effective surveillance strategy for reactor pressure vessel assessment in the long term operation perspective

    International Nuclear Information System (INIS)

    Chaouadi, R.; Gerard, R.

    2015-01-01

    The reactor pressure vessel (RPV) irradiation embrittlement is monitored by means of surveillance capsules containing the RPV belt-line materials, inserted inside the reactor pressure vessel (RPV) before the start of operation. These capsules are placed at location where they receive a higher neutron flux than the vessel wall, by a factor of the order of 2 to 3. They are regularly retrieved and tested to evaluate the RPV irradiation embrittlement according to specific regulatory procedures and standards, in order to guarantee the safe operation of the RPV throughout its lifetime. These procedures are often relying on empirical but conservative concepts. In parallel, material research reactor (MTR) irradiations are often used to support the surveillance data and to develop a better understanding of irradiation effects, not only qualitatively but also quantitatively. Taking advantage of the increased understanding of irradiation effects, analytical tools were developed to improve the evaluation embrittlement and quality assurance of the RPV embrittlement assessment. In this framework, an alternative but complementary surveillance program assessment was developed in Belgium, the so-called enhanced surveillance, in order to benefit from the latest developments in the area of materials science and irradiation effects. The neutron flux and fracture properties of the surveillance materials can be reliably characterized and correlated to each other using physically-based rather than empirical concepts. The enhanced surveillance approach is complementary to the mandatory regulatory procedure and allows quantifying the conservatism of the regulatory approach. The enhanced surveillance approach that uses the reconstitution technology to fabricate additional small size specimens, appropriate modeling tools and microstructural examination when required, makes it possible to rationalize all available information in a physically-based way

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

    International Nuclear Information System (INIS)

    Abdulrahman, H.O.

    1978-01-01

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

  5. Evaluation of the stress distribution on the pressure vessel head with multi-openings

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K.S.; Kim, T.W.; Jeong, K.H.; Lee, G.M. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    This report discusses and analyzes the stress distribution on the pressure vessel head with multi-openings(3 PSV nozzles, 2 SDS nozzles and 1 Man Way) according to patterns of the opening distance. The pressurizer of Korea Standardized Nuclear Power Plant(Ulchin 3 and 4), which meets requirements of the cyclic operation and opening design defined by ASME code, was used as the basic model for that. Stress changes according to the distance between openings were investigated and the factors which should be considered for the opening design were analyzed. Also, the nozzle loads at Level A, B conditions and internal pressure were applied in order to evaluate changes of head stress distributions due to nozzle loads. (author). 6 refs., 29 figs., 4 tabs.

  6. ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R AND P REACTOR VESSELS

    Energy Technology Data Exchange (ETDEWEB)

    Wiersma, B.

    2010-05-24

    The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D and D). D and D activities consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS and T) of the Savannah River National Laboratory (SRNL) has been requested to review and evaluate existing experimental and analytical studies of this issue to determine if any process constraints on the chemistry of the fill material and the fill operation are necessary. Various options exist for the type of grout material that may be used for D and D of the reactor vessels. The grout formulation options include ceramicrete (pH 6-8), low pH portland cement + silica fume grout (pH 10.4), or Portland cement groupt (pH 12.5). The assessment concluded that either ceramicrete or the silica fume grout may be used to safely grout the P-reactor vessel. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Portland cement grout, on the other hand, for the same range of process parameters does not provide a margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations in the P-reactor vessel. It is recommended that this grout not be utilized for this task. The R-reactor vessel cotnains significantly less aluminum based on current facility process knowledge, surface observations, and drawings. Therefore, a Portland cement grout may be considered for grouting operations as well as the other grout formulations. For example, if the grout fill rate is less than 1 inch/min and the grout temperature is maintained at 70 C or less, the risk of hydrogen accumulation during fill

  7. Expanded Fermilab pressure vessel directory program

    Energy Technology Data Exchange (ETDEWEB)

    Tanner, A.

    1983-01-01

    Several procedures have been written to manage the information pertaining to the vacuum tanks and pressure vessels for which the laboratory is responsible. These procedures have been named TANK1 for the vessels belonging to the Accelerator Division, TANK2 and TANK3 for the vessels belonging to the Research Division and to Technical Support respectively, and TANK4 for the vessels belonging to the Business Division. The operating procedures are otherwise identical in every respect.

  8. Expanded Fermilab pressure vessel directory program

    International Nuclear Information System (INIS)

    Tanner, A.

    1983-01-01

    Several procedures have been written to manage the information pertaining to the vacuum tanks and pressure vessels for which the laboratory is responsible. These procedures have been named TANK1 for the vessels belonging to the Accelerator Division, TANK2 and TANK3 for the vessels belonging to the Research Division and to Technical Support respectively, and TANK4 for the vessels belonging to the Business Division. The operating procedures are otherwise identical in every respect

  9. An experimental study on in-vessel debris retention through gap cooling

    International Nuclear Information System (INIS)

    Kang, Kyung Ho; Kim, Jong Whan; Cho, Young Ro; Chang, Young Cho; Park, Rae Jun; Gu, Kil Mo; Kim, Sang Baik; Kim, Hee Dong

    1999-04-01

    LAVA experiments have been performed using high temperature molten material to be relocated into the 1/8 linear scaled vessel of a reactor lower plenum filled with water. An Al 2 O 3 /Fe tehrmite melt (Al 2 O 3 only) was used as a corium simulant. In this study, the influence of various initial conditions, such as internal pressure load across the vessel wall, the material composition of the melt simulant, water subcooling and depth on gap formation were investigated. As well, the thermal and mechanical behaviors of the vessel were examined. In case the internal pressure load was imposed, the gap formation between the continuous solidified debris and the vessel wall was clearly shown with post-test examination. On the other hand, in case the internal pressure load was not imposed, the iron welded to the inner surface of the vessel and the vessel experienced ablation to about 5 mm. The cooling rate of the vessel was very slow in the tests using Al 2 O 3 /Fe thermite melt but it was rather fast using Al 2 O 3 melt. It is postulated that in the Al 2 O 3 /Fe thermite tests, the iron melt layer is so dense that water ingression into the gap is difficult due to the high pressure of escaping steam. On the other hand, in the porosity of an Al 2 O 3 melt layer could enhance water ingression into the gap by giving the flow path of the evaporated steam through the porous media. The water height and subcooling could affect the melt pool formation and the initial thermal attack to the vessel. However, at present stage, the effect of water subcooling on the thermal behavior of the vessel couldn't be generalized. For clear confirmation of the effect of water subcooling, the tests will be performed at the saturated water condition. (author). 19 refs., 3 tabs., 36 figs

  10. Development of PWR pressure vessel steels

    International Nuclear Information System (INIS)

    Druce, S.; Edwards, B.

    1982-01-01

    Requirements to be met by vessel steels for pressurized water reactors are analyzed. Chemicat composition of low-alloyed steels, mechanical properties of sheets and forgings made of these steels and changes in the composition and properties over the wall thickness of the reactor vessel are presented. Problems of the vessel manufacturing including welding and heat treatment processes of sheets and forgings are considered. Special attention is paid to steel embrittlement during vessel fabrication and operation (radiation embrittlement, thermal embrittlement). The role of non-metal inclusions and their effect on anisotropy of fracture toughness is discussed. Possible developments of vessel steels and procedures for producing reactor vessels are reviewed

  11. Development of PWR pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Druce, S.; Edwards, B.

    1982-01-01

    Requirements to be met by vessel steels for pressurized water reactors are analyzed. Chemicat composition of low-alloyed steels, mechanical properties of sheets and forgings made of these steels and changes in the composition and properties over the wall thickness of the reactor vessel are presented. Problems of the vessel manufacturing including welding and heat treatment processes of sheets and forgings are considered. Special attention is paid to steel embrittlement during vessel fabrication and operation (radiation embrittlement, thermal embrittlement). The role of non-metal inclusions and their effect on anisotropy of fracture toughness is discussed. Possible developments of vessel steels and procedures for producing reactor vessels are reviewed.

  12. Study on operation conditions and an operation system of a nuclear powered submersible research vessel, 'report of working group on application of a very small nuclear reactor to an ocean research'

    Energy Technology Data Exchange (ETDEWEB)

    Ura, Tamaki [Tokyo Univ., Tokyo (Japan); Takamasa, Tomoji [Tokyo Univ. of Mercantile Marine, Tokyo (Japan); Nishimura, Hajime [Japan Marine Science and Technology Center, Yokosuka, Kanagawa (JP)] [and others

    2001-07-01

    JAERI has studied on design of a nuclear powered submersible research vessel, which will navigate under sea mainly in the Arctic Ocean, as a part of the design activity of advanced marine reactors. This report describes operation conditions and an operating system of the vessel, which were discussed by the specialists of hull design, sound positioning, ship motions and oceanography, etc. The design conditions on ship motions for submersible vessels were surveyed considering regulations in our country, and ship motions were evaluated in the cases of underwater and surface navigations taking account of observation activities in the Arctic Ocean. The effect of ship motions on the compact nuclear reactor SCR was assessed. A submarine transponder system and an on-ice communication buoy system were examined as a positioning and communication system, supposing the activity under ice. The interval between transponders or communication buoys was recommended as 130 km. Procedures to secure safety of nuclear powered submersible research vessel were discussed according to accidents on the hull or the nuclear reactor. These results were reflected to the concept of the nuclear powered submersible research vessel, and subjects to be settled in the next step were clarified. (author)

  13. Seismic-load-induced human errors and countermeasures using computer graphics in plant-operator communication

    International Nuclear Information System (INIS)

    Hara, Fumio

    1988-01-01

    This paper remarks the importance of seismic load-induced human errors in plant operation by delineating the characteristics of the task performance of human beings under seismic loads. It focuses on man-machine communication via multidimensional data like that conventionally displayed on large panels in a plant control room. It demonstrates a countermeasure to human errors using a computer graphics technique that conveys the global state of the plant operation to operators through cartoon-like, colored graphs in the form of faces that, with different facial expressions, show the plant safety status. (orig.)

  14. Demonstration of an automated electromanometer for measurement of solution volume in accountability vessels

    International Nuclear Information System (INIS)

    Suda, S.; Keisch, B.; Hayashi, M.; Onuma, T.; Fukuari, Y.

    1981-09-01

    A system for measuring the liquid volume in input and plutonium product accountability vessels, based upon a desktop-computer-controlled electromanometer, was installed at the Tokai-Mura reprocessing plant. In-tank temperatures, pressure measurements relating to volume and density, and load-cell weights are measured cyclically and recorded. The system feasibility was demonstrated through a series of tests including vessel calibration and the effects of thermal expansion, and through use during thirteen months of on-line plant operation. The value to the operator of the recording, display, replay, data handling, and report generation features of the system was demonstrated as was the enhanced precision of the electromanometer as compared to the conventional water-filled manometer system. The automated electromanometer system consists of a pneumatic scanner, a precision electromanometer, electronic scanner, a digital voltmeter, and a desktop computer with disc and tape mass storage, cathode-ray tube (CRT) graphics display, and printer output. The desktop computer is used to control the pneumatic and electronic scanners and the digital voltmeter and to log in the measurement data

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

    International Nuclear Information System (INIS)

    Ishiwatari, Koji; Otsuki, Akira

    1998-01-01

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

  16. Evaluation of reactivity and Xe behavior during daily load following operation

    International Nuclear Information System (INIS)

    Sakamoto, Yasunori; Araki, Tsuneyasu; Yamamoto, Fumiaki

    1992-01-01

    A boiling water reactor (BWR) has an excellent load following capability provided by a core flow control, which is used for changing a reactor power level and for compensating the subsequent Xe concentration change. The core characteristics during load following operations are investigated in detail, using our reactor core simulator. Comparisons of changes of the Doppler reactivity, the void reactivity and the Xe reactivity during transients are performed. Also the features of Xe transient during load following operations are shown. It has been shown that the core flow change required to compensate the Xe reactivity change produces much greater change of the void reactivity than that required for power level changes, and that the resulting local power change in the lower part of the core is greater than that in the upper part, because the Xe concentration change in the lower part is hardly compensated by the core flow control. Also the effects of power level changes, cycle patterns, and initial concentration of Xe and I on the Xe transient behavior have been investigated. (author)

  17. Pressurized wet digestion in open vessels (T11)

    International Nuclear Information System (INIS)

    Kettisch, P.; Maichin, P.; Zischka, M.; Knapp, G.

    2002-01-01

    Full text: Pressurized wet digestion in closed vessels, microwave assisted or with conventional conductive heating, is the most important sample preparation technique for digestion or leaching procedures in element analysis. In comparison to open vessel digestion closed vessel digestion methods have many advantages, but there is one disadvantage - complex and expensive vessel designs. A new technique - pressurized wet digestion in open vessels - combine the advantages of closed vessel sample digestion with the application of simple and cheap open vessels made of quartz or PFA. The vessels are placed in a high pressure Asher HPA, which is adapted with a Teflon liner and filled partly with water. The analytical results with 30 ml quartz vessels, 22 ml PFA vessels and 1.5 ml PIA auto sampler cups will be shown. In principle every dimensions of vessels can be used. The vessels are loaded with sample material (max. 1.5 g with quartz vessels, max. 0.5 g with PFA vessels and 50 mg with auto sampler cups) and digestion reagent. Afterwards the vessels are simply covered with PTFE stoppers and not sealed. The vessels are transferred into a special adapted HPA and digested at temperatures up to 270 o C. The digestion time is 90 min. and cooling down to room temperature 30 min. The analytical results of CRM's are within the certified values and no cross contamination and losses of volatile elements could be observed. (author)

  18. Assessment of Workers' Exposure to Grain Dust and Bioaerosols During the Loading of Vessels' Hold: An Example at a Port in the Province of Québec.

    Science.gov (United States)

    Marchand, Geneviève; Gardette, Marie; Nguyen, Kiet; Amano, Valérie; Neesham-Grenon, Eve; Debia, Maximilien

    2017-08-01

    Longshoremen are exposed to large amounts of grain dust while loading of grain into the holds of vessels. Grain dust inhalation has been linked to respiratory diseases such as chronic bronchitis, hypersensitivity, pneumonitis, and toxic pneumonitis. Our objective was to characterize the exposure of longshoremen to inhalable and total dust, endotoxins, and cultivable bacteria and fungi during the loading of grain in a vessel's hold at the Port of Montreal in order to assess the potential health risks. Sampling campaigns were conducted during the loading of two different types of grain (wheat and corn). Environmental samples of microorganisms (bacteria, fungus, and actinomycetes) were taken near the top opening of the ship's holds while personal breathing zone measurements of dust and endotoxins were sampled during the worker's 5-hour shifts. Our study show that all measurements are above the recommendations with concentration going up to 390 mg m-3 of total dust, 89 mg m-3 of inhalable fraction, 550 000 EU m-3 of endotoxins, 20 000 CFU m-3 of bacteria, 61 000 CFU m-3 of fungus and 2500 CFU m-3 of actinomycetes. In conclusion, longshoremen are exposed to very high levels of dust and of microorganisms and their components during grain loading work. Protective equipment needs to be enforced for all workers during such tasks in order to reduce their exposure. © The Author 2017. Published by Oxford University Press on behalf of the British Occupational Hygiene Society.

  19. Failure probability analysis on mercury target vessel

    International Nuclear Information System (INIS)

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

    2005-03-01

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

  20. Stress analysis of hydride bed vessels used for tritium storage

    International Nuclear Information System (INIS)

    McKillip, S.T.; Bannister, C.E.; Clark, E.A.

    1991-01-01

    A prototype hydride storage bed, using LaNi 4.25 Al 0.75 as the storage material, was fitted with strain gages to measure strains occurring in the stainless steel bed vessel caused by expansion of the storage powder upon uptake of hydrogen. The strain remained low in the bed as hydrogen was added, up to a bed loading of about 0.5 hydrogen to metal atom ratio (H/M). The strain then increased with increasing hydrogen loading (∼ 0.8 H/M). Different locations exhibited greatly different levels of maximum strain. In no case was the design stress of the vessel exceeded

  1. Changes in the Strength of the Polymer Concrete Used in the Electroplating Vats Under Operational Load

    Directory of Open Access Journals (Sweden)

    Radna Lidia

    2017-12-01

    Full Text Available Due to the strong and aggressive electrolyte media and thermal load, design of the electroplating vats in the copper industry often relies on the resin concrete. The article presents the results of the strength tests of the polymer concrete based on the "Derakane" resin, used in the construction of electroplating vats. Samples were taken from the real vats - both new and 17-year old. Strength tests included compression and bending tensile strength test. To assess the effect of operational conditions the tests were performed on the same-age vats, some of which were never used while others were subjected to the operational load. During the operation, the vats sustained load of the anode and cathode weights, cyclic electrolyte loading with a temperatures up to 60°C. As a result, it was noted that the operational conditions led to the increased strength of the polymer concrete material.

  2. Changes in the Strength of the Polymer Concrete Used in the Electroplating Vats Under Operational Load

    Science.gov (United States)

    Radna, Lidia; Sakharov, Volodymyr

    2017-12-01

    Due to the strong and aggressive electrolyte media and thermal load, design of the electroplating vats in the copper industry often relies on the resin concrete. The article presents the results of the strength tests of the polymer concrete based on the "Derakane" resin, used in the construction of electroplating vats. Samples were taken from the real vats - both new and 17-year old. Strength tests included compression and bending tensile strength test. To assess the effect of operational conditions the tests were performed on the same-age vats, some of which were never used while others were subjected to the operational load. During the operation, the vats sustained load of the anode and cathode weights, cyclic electrolyte loading with a temperatures up to 60°C. As a result, it was noted that the operational conditions led to the increased strength of the polymer concrete material.

  3. Reactor pressure vessel structural integrity research

    International Nuclear Information System (INIS)

    Pennell, W.E.; Corwin, W.R.

    1994-01-01

    Development continues on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite-element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack tip of shallow surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil ductility temperature (NDT) performs better than the reference temperature for nil ductility transition (RT NDT ) as a normalizing parameter for shallow-flaw fracture toughness data, (3) biaxial loading can reduce the shallow-flaw fracture toughness, (4) stress-based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on shallow-flaw fracture toughness because in-plane stresses at the crack tip are not influenced by biaxial loading, and (5) an implicit strain-based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow-flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation-induced shift in Charpy V-notch vs temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties

  4. Prediction of Composite Pressure Vessel Failure Location using Fiber Bragg Grating Sensors

    Science.gov (United States)

    Kreger, Steven T.; Taylor, F. Tad; Ortyl, Nicholas E.; Grant, Joseph

    2006-01-01

    Ten composite pressure vessels were instrumented with fiber Bragg grating sensors in order to assess the strain levels of the vessel under various loading conditions. This paper and presentation will discuss the testing methodology, the test results, compare the testing results to the analytical model, and present a possible methodology for predicting the failure location and strain level of composite pressure vessels.

  5. Analysis and evaluation system for elevated temperature design of pressure vessels

    International Nuclear Information System (INIS)

    Hayakawa, Teiji; Sayawaki, Masaaki; Nishitani, Masahiro; Mii, Tatsuo; Murasawa, Kanji

    1977-01-01

    In pressure vessel technology, intensive efforts have recently been made to develop the elevated temperature design methods. Much of the impetus of these efforts has been provided mainly by the results of the Liquid Metal Fast Breeder Reactor (LMFBR) and more recently, of the High Temperature Gas-cooled Reactor (HTGR) Programs. The pressure vessels and associated components in these new type nuclear power plants must operate for long periods at elevated temperature where creep effects are significant and then must be designed by rigorous analysis for high reliability and safety. To carry out such an elevated temperature designing, numbers of highly developed analysis and evaluation techniques, which are so complicated as to be impossible by manual work, are indispensable. Under these circumstances, the authors have made the following approaches in the study: (1) Study into basic concepts and the associated techniques in elevated temperature design. (2) Systematization (Analysis System) of the procedure for loads and stress analyses. (3) Development of post-processor, ''POST-1592'', for strength evaluation based on ASME Code Case 1592-7. By linking the POST-1592 together with the Analysis System, an analysis and evaluation system is developed for an elevated temperature design of pressure vessels. Consequently, designing of elevated temperature vessels by detailed analysis and evaluation has easily and effectively become feasible by applying this software system. (auth.)

  6. Marine pollution originating from purse seine and longline fishing vessel operations in the Western and Central Pacific Ocean, 2003-2015.

    Science.gov (United States)

    Richardson, Kelsey; Haynes, David; Talouli, Anthony; Donoghue, Michael

    2017-03-01

    Fisheries observer data recorded between 2003 and 2015 on-board purse seine and longline vessels operating in the Western and Central Pacific Ocean reported more than 10 000 pollution incidents within the exclusive economic zones (EEZs) of 25 Pacific countries and territories, and in international waters. A majority of the reported purse seine pollution incidents related to dumping of plastics waste. Other common pollution incidents related to oil spillages and to abandoned, lost or dumped fishing gear. Data analysis highlighted the need for increased monitoring, reporting, and enforcement of pollution violations by all types of fishing vessels operating in the Pacific region; a regional outreach and compliance assistance programme on marine pollution prevention and improvements in Pacific port waste reception facilities.

  7. Tar loads on Omani beaches

    International Nuclear Information System (INIS)

    Badawy, M.I.; Al-Harthy, F.T.

    1991-01-01

    Owing to Oman's geographic position and long coastal line, the coastal areas of Oman are particularly vulnerable to oil pollution from normal tanker operations, illegal discharges, and accidental spills as well as local sources of oil input. UNEP carried out a survey on the coasts of Oman to determine the major sources of oil pollution and concluded that the major shoreline pollution problems in Oman arose from operational discharges of oil from passing vessels traffic. The oil, because of the high sea and air temperatures in the area, was subjected to relatively high rates of evaporation and photo-oxidation and tended to arrive at the coast as heavy petroleum particulate residues (tar balls). The aim of the present study was to measure the loads of tar balls in Omani coastal areas and to identify the source of oil pollutants on beaches

  8. Switch Panel wear loading - a parametric study regarding governing train operational factors

    Science.gov (United States)

    Hiensch, E. J. M.; Burgelman, N.

    2017-09-01

    The acting forces and resulting material degradation at the running surfaces of wheels and rail are determined by vehicle, track, interface and operational characteristics. To effectively manage the experienced wear, plastic deformation and crack development at wheels and rail, the interaction between vehicle and track demands a system approach both in maintenance and in design. This requires insight into the impact of train operational parameters on rail- and wheel degradation, in particular at switches and crossings due to the complex dynamic behaviour of a railway vehicle at a turnout. A parametric study was carried out by means of vehicle-track simulations within the VAMPIRE® multibody simulation software, performing a sensitivity analysis regarding operational factors and their impact on expected switch panel wear loading. Additionally, theoretical concepts were cross-checked with operational practices by means of a case study in response to a dramatic change in lateral rail wear development at specific switches in Dutch track. Data from train operation, track maintenance and track inspection were analysed, providing further insight into the operational dependencies. From the simulations performed in this study, it was found that switch rail lateral wear loading at the diverging route of a 1:9 type turnout is significantly influenced by the level of wheel-rail friction and to a lesser extent by the direction of travel (facing or trailing). The influence of other investigated parameters, being vehicle speed, traction, gauge widening and track layout is found to be small. Findings from the case study further confirm the simulation outcome. This research clearly demonstrates the contribution flange lubrication can have in preventing abnormal lateral wear at locations where the wheel-rail interface is heavily loaded.

  9. Reactor-vessel-sectioning demonstration

    International Nuclear Information System (INIS)

    Lundgren, R.A.

    1981-09-01

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

  10. Analysis of cracked pressure vessel nozzles by finite elements

    International Nuclear Information System (INIS)

    Reynen, J.

    1975-01-01

    In order to assess the safety of pressure vessel nozzles, the analysis should take into account cracks. The paper describes various algorithms, their computer implementations and relative merits to define in an effective way strain energy release rates along the tip front of arbitrary 3 D cracks under arbitary load including thermal strains. These techniques are basically equivalent to substructuring techniques and consequently they can be implemented to only FEM program able to deal with the data handling problems of the substructuring technique. Examples are given carried out with a substructure version of the BERSAFE system. These examples include a corner crack in a pressure vessel nozzle loaded by internal pressure and by thermal stresses. (Auth.)

  11. Guidelines for pressure vessel safety assessment

    Science.gov (United States)

    Yukawa, S.

    1990-04-01

    A technical overview and information on metallic pressure containment vessels and tanks is given. The intent is to provide Occupational Safety and Health Administration (OSHA) personnel and other persons with information to assist in the evaluation of the safety of operating pressure vessels and low pressure storage tanks. The scope is limited to general industrial application vessels and tanks constructed of carbon or low alloy steels and used at temperatures between -75 and 315 C (-100 and 600 F). Information on design codes, materials, fabrication processes, inspection and testing applicable to the vessels and tanks are presented. The majority of the vessels and tanks are made to the rules and requirements of ASME Code Section VIII or API Standard 620. The causes of deterioration and damage in operation are described and methods and capabilities of detecting serious damage and cracking are discussed. Guidelines and recommendations formulated by various groups to inspect for the damages being found and to mitigate the causes and effects of the problems are presented.

  12. Automatic Control of Reactor Temperature and Power Distribution for a Daily Load following Operation

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Keuk Jong; Kim, Han Gon [Korea Hydro and Nuclear Power Institute, Daejeon (Korea, Republic of)

    2010-10-15

    An automatic control method of reactor power and power distribution was developed for a daily load following operation of APR1400. This method used a model predictive control (MPC) methodology having second-order plant data. And it utilized a reactor power ratio and axial shape index as control variables. However, the reactor regulating system of APR1400 is operated by the difference between the average temperature of the reactor core and the reference temperature, which is proportional to the turbine load. Thus, this paper reports on the model predictive control methodology using fourth-order plant data and a reactor temperature instead of the reactor power shape. The purpose of this study is to develop a revised automatic controller and analyze the behavior of the nuclear reactor temperature (Tavg) and the axial shape index (ASI) using the MPC method during a daily load following operation

  13. Finite element analysis of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  14. Improving the extraction-and-loading process in the open mining operations

    Directory of Open Access Journals (Sweden)

    Cheban A. Yu.

    2017-09-01

    Full Text Available Using the explosions is the main way to prepare solid rocks for the excavation, and that results in the formation of a rock mass of uneven granulometric composition, which makes it impossible to use a conveyor quarry transport without the preliminary large crushing of the rock mass obtained during the explosion. A way to achieve the greatest technical and economic effect is the full conveyorization of quarry transport, what, in this case, ensures the sequenced-flow of transport operations, automation of management and high labor productivity. The extraction-and-loading machines are the determining factor in the performance of mining and transport machines in the technological flow of the quarry. When extracting a blasted rock mass with single-bucket excavators or loaders working in combination with bottom-hole conveyors, one uses self-propelled crushing and reloading units of various designs to grind large individual parts to fractions of conditioning size. The presence of a crushing and reloading unit in the pit-face along with the excavator requires an additional space for its placement, complicates the maneuvering of the equipment in the pit-face, and increases the number of personnel and the cost of maintaining the extraction-and-reloading operations. The article proposes an improved method for carrying out the extraction-and-loading process, as well as the design of extraction-and-grinding unit based on a quarry hydraulic excavator. The design of the proposed unit makes it possible to convert the cyclic process of scooping the rock mass into the continuous process of its loading on the bottom-hole conveyor. Using the extraction-and-grinding unit allows one to combine the processes of excavation, preliminary crushing and loading of the rock mass, which ensures an increase in the efficiency of mining operations.

  15. Metallurgical failure investigation of a pipe connector fracture of an expansion vessel; Werkstofftechnische Schadensuntersuchung des Abrisses einer Rohrverschraubung eines Ausgleichsbehaelters

    Energy Technology Data Exchange (ETDEWEB)

    Neidel, Andreas [Siemens AG, Power and Gas, Gas Turbines and Generators, Gasturbinenwerk Berlin (Germany). Werkstoffprueflabor

    2016-08-15

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

  16. Benchmark study of shear buckling of a cylindrical vessel

    International Nuclear Information System (INIS)

    Dostal, M.; Austin, N.M.; Peano, A.; Combescure, A.; Bastien, R.; Carnoy, E.G.

    1986-01-01

    The possibility of a buckling failure of the primary vessel subjected to seismic excitation has been considered, by all major designers of loop and pool type liquid metal cooled fast breeder reactors. The problem is particularly onerous in this type of reactor due to their large size, coupled with small wall thicknesses. This report details the results of the first phase in a joint European code validation exercise on the static shear buckling behaviour of thin, low aspect ratio stainless steel cylinders. Linear and non-linear finite element analyses were performed by four organizations using three different computer codes, i.e. NNC (UK)-ABAQUS, ISMES (Italy)-ABAQUS, CEA (France)-BILBO/INCA and NOVATOME (France)-NOVNL. The computed results were compared directly with experimental results. It was discovered that refined finite element models were essential if accurate buckling loads were to be calculated. Buckling analyses in 3D were therefore computationally expensive and 2D analyses, where applicable, proved an useful alternative. Traditional linear (Euler) bifurcation analysis seriously over-estimated the buckling loads by around 50 %. Extrapolation techniques can however be used to reduce this discrepancy. Elasto-plastic bifurcation analysis predicted conservative buckling loads close to the experimental value. Non-linear, large displacement analyses were performed on the vessel. The effect of geometrical imperfections in the vessel was considered. These analyses all over-estimated the experimental buckling load by 10 %-25 % and appeared to be largely insensitive to the initial imperfection size. Each of the codes appeared to predict reasonably well the final buckled geometry although the analytical load-deflection estimate did not agree exactly with the experiment

  17. Keep fatigue usage low for LTO. Benefits of load monitoring and related fatigue evaluations for long term operation

    International Nuclear Information System (INIS)

    Rothenhoefer, H.; Koenig, G.

    2012-01-01

    Design fatigue calculations normally cover a service life of 40 years. Based on design transients with a specified number of cycles the evaluations have to prove that the fatigue usage after 40 years will stay below 1. In 40+ years of operation real loads can differ much from design loads so that premature ageing can occur. For long term operation, monitoring of real loads and detailed fatigue analysis for selected locations can be used to optimize operational modes in order to reduce the loads causing fatigue. As a result fatigue usage can be kept below 1 even for 60+ years. (author)

  18. Pressure vessel for nuclear reactors

    International Nuclear Information System (INIS)

    1975-01-01

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

  19. Structural analysis of vacuum vessel and blanket support system for International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Kitamura, Kazunori; Koizumi, Kouichi; Takatsu, Hideyuki; Tada, Eisuke; Shimane, Hideo.

    1996-11-01

    Structural analyses of vacuum vessel and blanket support system have been performed to examine their integrated structural behavior under the design loads and to assess their structural feasibility, with two kinds of three-dimensional (3-D) FEM models; a detailed model with 18deg sector region to investigate the detailed mechanical behaviors of the blanket and vessel components under the several symmetric loads, and a 180deg torus model with relatively coarser meshes to assess the structural responses under the asymmetric VDE load. The analytical results obtained by both models were also compared for the several symmetric loads to check the equivalent mechanical stiffness of the 180deg torus model. As the results, most of the vessel and blanket components have sufficient mechanical integrities with the stress level below the allowable limit of the materials, while the lower parts of inboard/outboard back plate need to be reinforced by increasing the thickness and/or mounting a toroidal ring support at the lower edge of the back plate. Two types of eigenvalue analyses were also conducted with the 180deg torus model to investigate natural frequencies of the vessel torus support system and to assess the mechanical integrity of the elastic stability under the asymmetric VDE load. Analytical results show that the mechanical stiffness of the vessel gravity support should be higher in the view point of a seismic response, and that those of the blanket support structures should also be increased for the buckling strength against the VDE vertical force. (author)

  20. Use of plant operating history to define transient loads

    International Nuclear Information System (INIS)

    Dwivedy, K.K.

    1996-01-01

    Fatigue and crack growth analyses of components subjected to transient loads have been under continuous development during the recent past to include effects of environment on the components. The accuracy of the evaluation method on the predicted reliability of the components in the operating environment has become a focus of attention. Methods have integrated available material/component test data to improve evaluation techniques. However, in the area of definition of thermal transient loads the analyst still has to remain conservative, because no realistic guidelines have been developed to define thermal transients and their sequences. Fatigue re-evaluations of components are becoming increasingly necessary in operating plants as they age due to two reasons: (1) Components show age related degradation and cannot be repaired/replaced due to economic/logistic reasons. (2) Components experience transient conditions which were not considered in the original design. In either case, the evaluation of remaining life of components involves definition of transients and their sequence from the time the component was put in service until the end of life. As a common practice, initial plant design transients are used in a conservative definition of sequences to obtain results unrealistic for the situation, which sometimes leads to inaccurate estimate of the remaining life of components. The objective of this paper is to use plant operating history and plant monitoring data to provide procedures and techniques to define realistic transients for evaluation

  1. Interpreting ASME limits and philosophy in FEA of pressure vessel parts

    International Nuclear Information System (INIS)

    Bezerra, L.M.; Cruz, J.R.B.; Miranda, C.A.J.; Neto, M.M.

    1995-01-01

    In recent years there has been an effort to interpret finite element (FE) stress results on the light of the ASME B and PV rules and philosophy. Many task groups have issued guidelines on stress linearization and classifications. All those attempts have come up trying to cope modern FE techniques with the rules imposed by the ASME Code. This paper is an independent contribution to the Pressure Vessel Research Council (PVRC) groups which are studying the stress classification and the failure mechanism in a FE framework. This work tries to complement the interesting work by Hollinger and Hechmer presented in the PVP-94 in Minneapolis. In that paper, the authors examined a typical support skirt and showed relations between the skirt collapse load obtained by finite element analysis and the loads allowed from the ASME stress limits. To complement such paper, in the present article, different skirt geometry configurations are analyzed. The configurations here investigated consist of similar support skirts but with different angles of attachments between cylinder and cone parts. It will be possible to observe the influence of the bending stress in the collapse load and its relation to the allowable loads inferred from the ASME limits. A pressure vessel with torispherical head under internal pressure is also examined. Using elastic and limit load FEA, the present paper determines the collapse loads of the configurations. It sets up the relations between these collapse loads, stress categories, and limits dictated by the ASME Code Subsection NB. On the light of NB rules and philosophy, this paper shows how different methods of stress assessment, classification, and limits may influence in the design of a pressure vessel

  2. Tempest in a vessel

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-01-01

    As the ASN made some statements about anomalies of carbon content in the EPR vessel bottom and top, the author recalls and comments some technical issues to better understand the information published on this topic. He notably addresses the role of the vessel, briefly indicates its operating conditions, shape and structure, and mechanical components for the top, its material and mechanical properties, and test samples used to assess mechanical properties. He also comments the phenomenon of radio-induced embrittlement, the vessel manufacturing process, and evokes the applicable regulations. He quotes and comments statements made by the ASN and Areva which evoke further assessments of the concerned components

  3. Motor operated valve testing and the 'rate of loading' phenomenon

    International Nuclear Information System (INIS)

    Black, B.R.

    1991-01-01

    This paper discusses valve design features which affect the ability to predict motor operated valve (MOV) performance and reviews factors which should be considered when selecting switch settings to limit stem loads. Considerable attention is given to the rate of loading phenomenon which affects the relationship between valve stem thrust and actuator spring pack deflection. Equations are developed, and testing is discussed which permit the construction of an MOV dynamic model. Factors which must be considered when maintaining switch settings correct throughout the life of the plant are discussed. And switch setting acceptance criteria for use with baseline Static and Design Basis testing are suggested

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

    International Nuclear Information System (INIS)

    Parker, R.R.

    1998-01-01

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

  5. ITER in-vessel system design and performance

    Science.gov (United States)

    Parker, R. R.

    2000-03-01

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

  6. ITER in-vessel system design and performance

    International Nuclear Information System (INIS)

    Parker, R.R.

    2000-01-01

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

  7. 46 CFR 42.09-5 - All vessels-division into types.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false All vessels-division into types. 42.09-5 Section 42.09-5... BY SEA Load Line Assignments and Surveys-General Requirements § 42.09-5 All vessels—division into types. (a) For the purposes of this part, each vessel to which this part applies is either a Type “A” or...

  8. Experimental study and simulation of transformation induced plasticity, and multiphase behaviour of the 16MND5 vessel steel under aniso-thermal multiaxial loading

    International Nuclear Information System (INIS)

    Coret, M.

    2001-01-01

    This work deals with the aniso-thermal multiphase behaviour of the French vessel steel and more specially about the transformation plasticity in the cases of multiaxial non-proportional loadings paths. The first part of this report is devoted to the presentation of a high temperature tension-torsion experimental device enable of obtaining a large range of cooling rate. This experimental set-up is used to explore the transformation plasticity under proportional or non-proportional loading paths, during austenitic, bainitic and martensitic transformations. The results of the tests are compared to the Leblond's model. In the last part, we propose a two-scale behaviour model in which the type of each phase behaviour can be different. This meso-model is finally used to simulate two real tests on structures. (author) [fr

  9. Evaluation of buckling on containment metallic vessels

    International Nuclear Information System (INIS)

    Silveira, Renato Campos da; Mattar Neto, Miguel

    2000-01-01

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

  10. Robust nonlinear model predictive control for nuclear power plants in load following operations with bounded xenon oscillations

    International Nuclear Information System (INIS)

    Eliasi, H.; Menhaj, M.B.; Davilu, H.

    2011-01-01

    Research highlights: → In this work, a robust nonlinear model predictive control algorithm is developed. → This algorithm is applied to control the power level for load following. → The state constraints are imposed on the predicted trajectory during optimization. → The xenon oscillations are the main constraint for the load following problem. → In this algorithm, xenon oscillations are bounded within acceptable limits. - Abstract: One of the important operations in nuclear power plants is load-following in which imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation considered to be a constraint for the load-following operation. In this paper, a robust nonlinear model predictive control for the load-following operation problem is proposed that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO) strategy to maintain xenon oscillations to be bounded. The constant AO is a robust state constraint for load-following problem. The controller imposes restricted state constraints on the predicted trajectory during optimization which guarantees robust satisfaction of state constraints without restoring to a min-max optimization problem. Simulation results show that the proposed controller for the load-following operation is so effective so that the xenon oscillations kept bounded in the given region.

  11. Design optimization of a thin walled pressure vessel

    International Nuclear Information System (INIS)

    Sadiq, S.

    2001-01-01

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

  12. Guidelines for prediction of irradiation embrittlement of operating WWER-440 reactor pressure vessels

    International Nuclear Information System (INIS)

    2005-06-01

    This TECDOC has been developed under an International Atomic Energy Agency Coordinated Research Project (CRP) entitled Evaluation of Radiation Damage of WWER Reactor Pressure Vessels (RPV) using Database on RPV Materials to develop the guidelines for prediction of radiation damage to WWER-440 PRVs. The WWER-440 RPV was designed by OKB Gidropress, Russian Federation, the general designer. Prediction of irradiation embrittlement of RPV materials is usually done in accordance with relevant codes and standards that are based on the large amounts of information from surveillance and research programmes. The existing Russian code (standard for strength calculations of components and piping in NPPs - PNAE G 7-002-86) for the WWER RPV irradiation embrittlement assessment was approved more than twenty years ago and based mostly on the experimental data obtained in research reactors with accelerated irradiation. Nevertheless, it is still in use and generally consistent with new data. The present publication presents the analyses using all available data required for more precise prediction of radiation embrittlement of WWER-440 RPV materials. Based on the fact that it contains a large amount of data from surveillance programmes as well as research programmes, the IAEA International Database on RPV Materials (IDRPVM) is used for the detailed analysis of irradiation embrittlement of WWER RPV materials. Using IDRPVM, the guideline is developed for assessment of irradiation embrittlement of RPV ferritic materials as a result of degradation during operation. Two approaches, i.e. transition temperatures based on Charpy impact notch toughness, as well as based on static fracture toughness tests, are used in RPV integrity evaluation. The objectives of the TECDOC are the analysis of irradiation embrittlement data for WWER- 440 RPV materials using IDRPVM database, evaluation of predictive formulae depending on chemical composition of the material, neutron fluence, flux, and

  13. Assessment of environmentally assisted cracking in PWR pressure vessel steels

    International Nuclear Information System (INIS)

    Tice, D.R.

    1991-01-01

    There is a possibility that extension of pre-existing flaws in the reactor pressure vessel of a pressurised water reactor (PWR) may occur by environmentally assisted cracking, in particular by corrosion fatigue under cyclic transient loading. Crack growth predictions have usually been carried out using cyclic crack growth rate (da/dN) versus stress intensity range (δK) curves, such as those given in Section XI, Appendix A of the ASME Boiler and Pressure Vessel Code. However, the inherent time dependent nature of environmental cracking processes renders such an approach unrealistic. The present paper describes the development of an alternative time based assessment methodology. Illustrative calculations of expected crack growth of assumed defects made using the cyclic (ASME XIA) and time-based approaches are compared. The results illustrate that crack growth predicted by the time-based approach can be greater or less than that calculated by the traditional method. For a PWR operated with good control of water chemistry, actual crack growth rates are expected to be well below those predicted by the ASME code. (Author)

  14. Analysis of ex-vessel steam explosion with MC3D

    International Nuclear Information System (INIS)

    Leskovar, M.; Mavko, B.

    2007-01-01

    An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. In the paper, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which was developed for the simulation of fuel-coolant interactions. A comprehensive parametric study was performed varying the location of the melt release (central, left and right side melt pour), the cavity water subcooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to determine the most challenging ex-vessel steam explosion cases in a typical pressurized water reactor and to estimate the expected pressure loadings on the cavity walls. The performed analysis shows that for some ex-vessel steam explosion scenarios significantly higher pressure loads are predicted than obtained in the OECD programme SERENA Phase 1. (author)

  15. Stresses in reactor pressure vessel nozzles -- Calculations and experiments

    International Nuclear Information System (INIS)

    Brumovsky, M.; Polachova, H.

    1995-01-01

    Reactor pressure vessel nozzles are characterized by a high stress concentration which is critical in their low-cycle fatigue assessment. Program of experimental verification of stress/strain field distribution during elastic-plastic loading of a reactor pressure vessel WWER-1000 primary nozzle model in scale 1:3 is presented. While primary nozzle has an ID equal to 850 mm, the model nozzle has ID equal to 280 mm, and was made from 15Kh2NMFA type of steel. Calculation using analytical methods was performed. Comparison of results using different analytical methods -- Neuber's, Hardrath-Ohman's as well as equivalent energy ones, used in different reactor Codes -- is shown. Experimental verification was carried out on model nozzles loaded statically as well as by repeated loading, both in elastic-plastic region. Strain fields were measured using high-strain gauges, which were located in different distances from center of nozzle radius, thus different stress concentration values were reached. Comparison of calculated and experimental data are shown and compared

  16. Application of annealing for WWER vessels life extension

    International Nuclear Information System (INIS)

    Badanin, V.I.; Gorynin, I.V.; Nickolaev, V.A.; Dragunov, Y.G.; Fedorov, V.G.

    1989-01-01

    Safe operation of NPP is greatly dependent on the guarantee of reactor vessel brittle failure strength with account for the effect of radiation embrittlement of vessel material. Recovery of irradiated material properties is principally important way to extend radiation life of reactor vessel. The aim of this report is to demonstrate the efficiency of annealing for recovery of vessel material properties and extension of its service-life

  17. Resin-based preparation of HTGR fuels: operation of an engineering-scale uranium loading system

    International Nuclear Information System (INIS)

    Haas, P.A.

    1977-10-01

    The fuel particles for recycle of 233 U to High-Temperature Gas-Cooled Reactors are prepared from uranium-loaded carboxylic acid ion exchange resins which are subsequently carbonized, converted, and refabricated. The development and operation of individual items of equipment and of an integrated system are described for the resin-loading part of the process. This engineering-scale system was full scale with respect to a hot demonstration facility, but was operated with natural uranium. The feed uranium, which consisted of uranyl nitrate solution containing excess nitric acid, was loaded by exchange with resin in the hydrogen form. In order to obtain high loadings, the uranyl nitrate must be acid deficient; therefore, nitric acid was extracted by a liquid organic amine which was regenerated to discharge a NaNO 3 or NH 4 NO 3 solution waste. Water was removed from the uranyl nitrate solution by an evaporator that yielded condensate containing less than 0.5 ppM of uranium. The uranium-loaded resin was washed with condensate and dried to a controlled water content via microwave heating. The loading process was controlled via in-line measurements of the pH and density of the uranyl nitrate. The demonstrated capacity was 1 kg of uranium per hour for either batch loading contractors or a continuous column as the resin loading contractor. Fifty-four batch loading runs were made without a single failure of the process outlined in the chemical flowsheet or any evidence of inability to control the conditions dictated by the flowsheet

  18. Design of the ITER vacuum vessel

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  19. Assessment of Gearbox Operational Loads and Reliability under High Mean Wind Speeds

    DEFF Research Database (Denmark)

    Dabrowski, Dariusz; Natarajan, Anand

    2015-01-01

    wind turbine that operates in storm conditions with mean wind speeds less than 30 m/s is presented. In the study, normal shut-downs of a wind turbine in storm conditions were investigated. The analysis were conducted for two storm control strategies and different wind conditions from an extreme...... operating gust, normal turbulence model and extreme turbulence model. In the paper, loads in the planetary gear are quantified as well as the torsional moments in the main shaft. On the basis of simulation results the annual probability of failure of the gearbox in a wind turbine with soft storm controller...... is calculated, and compared with the one had the gearbox working in a wind turbine operating with hard storm controller. In the study, it was found that normal shut-downs do not have a significant influence on the ultimate loads in the gearbox, since they are related mostly to the gusts occurring during...

  20. Multi-purpose deployer for ITER in-vessel maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang-Hwan, E-mail: Chang-Hwan.CHOI@iter.org [ITER Organization, Route de Vinon-sur-Verdon, 13115 St Paul lez Durance (France); Tesini, Alessandro; Subramanian, Rajendran [ITER Organization, Route de Vinon-sur-Verdon, 13115 St Paul lez Durance (France); Rolfe, Alan; Mills, Simon; Scott, Robin; Froud, Tim; Haist, Bernhard; McCarron, Eddie [Oxford Technologies Ltd., 7 Nuffield Way, Abingdon, OXON (United Kingdom)

    2015-10-15

    Highlights: • ITER RH system called as the multi-purpose deployer (MPD) is introduced. • The MPD performs dust and tritium inventory control, in-service inspection. • The MPD performs leak localization, in-vessel diagnostics maintenance. • The MPD has nine degrees of freedom with a payload capacity up to 2 tons. - Abstract: The multi-purpose deployer (MPD) is a general purpose in-vessel remote handling (RH) system in the ITER RH system. The MPD provides the means for deployment and handling of in-vessel tools or components inside the vacuum vessel (VV) for dust and tritium inventory control, in-service inspection, leak localization, and in-vessel diagnostics. It also supports the operation of blanket first wall maintenance and neutral beam duct liner module maintenance operations. This paper describes the concept design of the MPD. The MPD is a cask based system, i.e. it stays in the hot cell building during the machine operation, and is deployed to the VV using the cask system for the in-vessel operations. The main part of the MPD is the articulated transporter which provides transportation and positioning of the in-vessel tools or components. The articulated transporter has nine degrees of freedom with a payload capacity up to 2 tons. The articulated transporter can cover the whole internal surface of the VV by switching between the four equatorial RH ports. Additionally it can use two non-RH equatorial ports to transfer large tools or components. A concept for in-cask tool exchange is developed which minimizes the cask transportation by allowing the MPD to stay in the VV during the tool exchange.

  1. Recent development for improving the PWR flexibility to load follow and frequency control operation

    International Nuclear Information System (INIS)

    Dubourg, M.

    1983-01-01

    The increasing production of nuclear electricity generated by PWR in the French network will modify the operating conditions of these plants for adjusting the electricity generation to the consumption. For assessing the adequacy of main components, FRAMATOME, in conjunction with Electricite de France and the Commissariat a l'Energie Atomique has undertaken a large R and D effort and initiated significant design changes for sustaining the new operating modes including. Daily load follow and frequency remote dispatch operation (+- 5% random fluctuation load around a present value). These new operating conditions generate additional mechanical and thermal sollicitations due to the frequent motion of control rod banks, consisting of: a) Mechanical fatigue cycling and wear at the level of control rod drive mechanisms (CRDM), control rods and guides tubes. b) Wear and thermal fatigue cycling at the level of fuel assemblies. This paper will present the various aspects of this program including: Identification of the most critical areas of components; Basic research in laboratories for resolving wear problems in PWR environment; Improvement of local hydraulics for reducing loads; Endurance testing of full scale components on testing facilities. (orig./GL)

  2. Is the bipolar vessel sealer device an effective tool in robotic surgery? A retrospective analysis of our experience and a meta-analysis of the literature about different robotic procedures by investigating operative data and post-operative course.

    Science.gov (United States)

    Ortenzi, Monica; Ghiselli, Roberto; Baldarelli, Maddalena; Cardinali, Luca; Guerrieri, Mario

    2018-04-01

    The latest robotic bipolar vessel sealing tools have been described to be effective allowing to perform procedures with reduced blood loss and shorter operative times. The aim of this study was to assess the efficacy and reliability of these devices applied in different robotic procedures. All robotic operations, between 2014 and 2016, were performed using the EndoWrist One VesselSealer (EWO, Intuitive Surgical, Sunnyvale, CA), a bipolar fully wristed device. Data, including age, gender, body mass index (BMI), were collected. Robot docking time, intraoperative blood loss, robot malfunctioning and overall operative time were analyzed. A meta-analysis of the literature was carried out to point the attention to three different parameters (mean blood loss, operating time and hospital stay) trying to identify how different coagulation devices may affect them. In 73 robotic procedures, the mean operative time was 118.2 minutes (75-125 minutes). Mean hospital stay was four days (2-10 days). There were two post-operative complications (2.74%). The bipolar vessel sealer offers the efficacy of bipolar diathermy and the advantages of a fully wristed instrument. It does not require any change of instruments for coagulation or involvement of the bedside assistant surgeon. These characteristics lead to a reduction in operative time.

  3. Development of a Remotely-operated Visual Inspection System for Reactor Vessel Bottommounted Instrument Penetrations of KSNP and Lessons Learned

    International Nuclear Information System (INIS)

    Jeong, Kyungmin; Choi, Youngsu; Lee, Sunguk; Seo, Yongchil; Kang, Jong Gyu; Kim, Seungho; Jung, Seungho

    2006-01-01

    In April 2003, South Texas Project Unit 1 made a surprising discovery of boron acid leakage from two nozzles from a bare-metal examination of the reactor vessel bottom-mounted instrument penetrations during a routine refueling outage. A small powdery substance about 150mg was found on the outside of two instrument guide penetration nozzles on the bottom of the reactor. The primary coolant water of pressurized water reactors has caused cracking in penetrations with Alloy 600 through a process called primary water stress corrosion cracking. In South Korea, it is required to conduct 100% visual inspection of the outside of instrument guide penetration nozzles on the bottom of PWRs to confirm the integrity of reactor vessel. This paper describes the remotely-operated visual inspection systems for reactor vessel bottom-mounted instrument penetrations dispatched two times to Youngkwang NPPs and discusses the lessons learned

  4. Structural evaluation of the Shippingport Reactor Pressure Vessel and Neutron Shield Tank package for impact and puncture loads

    International Nuclear Information System (INIS)

    Fischer, L.E.; Chou, C.K.; Lo, T.; Schwartz, M.W.

    1988-06-01

    A structural evaluation of Shippingport Reactor Pressure Vessel and Neutron Shield Tank package for impact and puncture loads under the normal and hypothetical accident conditions of 10 CFR 71 was performed. Component performance criteria for the Shippingport package and the corresponding structural acceptance criteria for these components were developed based on a review of the package geometry, the planned transport environment, and the external radiation standards and dispersal limits of 10 CFR 71. The evaluation was performed using structural analysis methods. A demonstration combining simplified model tests and nonlinear finite element analyses was made to substantiate the structural analysis methods used to evaluate the Shippingport package. The package was analyzed and the results indicate that the package meets external radiation standards and release limits of 10 CFR 71. 13 refs., 50 figs., 19 tabs

  5. Coupled thermo-mechanical creep analysis for boiling water reactor pressure vessel lower head

    International Nuclear Information System (INIS)

    Villanueva, Walter; Tran, Chi-Thanh; Kudinov, Pavel

    2012-01-01

    Highlights: ► We consider a severe accident in a BWR with melt pool formation in the lower head. ► We study the influence of pool depth on vessel failure mode with creep analysis. ► There are two modes of failure; ballooning of vessel bottom and a localized creep. ► External vessel cooling can suppress creep and subsequently prevent vessel failure. - Abstract: In this paper we consider a hypothetical severe accident in a Nordic-type boiling water reactor (BWR) at the stage of relocation of molten core materials to the lower head and subsequent debris bed and then melt pool formation. Nordic BWRs rely on reactor cavity flooding as a means for ex-vessel melt coolability and ultimate termination of the accident progression. However, different modes of vessel failure may result in different regimes of melt release from the vessel, which determine initial conditions for melt coolant interaction and eventually coolability of the debris bed. The goal of this study is to define if retention of decay-heated melt inside the reactor pressure vessel is possible and investigate modes of the vessel wall failure otherwise. The mode of failure is contingent upon the ultimate mechanical strength of the vessel structures under given mechanical and thermal loads and applied cooling measures. The influence of pool depth and respective transient thermal loads on the reactor vessel failure mode is studied with coupled thermo-mechanical creep analysis. Efficacy of control rod guide tube (CRGT) cooling and external vessel wall cooling as potential severe accident management measures is investigated. First, only CRGT cooling is considered in simulations revealing two different modes of vessel failure: (i) a ‘ballooning’ of the vessel bottom and (ii) a ‘localized creep’ concentrated within the vicinity of the top surface of the melt pool. Second, possibility of in-vessel retention with CRGT and external vessel cooling is investigated. We found that the external vessel

  6. Integrity of Magnox reactor steel pressure vessels

    International Nuclear Information System (INIS)

    Flewitt, P.E.J.; Williams, G.H.; Wright, M.B.

    1992-01-01

    The background to the safety assessment of the steel reactor pressure vessels for Magnox power stations is reviewed. The evolved philosophy adopted for the 1991 safety cases prepared for the continued operation of four Magnox power stations operated by Nuclear Electric plc is described, together with different aspects of the multi-legged integrity argument. The main revisions to the materials mechanical property data are addressed together with the assessment methodology adopted and their implications for the overall integrity argument formulated for the continued safe operation of these reactor pressure vessels. (author)

  7. Vacuum vessel port structures for ITER-FEAT

    International Nuclear Information System (INIS)

    Utin, Yu.; Ioki, K.; Komarov, V.; Krylov, V.; Kuzmin, E.; Labusov, I.; Miki, N.; Onozuka, M.; Rozov, V.; Sannazzaro, G.; Tesini, A.; Yamada, M.; Barthel, Th.

    2001-01-01

    The equatorial and the upper port structures are the most loaded among those of the ITER-FEAT vacuum vessel (VV). For all of these ports, the VV closure plate and the in-port components are integrated into the port plug. The plugs/port structures are affected by plasma events and must withstand high mechanical loads. Based on typical port plugs, this paper presents the conceptual design of the port structures (with emphasis on the supporting system), and the results of analyses performed

  8. Vacuum vessel port structures for ITER-FEAT

    Energy Technology Data Exchange (ETDEWEB)

    Utin, Yu.; Ioki, K.; Komarov, V.; Krylov, V.; Kuzmin, E.; Labusov, I.; Miki, N.; Onozuka, M.; Rozov, V.; Sannazzaro, G.; Tesini, A.; Yamada, M.; Barthel, Th

    2001-11-01

    The equatorial and the upper port structures are the most loaded among those of the ITER-FEAT vacuum vessel (VV). For all of these ports, the VV closure plate and the in-port components are integrated into the port plug. The plugs/port structures are affected by plasma events and must withstand high mechanical loads. Based on typical port plugs, this paper presents the conceptual design of the port structures (with emphasis on the supporting system), and the results of analyses performed.

  9. In-vessel remote maintenance of the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Tabor, M.A.; Hager, E.R.; Creedon, R.L.; Fisher, M.V.; Atkin, S.D.

    1987-01-01

    The Compact Ignition Tokamak (CIT) is the first deuterium-tritium (D-T) fusion device that will study the physics of an ignited plasma. The ability of the tokamak vacuum vessel to be maintained remotely while under vacuum has not been fully demonstrated on previous machines, and this ability will be critical to the efficient and safe operation of ignition devices. Although manned entry into the CIT vacuum vessel will be possible during the nonactivated stages of operation, remotely automated equipment will be used to assist in initial assembly of the vessel as well as to maintain all in-vessel components once the D-T burn is achieved. Remote maintenance and operation will be routinely required for replacement of thermal protection tiles, inspection of components, leak detection, and repair welding activities. Conceptual design to support these remote maintenance activities has been integrated with the conceptual design of the in-vessel components to provide a complete and practical remote maintenance system for CIT. The primary remote assembly and maintenance operations on CIT will be accomplished through two dedicated 37- x 100-cm ports on the main toroidal vessel. Each port contains a single articulated boom manipulator (ABM), which is capable of accessing half of the torus. The proposed ABM consists of a movable carriage assembly, telescoping two-part mast, and articulated link sections. 1 ref

  10. Preliminary study of S-CO{sub 2} cycle control logic for part load operation

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Yoonhan; Lee, Jeong Ik [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2016-05-15

    The benefits of S-CO{sub 2} cycle are relatively high efficiency under the mild turbine inlet temperature region, simple layout configuration and small foot-print. In addition, the safety of the SFR system can be inherently enhanced as the violent sodium-water reaction can be substituted with the mild sodium-CO{sub 2} reaction. 75MWe S-CO{sub 2} recompression cycle with radial type turbomachineries and PCHE was designed. Under various part load conditions (30-100% thermal load), off-design performance of the designed system was assessed, and different control logics were first tested. It was identified that the inventory control strategy is the most efficient logic for the part load operation. In the system operation, the compressor surge condition is seriously considered and controlled to avoid the system damage.

  11. The Impact of Climatological Conditions on Low Enriched Uranium Loading Station Operations for the HEU Blend Down Project

    International Nuclear Information System (INIS)

    Chang, R.C.

    2002-01-01

    A computer model was developed using COREsim to perform a time motion study for the Low Enriched Uranium (LEU) Loading Station operations. The project is to blend Highly Enriched Uranium (HEU) with Natural Uranium (NU) to produce LEU to be shipped to Tennessee Valley Authority (TVA) for further processing. To cope with a project cost reduction, the LEU Loading Station concept has changed from an enclosed building with air-conditioning to a partially enclosed building without air conditioning. The LEU Loading Station is within a radiological contaminated area; two pairs of coveralls and negative pressure respirator are required. As a result, inclement weather conditions, especially heat stress, will affect and impact the LEU loading operations. The purposes of the study are to determine the climatological impacts on LEU Loading operations, resources required for committed throughputs, and to find out the optimum process pathways for multi crews working simultaneously in the space-lim ited LEU Loading Station

  12. Contribution of materials investigations and operating experience of reactor vessel internals to PWRs' safety, performance and reliability

    International Nuclear Information System (INIS)

    Lemaire, E.; Monteil, N.; Jardin, N.; Doll, M.

    2015-01-01

    The Reactor Pressure Vessel Internals (RVI) include all the components inside the pressure vessel, except the nuclear fuel, the rod cluster assemblies and the instrumentation. The RVI consist of bolted and welded structures that are divided into two sub-assemblies: the upper internals which are removed at every refueling outage and the lower internals which are systematically removed for inspection at every 10-year outage. The main functions of the RVI are to position the core, to support it, and to provide a coolant flow by channeling the fluid. Moreover, the lower internals contribute to a neutron protection of the reactor pressure vessel by absorbing most of the neutron flux from the core. Depending on their location and material composition, the RVI components can face different ageing phenomena, that are actual or potential (such as wear, fatigue, stress corrosion cracking, irradiation assisted stress corrosion cracking, hardening and loss of ductility due to neutron irradiation, irradiation creep and irradiation swelling). EDF has developed a strategy for managing ageing and demonstrating the capacity of the RVI to perform their design functions over 40 years of operation. This overall approach is periodically revisited to take into account the most recent knowledge obtained from the following main topics: Safety Analyses, Research-Development programs, In-Service Inspection (ISI) results, Maintenance programs and Metallurgical Examinations. Based on continuous improvements in those fields, the goal of this paper is to present the way that materials investigations and operating experience obtained on RVI are managed by EDF to improve RVI safety, performance and reliability. It is shown that a perspective of 60 years of operation of RVI components is supported by large Research-Development efforts combined with field experience. (authors)

  13. 46 CFR 390.5 - Agreement vessels.

    Science.gov (United States)

    2010-10-01

    ... port to port, excluding equipment that needs frequent replacement due to normal wear and tear, and is... foreign country or points in two different foreign countries in the case of liquid and dry bulk cargo... vessel by means of wheeled technology; and (ii) That is: (A) Loaded at a port in the United States and...

  14. Ex-vessel boiling experiments: laboratory- and reactor-scale testing of the flooded cavity concept for in-vessel core retention. Pt. II. Reactor-scale boiling experiments of the flooded cavity concept for in-vessel core retention

    International Nuclear Information System (INIS)

    Chu, T.Y.; Bentz, J.H.; Slezak, S.E.; Pasedag, W.F.

    1997-01-01

    For pt.I see ibid., p.77-88 (1997). This paper summarizes the results of a reactor-scale ex-vessel boiling experiment for assessing the flooded cavity design of the heavy water new production reactor. The simulated reactor vessel has a cylindrical diameter of 3.7 m and a torispherical bottom head. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling mainly results from the gravity head, which in turn results from flooding the side of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid-solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion. The results show that, under prototypic heat load and heat flux distributions, the flooded cavity will be effective for in-vessel core retention in the heavy water new production reactor. The results also demonstrate that the heat dissipation requirement for in-vessel core retention, for the central region of the lower head of an AP-600 advanced light water reactor, can be met with the flooded cavity design. (orig.)

  15. Crashworthy sealed pressure vessel for plutonium transport

    International Nuclear Information System (INIS)

    Andersen, J.A.

    1980-01-01

    A rugged transportation package for the air shipment of radioisotopic materials was recently developed. This package includes a tough, sealed, stainless steel inner containment vessel of 1460 cc capacity. This vessel, intended for a mass load of up to 2 Kg PuO 2 in various isotopic forms (not to exceed 25 watts thermal activity), has a positive closure design consisting of a recessed, shouldered lid fastened to the vessel body by twelve stainless-steel bolts; sealing is accomplished by a ductile copper gasket in conjunction with knife-edge sealing beads on both the body and lid. Follow-on applications of this seal in newer, smaller packages for international air shipments of plutonium safeguards samples, and in newer, more optimized packages for greater payload and improved efficiency and utility, are briefly presented

  16. Flexible operation of a system of SPSs for load following

    Energy Technology Data Exchange (ETDEWEB)

    Collins, P Q; Tomkins, R

    1984-04-01

    Major conclusions of previous studies on the SPS (solar power satellite) project - the proposal to collect solar energy in orbit and to transmit it to earth as a coherent microwave beam for reconversion to electricity - are that the system appears likely to be technically feasible, but that it will be economically feasible only if major reductions are made in the cost of certain subsystems, most notably in the fields of photovoltaic cells and in space transportation and operations. The former field is currently experiencing rapid advances, along with the rest of the semiconductor industry, while the achievement of the necessary cost targets in the latter field will depend on the development of a family of fully reusable space transportation vehicles as a successor to the present U.S. space shuttle. In order to assess the system's full economic potential it is also necessary to examine the most profitable means of operation. Most studies to date have assumed that, because of the SPS's high capital cost and the corresponding need to maximise its usage, each satellite would be used to transmit base load power more or less continuously to a single rectenna. However, this assumption overlooks the fact that the value of electric power is not constant but varies during the day and over the year. It therefore narrows the scope for SPS operation, and ignores a major part of the system's potential value. It also sets the most demanding cost targets for the SPS, since it would be in competition with other base-load plant with the lowest operating costs.

  17. PWR vessel inspection performance improvements

    International Nuclear Information System (INIS)

    Blair Fairbrother, D.; Bodson, Francis

    1998-01-01

    A compact robot for ultrasonic inspection of reactor vessels has been developed that reduces setup logistics and schedule time for mandatory code inspections. Rather than installing a large structure to access the entire weld inspection area from its flange attachment, the compact robot examines welds in overlapping patches from a suction cup anchor to the shell wall. The compact robot size allows two robots to be operated in the vessel simultaneously. This significantly reduces the time required to complete the inspection. Experience to date indicates that time for vessel examinations can be reduced to fewer than four days. (author)

  18. 46 CFR 42.09-35 - Additional survey requirements for wood-hull vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Additional survey requirements for wood-hull vessels. 42.09-35 Section 42.09-35 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) LOAD LINES... Additional survey requirements for wood-hull vessels. (a) In addition to the requirements in § 42.09-25, the...

  19. An interior vessel viewing system for DIII-D

    International Nuclear Information System (INIS)

    Senior, R.

    1989-11-01

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

  20. Reactor vessel decommissioning project. Final report

    International Nuclear Information System (INIS)

    Schoonen, D.H.

    1984-09-01

    This report describes a reactor vessel decommissioning project; it documents and explains the project objectives, scope, performance results, and sodium removal process. The project was successfully completed in FY-1983, within budget and without significant problems or adverse impact on the environment. Waste generated by the operation included the reactor vessel, drained sodium, and liquid, solid, and gaseous wastes which were significantly less than project estimates. Personnel radiation exposures were minimized, such that the project total was one-half the predicted exposure level. Except for the sodium removed, the material remaining in the reactor vessel is essentially the same as when the vessel arrived for processing

  1. 75 FR 75486 - Maritime Security Directive 104-6 (Rev. 4); Guidelines for U.S. Vessels Operating in High Risk...

    Science.gov (United States)

    2010-12-03

    ... Directive 104-6 (Rev. 4); Guidelines for U.S. Vessels Operating in High Risk Waters AGENCY: Coast Guard, DHS... Maritime Transportation Security Act (MTSA) on international voyages through or in designated high risk... MARSEC Directives are available at your local Captain of the Port (COTP) office. Phone numbers and...

  2. Research on Operation and Control Strategy of 600MW PWR in Load Follow

    Energy Technology Data Exchange (ETDEWEB)

    Qu, Bing Yang; Cao, Xin Rong [Harbin Engineering University, Harbin (China); Li, Han Chen [China Nuclear Power Engineering Co., Beijing (China)

    2014-08-15

    600MW Pressurized Water Reactor (PWR) is designed to operate in Constant Axial Offset Control (CAOC) strategy with base load originally. By calculations over a typical load follow scenario '12-3-6-3 {sup (}100-50-100%FP) via the CASMO-4E and SIMULATE-3 package, values of core operating parameter have been examined. With the progress of the nuclear power industry, advanced reactors are considered to have a good performance in load follow, economy and flexibility. Under the premise of fuel loading and structural dimensions unchanged, two independent control rod groups M and AO are used in 600MW pressurized water reactor to provide fine control of both the core reactivity and axial power distribution, which is named ' Improved G strategy .' The influences of different control rod distributions, composition materials, and overlap steps had in power changes have been examined in a comparative study to choose the optimal one.Then we simulate a range of load follow scenarios of the redesigned 600MW core without adjusting soluble boron concentration in the begin, middle and end of first cycle. This paper additionally demonstrated the moderator temperature coefficient and shutdown margin values of the reactor in Improved G strategy to compare with the thermal safety design criteria. It's demonstrated that adequate adjustment of control rod groups enable the core to perform load follow through Improved G strategy in 80% of cycle and save a large volume of liquid effluent particularly toward the end of cycle.

  3. A quantitative methodology for reactor vessel pressurized thermal shock decision making

    International Nuclear Information System (INIS)

    Ackerson, D.S.; Balkey, K.R.; Meyer, T.A.; Ofstun, R.P.; Rupprecht, S.D.; Sharp, D.R.

    1983-01-01

    The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). Previous reactor vessel integrity concerns have led to changes in vessel and plant system design and to operating procedures, and increased attention to the PTS issue is causing consideration of further modifications. Events such as excess feedwater, loss of normal feedwater, and steam generator tube rupture have led to significant primary system cooldowns. Each of these cooldown transients occurred concurrently with a relatively high primary system pressure. Considerations of these and other postulated cooldown events has drawn attention to the impact of operator action and control system effects on reactor vessel PTS. A methodology, which couples event sequence analysis with probabilistic fracture mechanics analyses, was developed to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation. (orig./RW)

  4. Nuclear reactor pressure vessel integrity insurance by crack arrestability evaluation using load from CVN tests

    International Nuclear Information System (INIS)

    Fabry, A.

    1997-01-01

    The present work is undertaken in the framework of nuclear reactor pressure vessel (RPV) surveillance and aims at revisiting the crack arrest approach to structural integrity insurance. This approach, performed under normal plant operation conditions, can also offer an attractive alternative to the crack initiation philosophy promoted for accidental analysis. To this end, an accidental conservative, cost effective and robust methodology is forwarded and demonstrated: it makes use of the crack arrest information contained in the instrumented Charpy V-notch impact test and/or in the shear fracture appearance of broken samples. Particular attention is paid to the appraisal of uncertainties and the related safety margin. The resulting capability is placed in perspective with the state-of-the-art crack initiation methodology based on the slow bend testing of recracked specimens, presently under standardization world-wide. The investigation leads to highlight three conceptual weaknesses of current enfgineering and regulatory practices. Improved crack arrestability evaluation emerges as an optimal approach to insure safe PWR operation up to design end-of-life and beyond

  5. Nuclear reactor pressure vessel integrity insurance by crack arrestability evaluation using load from CVN tests

    Energy Technology Data Exchange (ETDEWEB)

    Fabry, A.

    1997-10-15

    The present work is undertaken in the framework of nuclear reactor pressure vessel (RPV) surveillance and aims at revisiting the crack arrest approach to structural integrity insurance. This approach, performed under normal plant operation conditions, can also offer an attractive alternative to the crack initiation philosophy promoted for accidental analysis. To this end, an accidental conservative, cost effective and robust methodology is forwarded and demonstrated: it makes use of the crack arrest information contained in the instrumented Charpy V-notch impact test and/or in the shear fracture appearance of broken samples. Particular attention is paid to the appraisal of uncertainties and the related safety margin. The resulting capability is placed in perspective with the state-of-the-art crack initiation methodology based on the slow bend testing of recracked specimens, presently under standardization world-wide. The investigation leads to highlight three conceptual weaknesses of current enfgineering and regulatory practices. Improved crack arrestability evaluation emerges as an optimal approach to insure safe PWR operation up to design end-of-life and beyond.

  6. 48 CFR 47.303-8 - F.a.s. vessel, port of shipment.

    Science.gov (United States)

    2010-10-01

    ..., up to this point; (3) Provide a clean dock or ship's receipt; (4) Be responsible for any loss of and... CONTRACT MANAGEMENT TRANSPORTATION Transportation in Supply Contracts 47.303-8 F.a.s. vessel, port of... order and condition alongside the ocean vessel and within reach of its loading tackle, at the point of...

  7. Optimization of Helium Vessel Design for ILC Cavities

    Energy Technology Data Exchange (ETDEWEB)

    Fratangelo, Enrico [Univ. of Pisa (Italy)

    2009-01-01

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

  8. Automatic segmentation of blood vessels from retinal fundus images ...

    Indian Academy of Sciences (India)

    The retinal blood vessels were segmented through color space conversion and color channel .... Retinal blood vessel segmentation was also attempted through multi-scale operators. A few works in this ... fundus camera at 35 degrees field of view. The image ... vessel segmentation is available from two human observers.

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

    International Nuclear Information System (INIS)

    1996-01-01

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

  10. Load Assembly Design of the FAST Machine

    International Nuclear Information System (INIS)

    Cucchiaro, A.; Albanese, R.; Ambrosino, G.; Brolatti, G.; Calabro, G.; Cocilovo, V.; Coletti, A.; Coletti, R.; Costa, P.; Frosi, P.; Crescenzi, F.; Crisanti, F.; Granucci, G.; Maddaluno, G.; Pericoli Ridolfini, V.; Pizzuto, A.; Rita, C.; Ramogida, G.

    2008-01-01

    The FAST [1][2] (Fusion Advanced Studies Torus) load assembly, which includes the Vacuum Vessel (VV) and its internal components, the magnet system and the poloidal field coils, is presented in this paper. FAST operates at a wide range [3][4] of parameters from high performance H-Mode (B T up to 8.5 T; I P up to 8 MA) as well as Advance Tokamak operation (I P =3 MA), and full non inductive current scenario (I P =2 MA). Helium gas at 30K is used for cooling the resistive copper magnets. That allows for a pulse duration up to 170 s (∼ 40 times τres) at 3MA/ 3.5T. To limit the TF magnet ripple ferromagnetic within acceptable values insert have been introduced inside the outboard area of the VV. The VV, segmented by 20 degree modules, is capable to accommodate 40 MW RF power systems. The machine has been designed to house10 MW Negative Neutral Beam injection (NNBI) systems. Tungsten (W) and Liquid Lithium (L-Li) have been chosen as the divertor plates material, and Argon and Neon as the injected impurities to mitigate the thermal loads

  11. Residential CCHP microgrid with load aggregator: Operation mode, pricing strategy, and optimal dispatch

    International Nuclear Information System (INIS)

    Gu, Wei; Lu, Shuai; Wu, Zhi; Zhang, Xuesong; Zhou, Jinhui; Zhao, Bo; Wang, Jun

    2017-01-01

    Highlights: •A bilateral transaction mode for the residential CCHP microgrid is proposed. •An energy pricing strategy for the residential CCHP system is proposed. •A novel integrated demand response for the residential loads is proposed. •Two-stage operation optimization model for the CCHP microgrid is proposed. •Operations of typical days and annual scale of the CCHP microgrid are studied. -- Abstract: As the global energy crisis, environmental pollution, and global warming grow in intensity, increasing attention is being paid to combined cooling, heating, and power (CCHP) systems that realize high-efficiency cascade utilization of energy. This paper proposes a bilateral transaction mechanism between a residential CCHP system and a load aggregator (LA). The variable energy cost of the CCHP system is analyzed, based on which an energy pricing strategy for the CCHP system is proposed. Under this pricing strategy, the electricity price is constant, while the heat/cool price is ladder-shaped and dependent on the relationship between the electrical, heat, and cool loads. For the LA, an integrated demand response program is proposed that combines electricity-load shifting and a flexible heating/cooling supply, in which a thermodynamic model of buildings is used to determine the appropriate range of heating/cooling supply. Subsequently, a two-stage optimal dispatch model is proposed for the energy system that comprises the CCHP system and the LA. Case studies consisting of three scenarios (winter, summer, and excessive seasons) are delivered to demonstrate the effectiveness of the proposed approach, and the performance of the proposed pricing strategy is also evaluated by annual operation simulations.

  12. Test of 6-in.-thick pressure vessels. Series 4: intermediate test vessels V-5 and V-9 with inside nozzle corner cracks

    International Nuclear Information System (INIS)

    Merkle, J.G.; Robinson, G.C.; Holz, P.P.; Smith, J.E.

    1977-01-01

    Failure testing is described for two 99-cm-diam (39-in.), 15.2-cm-thick (6-in.) steel pressure vessels, each containing one flawed nozzle. Vessel V-5 was tested at 88 0 C (190 0 F) and failed by leaking without fracturing after extensive stable crack growth. Vessel V-9 was tested at 25 0 C (75 0 F) and failed by fracturing. Material properties measured before the tests were used for pretest and posttest fracture analyses. Test results supported by analysis indicate that inside nozzle corner cracks are not subject to plane strain under pressure loading. The preparation of inside nozzle corner cracks is described in detail. Extensive experimental data are tabulated and plotted

  13. Proposal of Ex-Vessel dosimetry for pressure vessel Atucha II

    International Nuclear Information System (INIS)

    Chiaraviglio, N.; Bazzana, S.

    2013-01-01

    Nuclear reactor dosimetry has the purpose of guarantee that changes in material mechanical properties of critical materials do not compromise the reactor safety. In PWR in which the top of the reactor vessel is open once a year, is possible to use Charpy specimens to measure the change in mechanical properties. Atucha II nuclear power plant is a reactor with on-line refueling so there is no access to the inside of the pressure vessel. Because of this, ex-vessel dosimetry must be performed and mechanical properties changes must be inferred from radiation damage estimations. This damage can be calculated using displacement per atom cross sections and a transport code such as MCNP. To increase results reliability it is proposed to make a neutron spectrum unfolding using activation dosimeters irradiated during one operation cycle of the power plant. In this work we present a dosimetry proposal for such end, made in base of unfolding procedures and experimental background. (author) [es

  14. Subjective and objective symptoms of psychical load in nuclear power plant operators in connection with shift operation

    International Nuclear Information System (INIS)

    Kozeny, J.; Prochazkova, Z.

    1992-01-01

    A group of 56 operators (mean age=34.6 years, S.D. 3.4 year, range=30-40 years; university graduates of technical direction) in a nuclear power plant was followed during the morning, afternoon and night shift from the standpoint of subjective symptoms (an evaluation scale assessed at the beginning, in the middle and at the end of the shift) and objective symptoms (critical flicker frequency determined at the same time) of the psychical load. No significant differences were found by multivariation analysis between subjective perception of the work load and its objective reflection. The night shift, where the load lies particularly in the monotony of stimuli, appears to exert a higher psychical load aimed at maintaining mental activity when compared with the afternoon and morning shift. In contrast to subjective assessment of the psychical condition where the data fluctuated during the whole shift, the psychical activity reflected in the objective data, decreased until the middle of the shift when stabilization occurred. (author) 1 tab., 8 refs

  15. Assessment of Gearbox Operational Loads and Reliability under High Mean Wind Speeds

    OpenAIRE

    Dabrowski, Dariusz; Natarajan, Anand

    2015-01-01

    This paper investigates the dynamic loads occurring in the drivetrain of wind turbines with a focus on offshore applications. Herein a model of the gearbox of the 5 MW wind turbine is presented. The model is developed in a multi-body framework using commercial software MSC ADAMS. Validation of the model was based on the experimental data provided by NREL for 750 kW prototype gearbox. Failures of gearboxes caused by high dynamic loads have a significant influence on the cost of operation of wi...

  16. Integral reactor vessel related to power reactor safety

    International Nuclear Information System (INIS)

    Widart, J.; Scailteur, A.

    1978-01-01

    Integral design applied to PWR pressure vessels allows to reach a high level of safety because: 1) it presents a better balance of the material in the geometry, resulting in an improved stress level (mainly faulted condition loadings); 2) location and geometry of the welds are designed in order to get a very sound pressure boundary of the upper part of the vessel; 3) the new location and geometry of the welds allow an easy ISI in such a way that ambiguity surrounding defect size or locaton is practically suppressed. (author)

  17. Power system operation risk analysis considering charging load self-management of plug-in hybrid electric vehicles

    International Nuclear Information System (INIS)

    Liu, Zhe; Wang, Dan; Jia, Hongjie; Djilali, Ned

    2014-01-01

    Highlights: • The interactive mechanism between system and PHEVs is presented. • The charging load self-management without sacrificing user requirements is proposed. • The charging load self-management is coupled to system operation risk analysis. • The charging load self-management can reduce the extra risk brought by PHEVs. • The charging load self-management can shift charging power to the time with low risk. - Abstract: Many jurisdictions around the world are supporting the adoption of electric vehicles through incentives and the deployment of a charging infrastructure to reduce greenhouse gas emissions. Plug-in hybrid electric vehicles (PHEVs), with offer mature technology and stable performance, are expected to gain an increasingly larger share of the consumer market. The aggregated effect on power grid due to large-scale penetration of PHEVs needs to be analyzed. Nighttime-charging which typically characterizes PHEVs is helpful in filling the nocturnal load valley, but random charging of large PHEV fleets at night may result in new load peaks and valleys. Active response strategy is a potentially effective solution to mitigate the additional risks brought by the integration of PHEVs. This paper proposes a power system operation risk analysis framework in which charging load self-management is used to control system operation risk. We describe an interactive mechanism between the system and PHEVs in conjunction with a smart charging model is to simulate the time series power consumption of PHEVs. The charging load is managed with adjusting the state transition boundaries and without violating the users’ desired charging constraints. The load curtailment caused by voltage or power flow violation after outages is determined by controlling charging power. At the same time, the system risk is maintained under an acceptable level through charging load self-management. The proposed method is implemented using the Roy Billinton Test System (RBTS) and

  18. Code boiler and pressure vessel life assessment

    International Nuclear Information System (INIS)

    Farr, J.R.

    1992-01-01

    In the United States of America and in Canada, laws and controls for determining life assessment for continued operation of equipment exist only for those pressure vessels built to Section III and evaluated according to Section XI. In this presentation, some of those considerations which are made in the USA and Canada for deciding on life or condition assessment of boilers and pressure vessels designed and constructed to other sections of the ASME Boiler and Pressure Vessel Code are reviewed. Life assessment or condition assesssment is essential in determining what is necessary for continued operation. With no ASME rules being adopted by laws or regulations, other than OSHA in the USA and similar environmental controls in Canada, to control life assessment for continued operation, the equipment owner must decide if assessment is to be done and how much to do. Some of those considerations are reviewed along with methods and procedures to make an assessment along with a discussion of where the ASME B and PV Code currently stands regarding continued operation. (orig.)

  19. 40 CFR 1042.130 - Installation instructions for vessel manufacturers.

    Science.gov (United States)

    2010-07-01

    ...) AIR POLLUTION CONTROLS CONTROL OF EMISSIONS FROM NEW AND IN-USE MARINE COMPRESSION-IGNITION ENGINES...-speed operation, tell vessel manufacturers not to install the engines in variable-speed applications or... vessel manufacturers. (a) If you sell an engine for someone else to install in a vessel, give the engine...

  20. Application of annealing for extension of WWER vessel lives

    International Nuclear Information System (INIS)

    Badanin, V.; Dragunow, Yu.G.; Fedorov, V.; Gorynin, I.; Nickolaev, V.

    1992-01-01

    The safe operation of nuclear power plants (NPP) is dependent upon the assurance that the reactor pressure vessel will not fail in a brittle manner when the effects of radiation embrittlement are taken into account. The recovery of the properties of the irradiated materials is an important way of extending the operating life of a reactor vessel. The intent of this paper is to demonstrate the efficiency of thermal annealing for the recovery of reactor vessel material properties and to present the implications for extended service life. In order to substantiate the application of annealing to the extensior of the service life of vessels, detailed investigations were conducted which involved thermal annealing temperature and time, fast neutron fluence, and metallurgical factors (i.e. impurity contents) on the recovery of properties after the annealing of irradiated materials. Similar studies were continued to determine predictive methods for radiation embrittlement after repeated annealings. In May 1987 the first pilot annealing of a commercial reactor vessel (Novo-Voronezhskaya, III, NPP) was performed. The development of the annealing equipment and investigations performed to test the annealing process proved successful, and an improved safe operation for the reactor vessel was thus atttained providing for an extended service life. (orig.)

  1. Consequence evaluation of hypothetical reactor pressure vessel support failure

    International Nuclear Information System (INIS)

    Lu, S.C.; Holman, G.S.; Lambert, H.E.

    1991-01-01

    This paper describes a consequence evaluation to address safety concerns raised by the radiation embrittlement of the reactor pressure vessel (RPV) supports for the Trojan nuclear power plant. The study comprises a structural evaluation and an effects evaluation and assumes that all four reactor vessel supports have completely lost the load carrying capability. The structural evaluation concludes that the Trojan reactor coolant loop (RCL) piping is capable of transferring loads to the steam generator (SG) supports and the reactor coolant pump (RCP) supports and that the SG supports and the RCP supports have sufficient design margins to accommodate additional loads transferred to them through the RCL piping. The effects evaluation, employing a systems analysis approach, investigates initiating events and the reliability of the engineered safeguard systems as the RPV is subject to movements caused by the RPV support failure. The evaluation identifies a number of areas for further investigation and concludes that a hypothetical failure of the Trojan RPV supports due to radiation embrittlement will not result in consequences of significant safety concerns. (author)

  2. The Analysis of the Causes of Emergencies on the Vessels

    Directory of Open Access Journals (Sweden)

    Alicja Mrozowska

    2017-12-01

    Full Text Available The article discusses the results of research conducted on the vessels, covering a wide spectrum of issues relating to the exploitation of vessels of various flags, as well as operating security and safety systems on board. The main aim of the study was to collect numbers of data directly from the crew, for examples: indicate by the crew marine areas with the greatest probability of occurrence of casualties and incidents, trying to the definition the causes of their occurrence, prevention actions used on board and analyses operating safety systems used on the various type of vessels. The analysis of research became the basis to identify strengths and weaknesses areas of the vessel operation. The author proposes a solution to be implemented on board and emphasizes meaning of safety management system.

  3. Predicting Vessel Trajectories from Ais Data Using R

    Science.gov (United States)

    2017-06-01

    Source: Hampton (2009). A vessel operator with AIS is able to get useful information about the other vessels in the area by selecting a vessel icon ...random forest model on our computer. All calculations are done on a MacBook-Pro with 2.7GHz quad-core Intel Core i7, and 16GB of memory . H2O allows us

  4. LOGISTIC SYSTEM OF LOAD DELIVERY AND QUALITY OF ITS OPERATION

    Directory of Open Access Journals (Sweden)

    O. G. Drozdovskaya

    2006-01-01

    Full Text Available The paper considers an opportunity for obtaining a competitive advantage by a transport and dispatch service company in the market of transport services while establishing a logistic system of load delivery. A model of delivery system, an universal scheme of system designing for every specific case are presented and also indices for evaluation of its operational quality are proposed in the paper.

  5. Flexible operation of a system of SPSs for load following

    Energy Technology Data Exchange (ETDEWEB)

    Collins, P.Q.; Tomkins, R.

    1984-04-01

    Major conclusions of previous studies on the SPS (solar power satellite) project - the proposal to collect solar energy in orbit and to transmit it to earth as a coherent microwave beam for reconversion to electricity - are that the system appears likely to be technically feasible, but that it will be economically feasible only if major reductions are made in the cost of certain subsystems, most notably in the fields of photovoltaic cells and in space transportation and operations. The former field is currently experiencing rapid advances, along with the rest of the semiconductor industry, while the achievement of the necessary cost targets in the latter field will depend on the development of a family of fully reusable space transportation vehicles as a successor to the present U.S. space shuttle. In order to assess the system's full economic potential it is also necessary to examine the most profitable means of operation. Most studies to date have assumed that, because of the SPS's high capital cost and the corresponding need to maximise its usage, each satellite would be used to transmit base load power more or less continuously to a single rectenna. However, this assumption overlooks the fact that the value of electric power is not constant but varies during the day and over the year. It therefore narrows the scope for SPS operation, and ignores a major part of the system's potential value. It also sets the most demanding cost targets for the SPS, since it would be in competition with other base-load plant with the lowest operating costs.

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

    International Nuclear Information System (INIS)

    Reimann, K.J.

    1980-12-01

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

  7. Stress criteria for nuclear vessel concrete

    International Nuclear Information System (INIS)

    Costes, D.

    1975-01-01

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

  8. Neutronics studies for the design of the European DEMO vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  9. Interpretation of Strain Measurements on Nuclear Pressure Vessels

    DEFF Research Database (Denmark)

    Andersen, Svend Ib Smidt; Engbæk, Preben

    1980-01-01

    with a negligible zeroshift. However, deviations from linear behaviour are observed in several cases. This nonlinearity can be explained by friction (flange connections) or by gaps (concentrical nozzles) in certain regions, whereas local plastic deformations during the first pressure loadings of the vessel seem...

  10. Light-water-reactor pressure-vessel surveillance dosimetry using solid-state track recorders

    International Nuclear Information System (INIS)

    Ruddy, F.H.; Roberts, J.H.; Gold, R.; Preston, C.C.

    1983-07-01

    The accumulation of neutron dose by the pressure vessel of an operating nuclear power plant results in damage in the form of steel embrittlement. In order to ascertain the safe operating lifetime of the reactor pressure vessel, dosimetric measurements must be made to evaluate the neutron dose to the pressure vessel and relate this dose to the cumulative radiation damage. Advanced dosimetry techniques are being evaluated for surveillance of operating reactors. Solid-state track recorder (SSTR) techniques are included among these advanced dosimetry techniques. Described herein are low neutron fluence calibration and standardization measurements that are being carried out in pressure vessel mockup benchmark neutron fields in the USA, Belgium, and England. In addition, high fluence SSTR dosimetry capsules have been irradiated with metallurgical specimens in a pressure vessel mockup facility. The design and deployment of advances SSTR dosimetry capsules in operating power reactors are also described

  11. Power reactor pressure vessel benchmarks

    International Nuclear Information System (INIS)

    Rahn, F.J.

    1978-01-01

    A review is given of the current status of experimental and calculational benchmarks for use in understanding the radiation embrittlement effects in the pressure vessels of operating light water power reactors. The requirements of such benchmarks for application to pressure vessel dosimetry are stated. Recent developments in active and passive neutron detectors sensitive in the ranges of importance to embrittlement studies are summarized and recommendations for improvements in the benchmark are made. (author)

  12. 46 CFR 173.007 - Location of the hook load.

    Science.gov (United States)

    2010-10-01

    ... PERTAINING TO VESSEL USE Lifting § 173.007 Location of the hook load. When doing the calculations required in this subpart, the hook load must be considered to be located at the head of the crane. ...

  13. Commissioning result of the KSTAR in-vessel cryo-pump

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y. B.; Lee, H. J.; Park, Y.M. [National Fusion Research Institute, Daejeon (Korea, Republic of); and others

    2013-12-15

    KSTAR in-vessel cryo-pump has been installed in the vacuum vessel top and bottom side with up-down symmetry for the better plasma density control in the D-shape H-mode. The cryogenic helium lines of the in-vessel cryo-pump are located at the vertical positions from the vacuum vessel torus center 2,000 mm. The inductive electrical potential has been optimized to reduce risk of electrical breakdown during plasma disruption. In-vessel cryo-pump consists of three parts of coaxial circular shape components; cryo-panel, thermal shield and particle shield. The cryo-panel is cooled down to below 4.5 K. The cryo-panel and thermal shields were made by Inconel 625 tube for higher mechanical strength. The thermal shields and their cooling tubes were annealed in air environment to improve the thermal radiation emissivity on the surface. Surface of cryo-panel was electro-polished to minimize the thermal radiation heat load. The in-vessel cryo-pump was pre-assembled on a test bed in 180 degree segment base. The leak test was carried out after the thermal shock between room temperature to LN2 one before installing them into vacuum vessel. Two segments were welded together in the vacuum vessel and final leak test was performed after the thermal shock. Commissioning of the in-vessel cryo-pump was carried out using a temporary liquid helium supply system.

  14. Foundamental characteristics of layered pressure vessel

    International Nuclear Information System (INIS)

    Moriwaki, Yoshikazu; Fugino, Masayuki; Shimizu, Yasuhiro; Nakamura, Takeshi

    1978-01-01

    Pressure vessels become larger and the working pressure become higher with the remarkable development of petroleum, chemical, thermal power generation and atomic energy industries. Multi-layered pressure vessels can be manufactured cheaply without large installations, and large wall thickness can be made, therefore they are suitable for large pressure vessels. The stress and deformation behaviors of such vessels are very complex because of the effect of frictional force working between layers. In this study, the phenomena arising in multiple layers and the difference as compared with single wall were studied fundamentally as one step for analyzing multi-layered pressure vessels as a whole. Finite element technique was employed as the analyzing method, and the behavior of multiple layers was analyzed, regarding it as multiple contact problem. The behavior of multiple layers seems to appear conspicuously in case of bending load, therefore the basic characteristics regarding bending were examined. The evaluation of interfacial stiffness was carried out by experiment. The computer program for analyzing multiple contact problem was developed. In order to examine the validity of the program, comparison with the analytical solution heretofore and the result of calculation by finite element technique was carried out. Moreover, the experimental proof with multi-layered models was made. The frictional force between layers hardly contributes to the stiffness. (Kako, I.)

  15. Automation-aided Task Loads Index based on the Automation Rate Reflecting the Effects on Human Operators in NPPs

    International Nuclear Information System (INIS)

    Lee, Seungmin; Seong, Poonghyun; Kim, Jonghyun

    2013-01-01

    Many researchers have found that a high automation rate does not guarantee high performance. Therefore, to reflect the effects of automation on human performance, a new estimation method of the automation rate that considers the effects of automation on human operators in nuclear power plants (NPPs) was suggested. These suggested measures express how much automation support human operators but it cannot express the change of human operators' workload, whether the human operators' workload is increased or decreased. Before considering automation rates, whether the adopted automation is good or bad might be estimated in advance. In this study, to estimate the appropriateness of automation according to the change of the human operators' task loads, automation-aided task loads index is suggested based on the concept of the suggested automation rate. To insure plant safety and efficiency on behalf of human operators, various automation systems have been installed in NPPs, and many works which were previously conducted by human operators can now be supported by computer-based operator aids. According to the characteristics of the automation types, the estimation method of the system automation and the cognitive automation rate were suggested. The proposed estimation method concentrates on the effects of introducing automation, so it directly express how much the automated system support human operators. Based on the suggested automation rates, the way to estimate how much the automated system can affect the human operators' cognitive task load is suggested in this study. When there is no automation, the calculated index is 1, and it means there is no change of human operators' task load

  16. Pressurized thermal shock probabilistic fracture mechanics sensitivity analysis for Yankee Rowe reactor pressure vessel

    International Nuclear Information System (INIS)

    Dickson, T.L.; Cheverton, R.D.; Bryson, J.W.; Bass, B.R.; Shum, D.K.M.; Keeney, J.A.

    1993-08-01

    The Nuclear Regulatory Commission (NRC) requested Oak Ridge National Laboratory (ORNL) to perform a pressurized-thermal-shock (PTS) probabilistic fracture mechanics (PFM) sensitivity analysis for the Yankee Rowe reactor pressure vessel, for the fluences corresponding to the end of operating cycle 22, using a specific small-break-loss- of-coolant transient as the loading condition. Regions of the vessel with distinguishing features were to be treated individually -- upper axial weld, lower axial weld, circumferential weld, upper plate spot welds, upper plate regions between the spot welds, lower plate spot welds, and the lower plate regions between the spot welds. The fracture analysis methods used in the analysis of through-clad surface flaws were those contained in the established OCA-P computer code, which was developed during the Integrated Pressurized Thermal Shock (IPTS) Program. The NRC request specified that the OCA-P code be enhanced for this study to also calculate the conditional probabilities of failure for subclad flaws and embedded flaws. The results of this sensitivity analysis provide the NRC with (1) data that could be used to assess the relative influence of a number of key input parameters in the Yankee Rowe PTS analysis and (2) data that can be used for readily determining the probability of vessel failure once a more accurate indication of vessel embrittlement becomes available. This report is designated as HSST report No. 117

  17. Assessment of In-vessel corium retention for VVER-440/V213

    International Nuclear Information System (INIS)

    Matejovic, P.; Barnak, M.; Bachraty, M.; Berky, R.

    2011-01-01

    In-vessel corium retention (IVR) via external reactor vessel cooling (ERVC) has been recognised as a feasible and promising severe accident management strategy for VVER-440/V213 reactors. In general, the avoiding of boiling crisis on outer (cooled) RPV (reactor pressure vessel) surface is sufficient condition for preserving the RPV integrity. The crucial point of the proposed IVR concept for VVER-440/V213 is the narrow gap between elliptical lower head and thermal and biological shield. In the cold conditions the width of this gap is only about 2 cm and would be even lower in hot IVR conditions, when the reactor wall is subjected to large thermal gradients due to temperature difference between the hot inner surface (loaded by corium) and cold outer surface (which is cooled by water in flooded cavity). Sufficient gap should remain free for coolant flow for the success of the proposed IVR concept. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width are of primarily importance. Two different approaches were used for the estimation of the thermal load: a conservative approach and a transient approach, both were computed with the ASTEC code. The structural analysis of RPV subjected to IVR load was performed using the finite element method (FEM) code ANSYS release 10.0. From the results obtained it follows, that even when the RPV is subjected to limiting loading conditions during severe accident, there should be sufficient gap width (∼ 1 cm) between RPV wall and thermal/biological shield for the coolant flow in natural circulation regime alongside the outer surface of the RPV wall

  18. Advanced operational strategy for the IRIS reactor: Load follow through mechanical shim (MSHIM)

    International Nuclear Information System (INIS)

    Franceschini, Fausto; Petrovic, Bojan

    2008-01-01

    design suited for MSHIM operation has been devised. Nine load follow scenarios covering a wide range of possible operating requirements, including Westinghouse design basis plus others proposed by EPRI for Advanced LWRs, have been successfully performed through the control rod banks movement only, without soluble boron adjustment, and maintaining power peaking factors within the acceptable range. Thus, IRIS provides improved operation by enabling load follow through MSHIM

  19. Oxidation effect on steel corrosion and thermal loads during corium melt in-vessel retention

    Energy Technology Data Exchange (ETDEWEB)

    Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Sulatsky, A.A.; Almjashev, V.I. [Alexandrov Scientific-Research Technology Institute (NITI), Sosnovy Bor (Russian Federation); Bechta, S.V. [KTH, Stockholm (Sweden); Gusarov, V.V. [SPb State Technology University (SPbGTU), St. Petersburg (Russian Federation); Barrachin, M. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), St Paul lez Durance (France); Bottomley, P.D., E-mail: paul.bottomley@ec.europa.eu [EC-Joint Research Centre, Institute for Transuranium Elements (ITU), Karlsruhe (Germany); Fischer, M. [AREVA GmbH, Erlangen (Germany); Piluso, P. [Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA), Cadarache, St Paul lez Durance (France)

    2014-10-15

    Highlights: • The METCOR facility simulates vessel steel corrosion in contact with corium. • Steel corrosion rates in UO{sub 2+x}–ZrO{sub 2}–FeO{sub y} coria accelerate above 1050 K. • However corrosion rates can also be limited by melt O{sub 2} supply. • The impact of this on in-vessel retention (IVR) strategy is discussed. - Abstract: During a severe accident with core meltdown, the in-vessel molten core retention is challenged by the vessel steel ablation due to thermal and physicochemical interaction of melt with steel. In accidents with oxidizing atmosphere above the melt surface, a low melting point UO{sub 2+x}–ZrO{sub 2}–FeO{sub y} corium pool can form. In this case ablation of the RPV steel interacting with the molten corium is a corrosion process. Experiments carried out within the International Scientific and Technology Center's (ISTC) METCOR Project have shown that the corrosion rate can vary and depends on both surface temperature of the RPV steel and oxygen potential of the melt. If the oxygen potential is low, the corrosion rate is controlled by the solid phase diffusion of Fe ions in the corrosion layer. At high oxygen potential and steel surface layer temperature of 1050 °C and higher, the corrosion rate intensifies because of corrosion layer liquefaction and liquid phase diffusion of Fe ions. The paper analyzes conditions under which corrosion intensification occurs and can impact on in-vessel melt retention (IVR)

  20. Ocean passenger vessels : migrating south for the winter

    Science.gov (United States)

    2010-01-01

    In response to consumer demand, the passenger vessels that operate from seaports along the Atlantic, Gulf, and Pacific coasts alternate between north and south. Passenger vessels that sail out of ports such as New York, Baltimore and Seattle in the s...

  1. IMPROVEMENT OF PERFORMANCE OF DUAL FUEL ENGINE OPERATED AT PART LOAD

    Directory of Open Access Journals (Sweden)

    N. Kapilan

    2010-12-01

    Full Text Available Rising petroleum prices, an increasing threat to the environment from exhaust emissions, global warming and the threat of supply instabilities has led to the choice of inedible Mahua oil (MO as one of the main alternative fuels to diesel oil in India. In the present work, MO was converted into biodiesel by transesterification using methanol and sodium hydroxide. The cost of Mahua oil biodiesel (MOB is higher than diesel. Hence liquefied petroleum gas (LPG, which is one of the cheapest gaseous fuels available in India, was fumigated along with the air to reduce the operating cost and to reduce emissions. The dual fuel engine resulted in lower efficiency and higher emissions at part load. Hence in the present work, the injection time was varied and the performance of the dual fuel engine was studied. From the engine tests, it is observed that an advanced injection time results in higher efficiency and lower emissions. Hence, advancing the injection timing is one of the ways of increasing the efficiency of LPG+MOB dual fuel engine operated at part load.

  2. Dynamic response of the high flux isotope reactor structure caused by nearby heavy load drop

    International Nuclear Information System (INIS)

    Chang, Shih-Jung.

    1995-01-01

    A heavy load of 50,000 lb is assumed to drop from 10 ft above the bottom of the High Flux Isotope Reactor (HFIR) pool at the loading station. The consequences of the dynamic impact to the bottom slab of the pool and to the nearby HFIR reactor vessel are analyzed by applying the ABAQUS computer code The results show that both the BM vessel structure and its supporting legs are subjected to elastic disturbances only and, therefore, will not be damaged. The bottom slab of the pool, however, will be damaged to about half of the slab thickness. The velocity response spectrum at the concrete floor next to the HFIR vessel as a result of the vibration caused by the impact is obtained. It is concluded, that the damage caused by heavy load drop at the loading station is controlled by the slab damage and the nearby HFIR vessel and the supporting legs will not be damaged

  3. Proactive life extension of pressure vessels

    Science.gov (United States)

    Mager, Lloyd

    1998-03-01

    For a company to maintain its competitive edge in today's global market every opportunity to gain an advantage must be exploited. Many companies are strategically focusing on improved utilization of existing equipment as well as regulatory compliance. Abbott Laboratories is no exception. Pharmaceutical companies such as Abbott Laboratories realize that reliability and availability of their production equipment is critical to be successful and competitive. Abbott Laboratories, like many of our competitors, is working to improve safety, minimize downtime and maximize the productivity and efficiency of key production equipment such as the pressure vessels utilized in our processes. The correct strategy in obtaining these objectives is to perform meaningful inspection with prioritization based on hazard analysis and risk. The inspection data gathered in Abbott Laboratories pressure vessel program allows informed decisions leading to improved process control. The results of the program are reduced risks to the corporation and employees when operating pressure retaining equipment. Accurate and meaningful inspection methods become the cornerstone of a program allowing proper preventative maintenance actions to occur. Successful preventative/predictive maintenance programs must utilize meaningful nondestructive evaluation techniques and inspection methods. Nondestructive examination methods require accurate useful tools that allow rapid inspection for the entire pressure vessel. Results from the examination must allow the owner to prove compliance of all applicable regulatory laws and codes. At Abbott Laboratories the use of advanced NDE techniques, primarily B-scan ultrasonics, has provided us with the proper tools allowing us to obtain our objectives. Abbott Laboratories uses B-scan ultrasonics utilizing a pulse echo pitch catch technique to provide essential data on our pressure vessels. Equipment downtime is reduced because the nondestructive examination usually takes

  4. Method for offshore loading of a tanker and construction of said tanker. Fremgangsmte og farty for offshore oljelasting

    Energy Technology Data Exchange (ETDEWEB)

    Lerstad, A.; Breivik, K.; verli, O.; Egge, T.G.; Smedal, A.

    1994-07-04

    The invention relates to a vessel for offshore loading. The vessel comprises a number of centre tanks arranged in the longitudinal direction of the vessel, and a number of wing tanks located in pairs on either side of respective centre tanks. A loading line is connected to the centre tanks via respective valves, and adjacent centre tanks are connected to each other via bulkhead valves. The method comprises the steps firstly effecting filling of a desired number of the centre tanks. These are substantially narrower and having a correspondingly smaller volume than the wing tanks. Thereafter opening bulkhead valves between a centre tank and an appurtenant wing tank pair and sluicing out oil from centre tanks into the two wing tanks by means of the static pressure from the centre tank. The bulkhead valves between the centre tanks being open, so that the wing tank pair is filled as quickly as possible by suitable operation of the bulkhead valves between the centre tanks and of the valves of the loading line to the centre tanks. Closing the bulkhead valves to the two wing tanks when these and the appurtenant centre tank are filled to a desired level, and repeating the filling process to the steps and successively for additional wing tank pairs after that all centre tanks have been filled again. 10 figs.

  5. 75 FR 29358 - Maritime Security Directive 104-6 (Rev 2 and 3); Guidelines for U.S. Vessels Operating in High...

    Science.gov (United States)

    2010-05-25

    ... designated high risk waters, and provides additional anti-piracy guidance and mandatory measures for these vessels operating in these areas where acts of piracy and armed robbery against ships are prevalent... piratical activities. The combination of piracy and weak rule of law in the region offers a potential...

  6. DESIGN OF A VIBRATION AND STRESS MEASUREMENT SYSTEM FOR AN ADVANCED POWER REACTOR 1400 REACTOR VESSEL INTERNALS COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM

    OpenAIRE

    KO, DO-YOUNG; KIM, KYU-HYUNG

    2013-01-01

    In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful prepa...

  7. Limit analysis and design of containment vessels

    International Nuclear Information System (INIS)

    Save, M.

    1984-01-01

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

  8. Status and Perspectives of Nuclear Reactor Pressure Vessel Life Extension up to 60 Years Operation in Belgium

    Energy Technology Data Exchange (ETDEWEB)

    Lucon, E.; Chaouadi, R.; Scibetta, M.; Van Walle, E.

    2009-09-15

    The scope of this report involves a safety evaluation of the reactor pressure vessel (RPV) against neutron embrittlement, in the most severely irradiation region (belt line) and in the event of a pressurized thermal shock. The irreplaceable RPV is considered to be the most critical component for lifetime considerations of the nuclear power plant. However, an application for operation extension will also depend upon a number of additional considerations, including the technical assessment of other plant components, as well as non-technical arguments (e.g. political, environmental, economical, strategical that are outside the scope this report. In the hypothesis of a request for operation extension, it is the responsibility of the utilities to provide the safety authorities with an exhaustive dossier demonstrating that safe extended operation is guaranteed. The role of the safety authorities is to critically evaluate the safety dossier for eventually granting the operation extension.

  9. Status and Perspectives of Nuclear Reactor Pressure Vessel Life Extension up to 60 Years Operation in Belgium

    International Nuclear Information System (INIS)

    Lucon, E.; Chaouadi, R.; Scibetta, M.; Van Walle, E.

    2009-01-01

    The scope of this report involves a safety evaluation of the reactor pressure vessel (RPV) against neutron embrittlement, in the most severely irradiation region (belt line) and in the event of a pressurized thermal shock. The irreplaceable RPV is considered to be the most critical component for lifetime considerations of the nuclear power plant. However, an application for operation extension will also depend upon a number of additional considerations, including the technical assessment of other plant components, as well as non-technical arguments (e.g. political, environmental, economical, strategical that are outside the scope this report. In the hypothesis of a request for operation extension, it is the responsibility of the utilities to provide the safety authorities with an exhaustive dossier demonstrating that safe extended operation is guaranteed. The role of the safety authorities is to critically evaluate the safety dossier for eventually granting the operation extension.

  10. Electrical discharge machining for vessel sample removal

    International Nuclear Information System (INIS)

    Litka, T.J.

    1993-01-01

    Due to aging-related problems or essential metallurgy information (plant-life extension or decommissioning) of nuclear plants, sample removal from vessels may be required as part of an examination. Vessel or cladding samples with cracks may be removed to determine the cause of cracking. Vessel weld samples may be removed to determine the weld metallurgy. In all cases, an engineering analysis must be done prior to sample removal to determine the vessel's integrity upon sample removal. Electrical discharge machining (EDM) is being used for in-vessel nuclear power plant vessel sampling. Machining operations in reactor coolant system (RCS) components must be accomplished while collecting machining chips that could cause damage if they become part of the flow stream. The debris from EDM is a fine talclike particulate (no chips), which can be collected by flushing and filtration

  11. Evaluation of creep damage due to stress relaxation in SA533 grade B class 1 and SA508 class 3 pressure vessel steels

    International Nuclear Information System (INIS)

    Hoffmann, C.L.; Urko, W.

    1993-01-01

    Creep damage can result from stress relaxation of residual stresses in components when exposed to high temperature thermal cycles. Pressure vessels, such as the reactor vessel of the modular high-temperature gas reactor (MHTGR), which normally operate at temperatures well below the creep range can develop relatively high residual stresses in high stress locations. During short term excursions to elevated-temperatures, creep damage can be produced by the loadings on the vessel. In addition, residual stresses will relax out, causing greater creep damage in the pressure vessel material than might otherwise be calculated. The evaluation described in this paper assesses the magnitude of the creep damage due to relaxation of residual stresses resulting from short term exposure of the pressure vessel material to temperatures in the creep range. Creep relaxation curves were generated for SA533 Grade B, Class 1 and SA508 Class 3 pressure vessel steels using finite element analysis of a simple uniaxial truss loaded under constant strain conditions to produce an initial axial stress equal to 1.25 times the material yield strength at temperature. The strain is held constant for 1000 hours at prescribed temperatures from 700 F to 1000 F. The material creep law is used to calculate the relaxed stress for each time increment. The calculated stress relaxation versus time curves are compared with stress relaxation test data. Creep damage fractions are calculated by integrating the stress relaxation versus time curves and performing a linear creep damage summation using the minimum stress to rupture curves at the respective relaxation temperatures. Cumulative creep damage due to stress relaxation as a function of time and temperature is derived from the linear damage summation

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  13. Experimental investigation of creep behavior of reactor vessel lower head

    International Nuclear Information System (INIS)

    Chu, T.Y.; Pilch, M.; Bentz, J.H.; Behbahani, A.

    1998-03-01

    The objective of the USNRC supported Lower Head Failure (LHF) Experiment Program at Sandia National Laboratories is to experimentally investigate and characterize the failure of the reactor pressure vessel (RPV) lower head due to the thermal and pressure loads of a severe accident. The experimental program is complemented by a modeling program focused on the development of a constitutive formulation for use in standard finite element structure mechanics codes. The problem is of importance because: lower head failure defines the initial conditions of all ex-vessel events; the inability of state-of-the-art models to simulate the result of the TMI-II accident (Stickler, et al. 1993); and TMI-II results suggest the possibility of in-vessel cooling, and creep deformation may be a precursor to water ingression leading to in-vessel cooling

  14. Radio-frequency electrical design of the WEST long pulse and load-resilient ICRH launchers

    Energy Technology Data Exchange (ETDEWEB)

    Helou, Walid, E-mail: walid.helou@cea.fr [CEA, IRFM, F-13108 St-Paul-Lez-Durance (France); Colas, Laurent; Hillairet, Julien [CEA, IRFM, F-13108 St-Paul-Lez-Durance (France); Milanesio, Daniele [Department of Electronics, Politecnico di Torino, Torino (Italy); Mollard, Patrick [CEA, IRFM, F-13108 St-Paul-Lez-Durance (France); Argouarch, Arnaud [CEA DAM/DIF/DP2I, Bruyère le Chatel (France); Berger-By, Gilles; Bernard, Jean-Michel [CEA, IRFM, F-13108 St-Paul-Lez-Durance (France); Chen, Zhaoxi [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Delaplanche, Jean-Marc [CEA, IRFM, F-13108 St-Paul-Lez-Durance (France); Dumortier, Pierre; Durodié, Frédéric [Laboratoire de physique des plasmas de l’ERM, Laboratorium voor plasmafysica van de KMS – (LPP-ERM/KMS), Ecole royale militaire–Koninklijke militaire school, BE-1000 Brussels (Belgium); Ekedahl, Annika; Fedorczak, Nicolas; Ferlay, Fabien; Goniche, Marc [CEA, IRFM, F-13108 St-Paul-Lez-Durance (France); Jacquot, Jonathan [Max-Planck Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching (Germany); Joffrin, Emmanuel; Litaudon, Xavier; Lombard, Gilles [CEA, IRFM, F-13108 St-Paul-Lez-Durance (France); and others

    2015-10-15

    Highlights: • Three new ion cyclotron resonance heating launchers designed for WEST. • Operation at 3 MW/launcher for 30 s and 1 MW/launcher for 1000 s on H-mode plasmas. • Unique combination of continuous-wave operation at high power and load tolerance. • International team led by the CEA/IRFM. • RF design performed using electromagnetic solvers and electric circuit calculations. - Abstract: Three new ion cyclotron resonance heating (ICRH) launchers have been designed for the WEST project (W-Tungsten Environment in Steady-state Tokamak) in order to operate at 3 MW/launcher for 30 s and 1 MW/launcher for 1000 s on H-mode plasmas. These new launchers will be to date the first ICRH launchers to offer the unique combination of continuous-wave (CW) operation at high power and load tolerance capabilities for coupling on H-mode edge. The radio-frequency (RF) design optimization process has been carried out using full-wave electromagnetic solvers combined with electric circuit calculations. Cavity modes occurring between the launchers structures and the vacuum vessel ports have been evaluated and cleared out.

  15. Optimal Operation Method of Smart House by Controllable Loads based on Smart Grid Topology

    Science.gov (United States)

    Yoza, Akihiro; Uchida, Kosuke; Yona, Atsushi; Senju, Tomonobu

    2013-08-01

    From the perspective of global warming suppression and depletion of energy resources, renewable energy such as wind generation (WG) and photovoltaic generation (PV) are getting attention in distribution systems. Additionally, all electrification apartment house or residence such as DC smart house have increased in recent years. However, due to fluctuating power from renewable energy sources and loads, supply-demand balancing fluctuations of power system become problematic. Therefore, "smart grid" has become very popular in the worldwide. This article presents a methodology for optimal operation of a smart grid to minimize the interconnection point power flow fluctuations. To achieve the proposed optimal operation, we use distributed controllable loads such as battery and heat pump. By minimizing the interconnection point power flow fluctuations, it is possible to reduce the maximum electric power consumption and the electric cost. This system consists of photovoltaics generator, heat pump, battery, solar collector, and load. In order to verify the effectiveness of the proposed system, MATLAB is used in simulations.

  16. Behaviour of Viscoelastic - Viscoplastic Spheres and Cylinders - Partly Plastic Vessel Walls

    DEFF Research Database (Denmark)

    Ottosen, N. Saabye

    1985-01-01

    The material model consists of a viscoelastic Burgers element and an additional viscoplastic Bingham element when the effective stress exceeds the yield stress. For partly plastic vessel walls, expressions are derived for the stress and strain state in pressurised or relaxation loaded thick......-walled cylinders in plane strain and spheres. For the spherical problem, the material compressibility is accounted for. The influence of the different material parameters on the behaviour of the vessels is evaluated. It is shown that the magnitude of the Maxwell viscosity is of major importance for the long......-term behaviour of thick-walled partly plastic vessels....

  17. Behaviour of Viscoelastic - Viscoplastic Spheres and Cylinders - Fully Plastic Vessel Walls

    DEFF Research Database (Denmark)

    Ottosen, N. Saabye

    1985-01-01

    The material model consists of a viscoelastic Burgers element and an additional viscoplastic Bingham element when the effective stress exceeds the yield stress. For fully plastic vessel walls, exact closed-form expressions arc derived for the stress and strain state in pressurised or relaxation...... loaded thick-walled cylinders in plane strain and spheres. For the spherical problem, the material compressibility is accounted for. The influence of the different material parameters on the behaviour of the vessels is evaluated. It is shown that the magnitude of the Maxwell viscosity is of major...... importance for the long-term behaviour of thick-walled fully plastic vessels....

  18. Extension of Operating Range in Pump-Turbines. Influence of Head and Load

    Directory of Open Access Journals (Sweden)

    Carme Valero

    2017-12-01

    Full Text Available Due to the increasing share of new renewable energies like wind and solar in the generation of electricity the need for power regulation and energy storage is becoming of paramount importance. One of the systems to store huge amounts of energy is pumped storage using reversible hydropower units. The machines used in these power plants are pump-turbines, which can operate as a pump and as a turbine. The surplus of electrical energy during low consumption hours can be converted into potential hydraulic energy by pumping water to a higher level. The stored energy can be converted into electricity again by operating the runner as a turbine. Due to new regulation requirements machines have to extend the operating range in order to match energy generation with consumption for the grid stability. In this paper the consequences of extending the operating range in existing pump-turbines have been studied. For that purpose, the data obtained after two years of condition monitoring were analyzed. Vibrations and pressure fluctuations of two pump-turbines of 85 MW each have been studied during pump and turbine operation. For turbine operation the effects of extending the operating range from the standard range of 45–85 MW to and increased range of 20–85 MW were analyzed. The change in vibration levels and signatures at very low load are presented with the identification of the phenomena that occur under these conditions. The influence of head in the vibration behavior is also presented. The appearance of fluid instabilities generated at part load that may produce power swing is also presented. Finally, the effect of head on the vibration levels for pump operation is shown and analyzed.

  19. 76 FR 2402 - Maritime Security Directive 104-6 (Rev 5); Guidelines for U.S. Vessels Operating in High Risk Waters

    Science.gov (United States)

    2011-01-13

    ... Directive 104-6 (Rev 5); Guidelines for U.S. Vessels Operating in High Risk Waters AGENCY: Coast Guard, DHS... designated high risk waters, and provides additional counter-piracy guidance and mandatory measures for these... MARSEC Directives are available at your local Captain of the Port (COTP) office. Phone numbers and...

  20. Field measurement of the piping system vibration of Ko-Ri unit 4 during the load-following operation

    International Nuclear Information System (INIS)

    Chung, Tae-Young; Hong, Sung-Yull; Kim, Bum-Nyun.

    1989-01-01

    During the load-following operation of nuclear power plants, flow rate, temperature, and pressure in the piping system can be varied by changing the electric power output level, and these variations can cause different vibration phenomena in the piping system. The piping system vibration is important because it is directly related to the dynamic stress of the piping system and can affect the life of the piping system through structural fatigue. An assessment of vibration levels for the classes II and III piping systems of the Ko-Ri Unit 4950-MW nuclear power plant was performed according to the given pattern of the load-following operation to study its feasibility from the viewpoint of piping system vibration. The classes II and III piping system vibration of the Ko-Ri Unit 4 may not cause any potential problem under the given pattern of the load-following operation; however, it is recommended that long-term operation in the 85 to 95% power range be avoided as much as possible

  1. Economic Aspects of Load Following in Flexible Operation. Annex IV

    International Nuclear Information System (INIS)

    2018-01-01

    To supplement this publication, the IAEA has prepared an economic study to quantify the cost–revenue aspects for nuclear power plant operations in mixed energy systems, including flexible operation. The purpose of this study was to explore economic opportunities for using nuclear energy in future power markets with increasing deployment of renewable energy, specifically investigating the following: — How much power ramping could future reactors support and at what penalty? — What type of flexibility is required on the grid — larger down power and faster ramp rate for load following or increased start/stop capability? What is the optimal level of flexible capacity in a system? — How can nuclear energy be used in different grid environments (e.g. different sizes or complexities) to cope with variable power demand and fluctuating supply from renewable energy sources? — What are the economic implications (including different revenue streams) for using large reactors? — How do market regulation and real market interaction affect the economic viability of using nuclear power under the various system modes? To answer these questions, an analytical case study approach, which directly assesses the difference between energy specific production costs for a baseload nuclear system and a load following nuclear system, was selected. Such a study, based on a large scale country level power plant dispatching model, could assess the difference between the total short term costs of producing a given amount of electricity by both types of operation in a particular system. This approach may answer the questions of whether adding flexible nuclear operation to the power system is capable of generating overall system benefits by minimizing total system costs at different levels. This annex briefly summarizes the IAEA study, together with insights from the available sources, and underlines the pressing needs in filling the missing knowledge gaps in the economic aspects of

  2. Power Generation by Zinc Antimonide Thin Film under Various Load Resistances at its Critical Operating Temperature

    DEFF Research Database (Denmark)

    Mir Hosseini, Seyed Mojtaba; Rezaniakolaei, Alireza; Rosendahl, Lasse Aistrup

    slightly reduces during unload conditions, although it is expected that by eliminating load in each step, the initial amount of voltage exactly repeats. Similar behavior is observed for Seebeck coefficient distribution versus time of working particularly in lower load resistances. Based on variation...... thin films operating under different load resistances at around its critical operating temperature, 400 ᵒC. The thermoelement is subjected to constant hot side temperature and to room temperature at the cold junction in order to measure the thin film TEG’s sample performance. The nominal loads equal...... to 10, 15, 20, 25, 30, 35, 40, 45… 175, and also 200 Ohms were applied. The results show that the value of the Seebeck coefficient is 0.0002 [V/K] for the specimen, which is in agreement with quantities of other zinc antimonide bulks materials in literature. The results also show that the voltage...

  3. Automation-aided Task Loads Index based on the Automation Rate Reflecting the Effects on Human Operators in NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seungmin; Seong, Poonghyun [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kim, Jonghyun [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2013-05-15

    Many researchers have found that a high automation rate does not guarantee high performance. Therefore, to reflect the effects of automation on human performance, a new estimation method of the automation rate that considers the effects of automation on human operators in nuclear power plants (NPPs) was suggested. These suggested measures express how much automation support human operators but it cannot express the change of human operators' workload, whether the human operators' workload is increased or decreased. Before considering automation rates, whether the adopted automation is good or bad might be estimated in advance. In this study, to estimate the appropriateness of automation according to the change of the human operators' task loads, automation-aided task loads index is suggested based on the concept of the suggested automation rate. To insure plant safety and efficiency on behalf of human operators, various automation systems have been installed in NPPs, and many works which were previously conducted by human operators can now be supported by computer-based operator aids. According to the characteristics of the automation types, the estimation method of the system automation and the cognitive automation rate were suggested. The proposed estimation method concentrates on the effects of introducing automation, so it directly express how much the automated system support human operators. Based on the suggested automation rates, the way to estimate how much the automated system can affect the human operators' cognitive task load is suggested in this study. When there is no automation, the calculated index is 1, and it means there is no change of human operators' task load.

  4. Development of advanced manufacturing technologies for low cost hydrogen storage vessels

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-29

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

  5. Fatigue Load Modeling and Control for Wind Turbines based on Hysteresis Operators

    DEFF Research Database (Denmark)

    Barradas Berglind, Jose de Jesus; Wisniewski, Rafal; Soltani, Mohsen

    2015-01-01

    method based on hysteresis operators, which can be used in control loops. Furthermore, we propose a model predictive control (MPC) strategy that incorporates the online fatigue estimation through the objective function, where the ultimate goal in mind is to reduce the fatigue load of the wind turbine......The focus of this work is on fatigue load modeling and controller design for the wind turbine level. The main purpose is to include a model of the damage effects caused by the fatigue of the wind turbine components in the controller design process. This paper addresses an online fatigue estimation...

  6. 46 CFR 97.30-1 - Repairs to boilers and pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Repairs to boilers and pressure vessels. 97.30-1 Section... VESSELS OPERATIONS Reports of Accidents, Repairs, and Unsafe Equipment § 97.30-1 Repairs to boilers and pressure vessels. (a) Before making any repairs to boilers or unfired pressure vessels, the chief engineer...

  7. 46 CFR 196.30-1 - Repairs to boilers and pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Repairs to boilers and pressure vessels. 196.30-1... VESSELS OPERATIONS Reports of Accidents, Repairs, and Unsafe Equipment § 196.30-1 Repairs to boilers and pressure vessels. (a) Before making any repairs to boilers or unfired pressure vessels, the Chief Engineer...

  8. Firefighter's compressed air breathing system pressure vessel development program

    Science.gov (United States)

    Beck, E. J.

    1974-01-01

    The research to design, fabricate, test, and deliver a pressure vessel for the main component in an improved high-performance firefighter's breathing system is reported. The principal physical and performance characteristics of the vessel which were required are: (1) maximum weight of 9.0 lb; (2) maximum operating pressure of 4500 psig (charge pressure of 4000 psig); (3) minimum contained volume of 280 in. 3; (4) proof pressure of 6750 psig; (5) minimum burst pressure of 9000 psig following operational and service life; and (6) a minimum service life of 15 years. The vessel developed to fulfill the requirements described was completely sucessful, i.e., every category of performence was satisfied. The average weight of the vessel was found to be about 8.3 lb, well below the 9.0 lb specification requirement.

  9. In-vessel tritium retention and removal in ITER-FEAT

    International Nuclear Information System (INIS)

    Federici, G.; Brooks, J.N.; Iseli, M.; Wu, C.H.

    2001-01-01

    Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points, to withstand the high thermal loads expected during off-normal events, e.g., disruption, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., ∝350 g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R and D work is required to narrow the remaining uncertainties are also briefly discussed. (orig.)

  10. In-vessel tritium retention and removal in ITER-FEAT

    Energy Technology Data Exchange (ETDEWEB)

    Federici, G. [ITER Garching Joint Work Site, Garching (Germany); Brooks, J.N. [Argonne National Lab., IL (United States); Iseli, M. [ITER Naka Joint Work Site, Naka-gun (Japan); Wu, C.H. [EFDA Close Support Unit, Garching (Germany)

    2001-07-01

    Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points, to withstand the high thermal loads expected during off-normal events, e.g., disruption, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., {proportional_to}350 g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R and D work is required to narrow the remaining uncertainties are also briefly discussed. (orig.)

  11. In-Vessel Tritium Retention and Removal in ITER-FEAT

    Science.gov (United States)

    Federici, G.; Brooks, J. N.; Iseli, M.; Wu, C. H.

    Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points, to withstand the high thermal loads expected during off-normal events, e.g., disruptions, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., ˜350 g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R&D work is required to narrow the remaining uncertainties are also briefly discussed.

  12. Power control method for load-frequency control operation in BWRs

    International Nuclear Information System (INIS)

    Ie, Shin-ichiroo; Ohgo, Yu-kiharu; Itou, Tetsuo; Shida, Tooichi

    1991-01-01

    The preliminary design of an advanced power control method for fast load-following [load frequency control (LFC)] maneuvers in a boiling water reactor (BWR) is described in this paper. Application of a multivariable control method using an optimal linear quadratic (LQ) regulator theory effectively improves control system performance when system variables have significant interactions such as in BWRs. The control problem, however, demands strict constraints on system variable from the standpoint of plant operation. These constraints require the control system to have a nonlinear property for better improvement. Therefore, the effectiveness of LQ control is limited by these constraints, because it is based on a linear model. A new method is needed to compensate for the nonlinear property. In this study, the authors propose a new method using fuzzy reasoning with LQ control to achieve nonlinear compensation

  13. A 2-megawatt load for testing high voltage DC power supplies

    International Nuclear Information System (INIS)

    Horan, D.; Kustom, R.; Ferguson, M.; Primdahl, K.

    1993-01-01

    A high power water-cooled resistive load, capable of dissipating 2 Megawatts at 95 kilovolts is being designed and built. The load utilizes wirewound resistor elements suspended inside insulating tubing contained within a pressure vessel which is supplied a continuous flow of deionized water for coolant. A sub-system of the load is composed of non-inductive resistor elements in an oil tank. Power tests conducted on various resistor types indicate that dissipation levels as high as 22 times the rated dissipation in air can be achieved when the resistors are placed in a turbulent water flow of at least 15 gallons per minute. Using this data, the load was designed using 100 resistor elements in a series arrangement. A single-wall 316 stainless steel pressure vessel with flanged torispherical heads is built to contain the resistor assembly and deionized water. The resistors are suspended within G-11 tubing which span the cylindrical length of the vessel. These tubes are supported by G-10 baffles which also increase convection from the tubes by promoting turbulence within the surrounding water

  14. Evaluation of HFIR vessel surveillance data and hydro-test conditions

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Nanstad, R.K.

    1994-01-01

    Surveillance specimens for the High Flux Isotope Reactor (HFIR) pressure vessel were removed and tested during 1993, after the vessel had accumulated 701,469 MWd of operation. The data agree well with HFIR surveillance data obtained in previous years. In conjunction with this effort, the vessel hydro-test conditions were reevaluated and found to be more than adequate. In view of this result, and because there are economic incentives for reducing the frequency of hydro testing, an analysis was performed to determine the minimum permissible frequency. The value obtained is substantially less than that presently specified. It was also determined that a somewhat lower cooling-tower-basin temperature is acceptable (improves operational flexibility). In 1986, after ∼20 years of reactor operation, it was discovered that the vessel embrittlement rate was substantially greater than expected. Possible reasons for the accelerated rate are reviewed in this report

  15. Interpretation of strain measurements on nuclear pressure vessels

    International Nuclear Information System (INIS)

    Andersen, S.I.; Engbaek, P.

    1979-11-01

    Selected results from strain measurements on 4 nuclear pressure vessels are presented and discussed. The measurements were made in several different regions of the vessels: transition zones in vessel heads, flanges and bottom parts, nozzels, internal vessel structure and flange bolts. The results presented are based on data obtained by approximately 700 strain-gauges, and a comprehensive knowledge of the quality obtained by such measurements is established. It is shown that a thorough control procedure before and after the test as well as detailed knowledge of the behaviour of the signal from the individual gauges during the test is necessary. If this is omitted, it can be extremely difficult to distinguish between the real structural behaviour and a malfunctioning of a specific gauge installation. In general, most of the measuring results exhibit a very linear behaviour with a negligible zeroshift. However, deviations from linear behaviour are observed in several cases. This nonlinearity can be explained by friction (flange connections) or by gaps (concentrical nozzles) in certain regions, whereas local plastic deformations during the first pressure loadings of the vessel seem to be the reason in other regions. (author)

  16. A powerful methodology for reactor vessel pressurized thermal shock analysis

    International Nuclear Information System (INIS)

    Boucau, J.; Mager, T.

    1994-01-01

    The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). More specifically, the review of the old WWER-type of reactors (WWER 440/230) has indicated a sensitive behaviour to neutron embrittlement. This led already to some remedial actions including safety injection water preheating or vessel annealing. Such measures are usually taken based on the analysis of a selected number of conservative PTS events. Consideration of all postulated cooldown events would draw attention to the impact of operator action and control system effects on reactor vessel PTS. Westinghouse has developed a methodology which couples event sequence analysis with probabilistic fracture mechanics analyses, to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation. Once the event sequences of concern are identified, detailed deterministic thermal-hydraulic and structural evaluations can be performed to determine the conditions required to minimize the extension of postulated flaws or enhance flaw arrest in the reactor vessel. The results of these analyses can then be used to better define further modifications in vessel and plant system design and to operating procedures. The purpose of the present paper will be to describe this methodology and to show its benefits for decision making. (author). 1 ref., 3 figs

  17. Determination of the wind power systems load to achieve operation in the maximum energy area

    Science.gov (United States)

    Chioncel, C. P.; Tirian, G. O.; Spunei, E.; Gillich, N.

    2018-01-01

    This paper analyses the operation of the wind turbine, WT, in the maximum power point, MPP, by linking the load of the Permanent Magnet Synchronous Generator, PMSG, with the wind speed value. The load control methods at wind power systems aiming an optimum performance in terms of energy are based on the fact that the energy captured by the wind turbine significantly depends on the mechanical angular speed of the wind turbine. The presented control method consists in determining the optimal mechanical angular speed, ωOPTIM, using an auxiliary low power wind turbine, WTAUX, operating without load, at maximum angular velocity, ωMAX. The method relies on the fact that the ratio ωOPTIM/ωMAX has a constant value for a given wind turbine and does not depend on the time variation of the wind speed values.

  18. Fluid-solid contact vessel having fluid distributors therein

    Science.gov (United States)

    Jones, Jr., John B.

    1980-09-09

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

  19. Intraabdominal laparoscopy-assisted "open" vessel ligation of testicular vessels: a potential treatment for varicocele.

    Science.gov (United States)

    Miyano, Go; Miyahara, Katsumi; Halibieke, Abudebieke; Lane, Geoffrey J; Okazaki, Tadaharu; Yamataka, Atsuyuki

    2011-10-01

    We tested our laparoscopy-assisted "open" ligation (LOL) technique on testicular vessels. We ligated the left testicular artery and vein (TAV) in 8-week-old male Wister rats using LOL (LOL group; n=10) or laparotomy (open group; n=10). In LOL, a 0-degree laparoscope was introduced through a 5-mm epigastric trocar. A 3-mm grasper was used to expose the left TAV. A lapa-her-closure (LHC) needle loaded with 3-0 SurgiPro was directly inserted into the left lower quadrant where the left TAV should be and advanced under the vessels, and the suture material was released leaving one end outside. The LHC was then withdrawn a little and advanced again over the vessels to grasp the end of the suture material just released to bring it outside. This was proximally repeated. The two ends of both sutures were conventionally tied outside, and the knot was passed through the insertion site and tightened around the vessels. In the open group, the left TAV were ligated using two 3-0 SurgiPro ties. In both groups, the right side was left intact. All rats were sacrificed 2 weeks postoperatively, and both testes were examined with hematoxylin and eosin. Treatment time was 5-7 minutes for LOL and 7-8 minutes for the open group. Postoperative recovery was uneventful. No adhesions were present between the ligated vessels and bowel in any rat. Histopathology of all left testes showed coagulative necrosis of germinal cells and seminiferous tubules; all right testes were normal. LOL appears to be as effective as open ligation and may find application for treating varicocele.

  20. TMI-2 reactor vessel head removal

    International Nuclear Information System (INIS)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1984-12-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training the head was safely removed and stored and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

  1. TMI-2 reactor vessel head removal

    International Nuclear Information System (INIS)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1985-09-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

  2. Mobile loading transuranic waste at small quantity sites in the Department of Energy complex-10523

    International Nuclear Information System (INIS)

    Carter, Mitch; Howard, Bryan; Weyerman, Wade; Mctaggart, Jerri

    2009-01-01

    Los Alamos National Laboratory, Carlsbad Office (LANL-CO), operates mobile loading operations for all of the large and small quantity transuranic (TRU) waste sites in the Department of Energy (DOE) complex. The mobile loading team performs loading and unloading evolutions for both contact handled (CH) and remote handled (RH) waste. For small quantity sites, many of which have yet to remove their TRU waste, the mobile loading team will load shipments that will ship to Idaho National Laboratory, a centralization site, or ship directly to the Waste Isolation Pilot Plant (WIPP). For example, Argonne National Laboratory and General Electric Vallecitos Nuclear Center have certified programs for RH waste so they will ship their RH waste directly to WIPP. Many of the other sites will ship their waste to Idaho for characterization and certification. The Mobile Loading Units (MLU) contain all of the necessary equipment needed to load CH and RH waste into the appropriate shipping vessels. Sites are required to provide additional equipment, such as cranes, fork trucks, and office space. The sites are also required to provide personnel to assist in the shipping operations. Each site requires a site visit from the mobile loading team to ensure that all of the necessary site equipment, site requirements and space for shipping can be provided. The mobile loading team works diligently with site representatives to ensure that all safety and regulatory requirements are met. Once the waste is ready and shipping needs are met, the mobile loading team can be scheduled to ship the waste. The CH MLU is designed to support TRUPACT-II and HalfPACT loading activities wherever needed within the DOE complex. The team that performs the mobile loading operation has obtained national certification under DOE for TRUPACT-II and HalfPACT loading and shipment certification. The RH MLU is designed to support removable lid canister (RLC) and RH-72B cask loading activities wherever needed within the DOE

  3. The Bottom supported fast breeder reactor vessel - an alternative approach to seismic accommodation and reduced cost

    International Nuclear Information System (INIS)

    Nakagawa, H.; Golan, S.; Petrozelli, J.; Kumaoka, Y.; Kawamura, Y.

    1988-01-01

    Most FBR vessels are supported by hanging from their top portions. A disadvantage of such a top supported reactor vessel (TSRV) structural configuration is that it may generate high reactor core accelerations. This is due to the long path the seismic vibrations must travel from the basemat up through the building and then down through the RV block to the core. To compensate for this disadvantage, TSRV blocks are often strengthened beyond what is required for other considerations, such as pressure, to satisfy seismic response criteria, thus increasing weights and costs. In addition to long load paths, TSRVs also have common load paths. For example, in a TSRV (with the core supported from the bottom of the RV) the sodium and core loads both travel along the RV pressure boundary. Therefore, one of these loads will likely control the RV thickness leaving excess margin for the other loads. It is the premise of this paper that the revision of a large pool FBR from a TSRV configuration to a specific bottom supported reactor vessel (BSRV) configuration can resolve the above TSRV disadvantages related to load path length and diversity, thereby improving seismic performance and simultaneously reducing RV block costs by reducing weights. This paper demonstrates this premise by comparing a reference TSRV block with a specific BSRV block design

  4. 36 CFR 13.1180 - Closed waters, motor vessels and seaplanes.

    Science.gov (United States)

    2010-07-01

    ... 36 Parks, Forests, and Public Property 1 2010-07-01 2010-07-01 false Closed waters, motor vessels... and Preserve Vessel Operating Restrictions § 13.1180 Closed waters, motor vessels and seaplanes. (a... Hugh Miller Inlet. (4) Waters within the Beardslee Island group (except the Beardslee Entrance), that...

  5. Final report for the 2nd Ex-Vessel Neutron Dosimetry Installations and Evaluations for Yonggwang Unit 2 Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Gong, Un Sik; Choi, Kwon Jae; Chung, Kyoung Ki; Kim, Kwan Hyun; Chang, Jong Hwa; Ha, Jea Ju

    2008-01-15

    This report describes a neutron fluence assessment performed for the Yonggwang Unit 2 pressure vessel beltline region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the beltline region of the pressure vessel. During Cycle 16 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Yonggwang Unit 2 to provide continuous monitoring of the beltline region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 16.

  6. Final report for the 1st ex-vessel neutron dosimetry installations and evaluations for Kori unit 2 reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

    2006-11-15

    This report describes a neutron fluence assessment performed for the Kori unit 2 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 20 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 2 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 20.

  7. Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Yonggwang Unit 1 Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Li, Nam Jin; Hong, Joon Wha

    2007-01-15

    This report describes a neutron fluence assessment performed for the Yonggwang Unit 1 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During Cycle 16 of reactor operation, 2nd Ex-Vessel Neutron Dosimetry Program was instituted at Yonggwang Unit 1 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 16.

  8. Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Yonggwang Unit 1 Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Gong, Un Sik; Choi, Kwon Jae; Chung, Kyoung Ki; Kim, Kwan Hyun; Chang, Jong Hwa; Ha, Jea Ju

    2008-01-15

    This report describes a neutron fluence assessment performed for the Kori Unit 2 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During Cycle 21 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 2 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 21.

  9. Final report for the 1st ex-vessel neutron dosimetry installation and evaluations for Kori unit 4 reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

    2006-11-15

    This report describes a neutron fluence assessment performed for the Kori unit 4 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 16 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 4 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 16.

  10. Optimization for steady-state and hybrid operations of ITER by using scaling models of divertor heat load

    International Nuclear Information System (INIS)

    Murakami, Yoshiki; Itami, Kiyoshi; Sugihara, Masayoshi; Fujieda, Hirobumi.

    1992-09-01

    Steady-state and hybrid mode operations of ITER are investigated by 0-D power balance calculations assuming no radiation and charge-exchange cooling in divertor region. Operation points are optimized with respect to divertor heat load which must be reduced to the level of ignition mode (∼5 MW/m 2 ). Dependence of the divertor heat load on the variety of the models, i.e., constant-χ model, Bohm-type-χ model and JT-60U empirical scaling model, is also discussed. The divertor heat load increases linearly with the fusion power (P FUS ) in all models. The possible highest fusion power much differs for each model with an allowable divertor heat load. The heat load evaluated by constant-χ model is, for example, about 1.8 times larger than that by Bohm-type-χ model at P FUS = 750 MW. Effect of reduction of the helium accumulation, improvements of the confinement capability and the current-drive efficiency are also investigated aiming at lowering the divertor heat load. It is found that NBI power should be larger than about 60 MW to obtain a burn time longer than 2000 s. The optimized operation point, where the minimum divertor heat load is achieved, does not depend on the model and is the point with the minimum-P FUS and the maximum-P NBI . When P FUS = 690 MW and P NBI = 110 MW, the divertor heat load can be reduced to the level of ignition mode without impurity seeding if H = 2.2 is achieved. Controllability of the current-profile is also discussed. (J.P.N.)

  11. Performance of core modifications to reduce the reactor pressure vessel fluence

    International Nuclear Information System (INIS)

    Kiehlmann, H.D.; Lisdat, R.; Sommer, D.

    1997-01-01

    It's often discussed that nuclear power plants (NPP) are designed for an operation of 40 years equivalent to 32 full power years (FPY) assuming a load factor of 0.8. Such fixed plant life times are subjects of US operating licenses but not, as in most other countries, in the Federal Republic of Germany. Here the operating licenses are issued for an indefinite period. However, the German utilities are continuously upgrading their plants to attain a safety level that meets all current requirements. These upgrading measures also include the replacement of bigger components like e.g. the steam generator. The reactor pressure vessel (RPV), however, has a special status. Unlike most other components of a NPP which most likely will be exchanged during its service life a replacement or annealing treatment of the RPV certainly require more efforts to be economically justified. Thus the embrittlement of the RPV has an essential impact on the life time of a NPP. The end-of-life (EOL) RPV material toughness in essential depends on the steel quality and the accumulated neutron fluence. For a given NPP the reduction of the neutron flux at the inner surface of the RPV is the only way to limit its embrittlement. The resulting modifications for the core loadings in combination with the insertion of additional core components like steel elements are described and the impact on core performance and RPV fluence considered. (UK)

  12. Effect of a new specimen size on fatigue crack growth behavior in thick-walled pressure vessels

    International Nuclear Information System (INIS)

    Shariati, Mahmoud; Mohammadi, Ehsan; Masoudi Nejad, Reza

    2017-01-01

    Fatigue crack growth in thick-walled pressure vessels is an important factor affecting their fracture. Predicting the path of fatigue crack growth in a pressure vessel is the main issue discussed in fracture mechanics. The objective of this paper is to design a new geometrical specimen in fatigue to define the behavior of semi-elliptical crack growth in thick-walled pressure vessels. In the present work, the importance of the behavior of fatigue crack in test specimen and real conditions in thick-walled pressure vessels is investigated. The results of fatigue loading on the new specimen are compared with the results of fatigue loading in a cylindrical pressure vessel and a standard specimen. Numerical and experimental methods are used to investigate the behavior of fatigue crack growth in the new specimen. For this purpose, a three-dimensional boundary element method is used for fatigue crack growth under stress field. The modified Paris model is used to estimate fatigue crack growth rates. In order to verify the numerical results, fatigue test is carried out on a couple of specimens with a new geometry made of ck45. A comparison between experimental and numerical results has shown good agreement. - Highlights: • This paper provides a new specimen to define the behavior of fatigue crack growth. • We estimate the behavior of fatigue crack growth in specimen and pressure vessel. • A 3D finite element model has been applied to estimate the fatigue life. • We compare the results of fatigue loading for cylindrical vessel and specimens. • Comparison between experimental and numerical results has shown a good agreement.

  13. Diagnostics carried by a light multipurpose deployer for vacuum vessel interventions

    Energy Technology Data Exchange (ETDEWEB)

    Houry, M., E-mail: Michael.houry@cea.fr [CEA-IRFM, F-13108 Saint-Paul-Lez-Durance (France); Gargiulo, L.; Balorin, C.; Bruno, V.; Keller, D.; Roche, H. [CEA-IRFM, F-13108 Saint-Paul-Lez-Durance (France); Kammerer, N.; Measson, Y. [CEA, LIST, F-92265 Fontenay-aux-Roses (France); Carrel, F.; Schoepff, V. [CEA, LIST, F-91191 Gif-sur-Yvette (France)

    2011-10-15

    ITER will greatly rely on remote-handling operations to accomplish its scientific missions. Robotic systems will also be required to operate inside vacuum vessels in order to limit or replace human access, to intervene quickly between experimental sessions for in-vessel inspections and measurements, and to preserve the machine conditioning and thus improve machine availability. In this prospect, a multipurpose carrier prototype called Articulated Inspection Arm (AIA) was developed by CEA laboratories within the European work program. With an embedded camera, it successfully demonstrated close inspection feasibility inside Tore Supra tokamak. The AIA robot was designed for mini-invasive operations with interchangeable diagnostics to be plugged at its head. This covers various applications for the safety, the operation and the scientific mission (in-vessel inspection, plasma diagnostics calibrations or inner components analysis and treatments). This paper presents recent analysis and results obtain with diagnostics developed by CEA for in-vessel remote-handling intervention.

  14. Diagnostics carried by a light multipurpose deployer for vacuum vessel interventions

    International Nuclear Information System (INIS)

    Houry, M.; Gargiulo, L.; Balorin, C.; Bruno, V.; Keller, D.; Roche, H.; Kammerer, N.; Measson, Y.; Carrel, F.; Schoepff, V.

    2011-01-01

    ITER will greatly rely on remote-handling operations to accomplish its scientific missions. Robotic systems will also be required to operate inside vacuum vessels in order to limit or replace human access, to intervene quickly between experimental sessions for in-vessel inspections and measurements, and to preserve the machine conditioning and thus improve machine availability. In this prospect, a multipurpose carrier prototype called Articulated Inspection Arm (AIA) was developed by CEA laboratories within the European work program. With an embedded camera, it successfully demonstrated close inspection feasibility inside Tore Supra tokamak. The AIA robot was designed for mini-invasive operations with interchangeable diagnostics to be plugged at its head. This covers various applications for the safety, the operation and the scientific mission (in-vessel inspection, plasma diagnostics calibrations or inner components analysis and treatments). This paper presents recent analysis and results obtain with diagnostics developed by CEA for in-vessel remote-handling intervention.

  15. Operational transparency: an advanced safeguards strategy for future on-load refuelled reactors

    International Nuclear Information System (INIS)

    Whitlock, J.J.; Trask, D.

    2012-01-01

    The IAEA's system for tracking fuel movement in an on-load refuelled heavy-water reactor is robust, but an opportunity remains to exploit the wealth of data streaming from the reactor vault during operation and provide real-time, third-party monitoring of reactor status and history. This concept of Operational Transparency would require that large amounts of operational data be reduced in near-real time to a small subset of high-level information. Operational Transparency would enhance the IAEA's ability to monitor the state of the core to an unprecedented level. This paper provides an overview of the novel concept of Operational Transparency in heavy water reactors, using potential application to CANDU reactors as an example, and explores some of the technical challenges that will need to be solved for efficient implementation. (author)

  16. A loading/unloading buoy; Laste/losseboeye

    Energy Technology Data Exchange (ETDEWEB)

    Breivik, K.; Smedal, A.; Syvertsen, K.

    1994-10-10

    The invention relates to a buoy design for use in the offshore loading or unloading of crude oil in particular. The buoy comprises an outer buoyancy member arranged to be introduced and secured in a submerged downwardly open receiving space in a floating vessel, and a central member which is rotatably mounted in the outer member and is intended for anchoring to the sea bed and arranged for passage of medium between a transfer line which, in operation, is coupled to the lower end of the central member and a tube system on the vessel. The central member is provided with a lower extension body having an outer peripheral portion abutting on and essentially corresponding to the outer periphery of the adjacent end of the outer buoyancy member, and having a lower portion which is downwardly tapering from the outer peripheral portion. A number of fastening means for fastening of the upper ends of anchoring lines for anchoring of the buoy are fastened at intervals along the periphery of the outer peripheral portion of the extension body, and the extension body comprises at least one buoyancy chamber for buoyancy or ballast material. 6 figs.

  17. The inclusion of weld residual stress in fracture margin assessments of embrittled nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Dickson, T.L.; Bass, B.R.; McAfee, W.J.

    1998-01-01

    Analyses were performed to determine the impact of weld residual stresses in a reactor pressure vessel (RPV) on (1) the generation of pressure temperature (P-T) curves required for maintaining specified fracture prevention margins during nuclear plant startup and shutdown, and (2) the conditional probability of vessel failure due to pressurized thermal shock (PTS) loading. The through wall residual stress distribution in an axially oriented weld was derived using measurements taken from a shell segment of a canceled RPV and finite element thermal stress analyses. The P-T curve derived from the best estimate load analysis and a t / 8 deep flaw, based on K Ic , was less limiting than the one derived from the current methodology prescribed in the ASME Boiler and Pressure Vessel Code. The inclusion of the weld residual stresses increased the conditional probability of cleavage fracture due to PTS loading by a factor ranging from 2 to 4

  18. Exact solution for stresses/displacements in a multilayered hollow cylinder under thermo-mechanical loading

    International Nuclear Information System (INIS)

    Yeo, W.H.; Purbolaksono, J.; Aliabadi, M.H.; Ramesh, S.; Liew, H.L.

    2017-01-01

    In this study, a new analytical solution by the recursive method for evaluating stresses/displacements in multilayered hollow cylinder under thermo-mechanical loading was developed. The results for temperature distribution, displacements and stresses obtained by using the proposed solution were shown to be in good agreement with the FEM results. The proposed analytical solution was also found to produce more accurate results than those by the analytical solution reported in literature. - Highlights: • A new analytical solution for evaluating stresses in multilayered hollow cylinder under thermo-mechanical loading. • A simple computational procedure using a recursive method. • A promising technique for evaluating the operating axial and hoop stresses in pressurized composite vessels.

  19. The LG-bank control concept: An improved method for PWR load-following operation

    International Nuclear Information System (INIS)

    Park, Won Seok; Christenson, J.M.

    1990-01-01

    In this paper the authors present the results of an investigation of a new pressurized water reactor load-following control concept that utilizes light gray (LG) banks in combination with a single high-worth bank. The investigation determined a control strategy and a set of nuclear design parameters for the control banks that permits unrestricted load-following operation over a wide power range at both beginning-of-cycle and end-of-cycle conditions. Advantages of the LG-bank control concept are that flexible load-following maneuvers can be performed without either making changes in the boron concentration or requiring the continuous insertion of a high-worth control bank. These features remove both of the disadvantages of current gray-bank load-following designs, which generally require the continuous insertion of a high-worth bank and in some cases also involve changes in the boron concentration

  20. TPX vacuum vessel transient thermal and stress conditions

    International Nuclear Information System (INIS)

    Feldshteyn, Y.; Dinkevich, S.; Feng, T.; Majumder, D.

    1995-01-01

    The TPX vacuum vessel provides the vacuum boundary for the plasma and the mechanical support for the internal components. Another function of the vacuum vessel is to contain neutron shielding water in the double wall space during normal operation. This double wall space serves as a heat reservoir for the entire vacuum vessel during bakeout. The vacuum vessel and the internal components are subjected to thermal stresses induced by a nonuniform temperature distribution within the structure during bakeout. A successful Conceptual Design Review in March 1993 has established superheated steam as the heating source of the vacuum vessel. A transient bakeout mode of the vacuum vessel and in-vessel components has been analyzed to evaluate transient period duration, proper temperature level, actual thermal stresses and performance of the steam equipment. Thermally, the vacuum vessel structure may be considered as an adiabatic system because it is perfectly insulated by the strong surrounding vacuum and multiple layers of superinsulation. Important aspects of the analysis are described herein

  1. Operating experience with high beam currents and transient beam loading in the SLC damping rings

    International Nuclear Information System (INIS)

    Minty, M.G.; Akre, R.; Krejcik, P.; Siemann, R.H.

    1995-01-01

    During the 1994 SLC run the nominal operating intensity in the damping rings was raised from 3.5 x 10 10 to greater than 4 x 10 10 particles per bunch (ppb). Stricter regulation of rf system parameters was required to maintain stability of the rf system and particle beam. Improvements were made in the feedback loops which control the cavity amplitude and loading angles. Compensation for beam loading was also required to prevent klystron saturation during repetition rate changes. To minimize the effects of transient loading on the rf system, the gain of the direct rf feedback loop and the loading angles were optimized

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  4. Experimental Investigation of Creep Behavior of Reactor Vessel Lower Head

    International Nuclear Information System (INIS)

    Chu, T.Y.; Pilch, M.; Bentz, J.H.; Behbahani, A.

    1999-01-01

    The authors report a study which aimed at experimentally and numerically investigating and characterizing the failure of a reactor pressure vessel (RPV) lower head due to thermal and pressure loads generated by a severe accident. They present the experimental apparatus which is based on a scaled version of the lower part of a TMI-like reactor pressure vessel without vessel skirt. They report and comment the results obtained during the first five experiments: uniform heating and non penetrations, centre-peaked heat flux and no penetrations, edge-peaked heat flux and no penetrations, uniform heating with penetrations, edge-peaked heat flux with penetrations. They compare the third and fifth experience (those with edge-peaked heat flux)

  5. Probabilistic fracture mechanics analysis of reactor vessels with low upper-shelf fracture toughness

    International Nuclear Information System (INIS)

    Yoon, K.K.

    1993-01-01

    A class of submerged-arc welds used in fabricating early reactor vessels has relatively high copper contents. Studies have shown that when such vessels are irradiated, the copper contributes to lowering the Charpy upper-shelf energy level. To address this concern, 10CFR50, Appendix G requires a fracture mechanics analysis to demonstrate an adequate margin of safety for continued service. The B and W Owners Group (B and WOG) has been accumulating J-resistance fracture toughness data for these weld metals. Based on a mathematical model derived from this B and WOG data base, the first Appendix G analysis was performed. Another important issue affecting reactor vessel integrity is pressurized thermal shock (PIS) transients. In the early 1980s, probabilistic fracture mechanics analyses were performed on a reactor vessel to determine the probability of failure under postulated accident scenarios. Results of such analyses were used by the Nuclear Regulatory Commission (NRC) to establish the screening criteria for assessing reactor vessel integrity under PTS transient loads. This paper addresses the effect of low upper-shelf toughness on the probability of failure of reactor vessels under PTS loads. Probabilistic fracture mechanics codes were modified to include the low upper-shelf toughness model used in a reference and a series of analyses was performed using plant-specific material conditions and realistic PTS scenarios. The results indicate that low upper-shelf toughness has an insignificant effect on the probability of reactor vessel failures. This is mostly due to PTS transients being susceptible to crack initiation at low temperatures and not affected by upper-shelf fracture toughness

  6. Efficient scanning of thick lead vessels

    International Nuclear Information System (INIS)

    Raghunath, V.M.; Bhatnagar, P.K.; Meenakshisundaram, V.

    1978-01-01

    Lead containers fabricated for transport of radioactive materials need to be evaluated for their shielding integrity. The common method of locating a strong gamma source inside the vessel and scanning the external surface by conventional detectors suffers from high radiation dose and low sensitivity. A new method has been proposed and tried. It is found to be more efficient. In the new method, 60 Co source is loaded at the centre of the lead vessel and the outer surface is scanned by NaI(Tl) detector. The transmitted virgin flux is scanned under the 60 Co channel in a single channel analyser. An area of 25 cm 2 is scanned for 10 to 20 seconds each time. The source strength required is considerably reduced by a factor of 10 or more as compared to the common method and external dose rates do not exceed 50 mR/h (130 nC. kg -1 h -1 ) on the vessel surface. The advantages are improved sensitivity, no interference from scattered radiation and assurance in repeatability of measurements. (M.G.B.)

  7. Demonstration of fuel switching on oceangoing vessels in the Gulf of Mexico.

    Science.gov (United States)

    Browning, Louis; Hartley, Seth; Bandemehr, Angela; Gathright, Kenneth; Miller, Wayne

    2012-09-01

    Switching fuels from high-sulfur heavy fuel oils (HFO) to lower sulfur marine gas oils (MGO) on an oceangoing vessel (OGV) can substantially reduce both PM and SO(x) ship stack emissions, potentially resulting in significant human health and environmental benefits in Gulf of Mexico port communities. The International Maritime Organization (IMO) established an emission control area (ECA) within 200 nautical miles of the US. and Canadian coastlines and French territories off the coast of Canada with lower fuel sulfur standards effective beginning August 2012, where OGVs will need to switch from HFO to MGO. However some operators and other stakeholders, particularly in the Gulf of Mexico, may be unfamiliar with the benefits and requirements and ship operators may be concerned over potential implications for cost and operations. This first-ever US. Environmental Protection Agency (EPA)-sponsored fuel switching demonstration in the Gulf of Mexico was initiated to showcase the environmental and health benefits of as well as operational issues associated with, fuel switching through the following activities: (1) Fuel switching was conducted on typical container ships operating the Gulf of Mexico, as routine fuel switching has been demonstrated in California in recent years. Two vessels were employed in the demonstration: the Maersk Roubaix, which switched fuels entering Port of Houston, TX, and the Port of Progreso, Mexico, and the Hamburg Süd vessel Cap San Lorenzo, which switched fuels entering the Port of Houston and the Mexican Ports of Veracruz and Altamira. Operational and cost aspects were also noted. (2) Emissions reductions were quantified through both a calculation approach based on fuel use of the Maersk Line vessel Roubaix and in-stack monitoring of emissions from the Hamburg Süd Cap San Lorenzo. Pollutant emissions including PM, SO(x), NO(x), and PM component speciation were sampled during use of both fuels. These observations showed reductions (1-6%) in NO

  8. Sealing analysis for nuclear vessel of PWR

    International Nuclear Information System (INIS)

    Qu, J.; Dou, Y.

    1987-01-01

    Although design by analysis of pressure vessel has become a requirement in all codes for more than 20 years, sealing design for nuclear components is still too complicated and there are yet no criteria about this aspect, even though in the well-known ASME Boiler and Pressure Vessel Code. Thus it is of significance to undertake researches of transient sealing tests and analysis for nuclear vessel. Since 1960s great progress has been made in analytic computer program, which takes flange as a rigid ring. Actually, however, there are elastic or elastoplastic contacts on flange mating surface. Chen (1979) gave a mixed finite element method, using a condensing flexible matrix skill, to solve two-body contact problem. On the basis of axisymmetric stress and thermal analysis of finite element method and on accepting Chen's (1979) idea of mixed finite element method, we have developed a computer program for sealing analysis, named SMEC, which considers bolt loading changes and temperature effects. (orig./GL)

  9. A Study on Load Following Operation Strategy to Minimize Boric-acid Adjustment

    International Nuclear Information System (INIS)

    Yoo, Jae Woon

    1997-02-01

    Applicability of the MSHIM control strategy, which has been developed by Westinghouse and characterized by elimination of the adjustment to the boron concentration during load maneuvering, to 1300MWe KNGR Cycle-1 is examined by simulation based on one-dimensional transient analyses code, ONED94, under the conditions of relatively large xenon worth and high amplitude of xenon oscillation. CASMO-3/MASTER code system is used to generate the ONED94 model. A processing code, which is referred to as MAS2ONED, is also developed to transform the MASTER calculation results to ONED94 cross section library. The three types of power maneuvering, daily maneuvering, extended weekend load follow, and spinning reserve capacity are simulated to evaluate the capability of the selected load follow technology, by using ONED94 code. By showing that the calculated total peaking factor is maintained below safety limit, it is proved that load follow operation without boron adjustment is feasible with simple control rod motion only. It is expected that the results of this study, such as control rod parameters could give a contribution to the development of load following technology in our country. The developed process code could be utilized in the another simulation of power maneuver in the future

  10. The JET high temperature in-vessel inspection system

    International Nuclear Information System (INIS)

    Businaro, T.; Cusack, R.; Calbiati, L.; Raimondi, T.

    1989-01-01

    The JET In-vessel Inspection System (IVIS) has been enhanced for operation under the following nominal conditions: vacuum vessel at 350 degC; vacuum vessel evacuated (∼10 -9 mbar); radiation dose during D-T phase 10 rads. The target resolution of the pictures is 2 mm at 5 m distance and tests on radiation resistance of the IVIS system are being carried out. Since June 1988, the new system is installed in the JET machine and the first inspections of the intire vessel at 250 degC have been satisfactory done. (author). 3 refs.; 6 figs.; 1 tab

  11. Diagnostic value for coronary artery disease of ST depression and chest pain during dipyridamole loading

    International Nuclear Information System (INIS)

    Yonezawa, Yoshihiro; Doi, Yoshinori; Aono, Tadashi; Odawara, Hiroaki; Chikamori, Taishiro; Yamada, Mitsutoshi; Takata, Jun; Ozawa, Toshio; Hamashige, Naohisa.

    1991-01-01

    The diagnostic significance of both ST depression and chest pain during dipyridamole loading was assessed in 437 patients with coronary artery disease (CRD) who have received dipyridamole-loading thallium-201 myocardial scintigraphy and coronary arteriography. ST depression and chest pain were induced in 35% and 42%, respectively. ST depression occurred in 23% for one vessel disease, 55% for two-vessel disease, 67% for three-vessel disease, and 11% for non-significant disease. In the absence of myocardial infarction, it occurred in as many as 78% for two-vessel disease and 94% for three-vessel disease. ST depression had a sensitivity of 51% and a specificity of 89%. Chest pain had a lower specificity (77%), whereas the sensitivity remained the same (51%). In 42 (31%) of 134 patients with ST depression, coronary revascularization was necessary in the early stage. With a median follow up of 29 months, 3 patients (2%) died of cardiac events, and 12 (9%) had nonfatal cardiac complications. There was a significant correlation between reversible defects (RD) on myocardial scintigrams and ST depression; RD were seen in 81% of ST depression cases, and ST depression occurred in 51% of 210 RD cases. The more diseased vessels, the more frequently ST depression occurred in accordance with RD on myocardial scintigrams. Both ST depression and chest pain during dipyridamole loading tended to be associated with myocardial ischemia, suggesting the diagnostic value in CRD patients with limited exercise loading. RD on myocardial scintigrams was considered attributable to coronary steal phenomenon for multi-vessel disease and to the difference in the relative increase of the coronary flow for single vessel disease. (N.K.)

  12. The pressure vessel for the NSF tandem

    International Nuclear Information System (INIS)

    Jones, C.W.

    1979-04-01

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

  13. Fission gas release behavior of MOX fuels under simulated daily-load-follow operation condition. IFA-554/555 test evaluation with FASTGRASS code

    International Nuclear Information System (INIS)

    Ikusawa, Yoshihisa; Ozawa, Takayuki

    2008-03-01

    IFA-554/555 load-follow tests were performed in HALDEN reactor (HBWR) to study the MOX fuel behavior under the daily-load-follow operation condition in the framework of ATR-MOX fuel development in JAEA. IFA-554/555 rig had the instruments of rod inner pressure, fuel center temperature, fuel stack elongation, and cladding elongation. Although the daily-load-follow operation in nuclear power plant is one of the available options for economical improvement, the power change in a short period in this operation causes the change of thermal and mechanical irradiation conditions. In this report, FP gas release behavior of MOX fuel rod was evaluated under the daily-load-follow operation condition with the examination data from IFA-554/555 by using the computation code 'FASTGRASS'. From the computation results of FASTGRASS code which could compute the FP gas release behavior under the transient condition, it could be concluded that FP gas was released due to the relaxation of fuel pellet inner stress and pellet temperature increase, which were caused by the cyclic power change during the daily-load-follow operation. In addition, since the amount of released FP gas decreased during the steady operation after the daily-load-follow, it could be mentioned that the total of FP gas release at the end of life with the daily-load-follow is not so much different from that without the daily-load-follow. (author)

  14. To selecting the characteristics of saturated steam direct cycle NPPs for operation under variable loads

    International Nuclear Information System (INIS)

    Khrustalev, V.A.; Demidov, O.I.

    1986-01-01

    Problems for operating process optimization of NPPs with RBMK type reactors under load swings in the power system is considered. Determination technique for optimal values of such parameters as initial steam pressure and fuel enrichment for NPP different load factors is developed. Optimization of these parameters gives a 150000 rouble saving of annual expenditures per each 3200 MW of reactor heat output

  15. Structural Loading of Cross Deck Connections for Trimaran Vessels

    National Research Council Canada - National Science Library

    Rhoads, Jason

    2004-01-01

    ...: longitudinal bending, transverse bending, torsional bending, spreading and squeezing of hulls, inner and outer hull slam pressures, wet deck slam pressures, loading from ship's motions, and whipping of slender hulls...

  16. Neurocontrol of Pressurized Water Reactors in Load-Follow Operations

    International Nuclear Information System (INIS)

    Lin Chaung; Shen Chihming

    2000-01-01

    The neurocontrol technique was applied to control a pressurized water reactor (PWR) in load-follow operations. Generalized learning or direct inverse control architecture was adopted in which the neural network was trained off-line to learn the inverse model of the PWR. Two neural network controllers were designed: One provided control rod position, which controlled the axial power distribution, and the other provided the change in boron concentration, which adjusted core total power. An additional feedback controller was designed so that power tracking capability was improved. The time duration between control actions was 15 min; thus, the xenon effect is limited and can be neglected. Therefore, the xenon concentration was not considered as a controller input variable, which simplified controller design. Center target strategy and minimum boron strategy were used to operate the reactor, and the simulation results demonstrated the effectiveness and performance of the proposed controller

  17. Response Load Extrapolation for Wind Turbines during Operation Based on Average Conditional Exceedance Rates

    DEFF Research Database (Denmark)

    Toft, Henrik Stensgaard; Naess, Arvid; Saha, Nilanjan

    2011-01-01

    to cases where the Gumbel distribution is the appropriate asymptotic extreme value distribution. However, two extra parameters are introduced by which a more general and flexible class of extreme value distributions is obtained with the Gumbel distribution as a subclass. The general method is implemented...... within a hierarchical model where the variables that influence the loading are divided into ergodic variables and time-invariant non-ergodic variables. The presented method for statistical response load extrapolation was compared with the existing methods based on peak extrapolation for the blade out......The paper explores a recently developed method for statistical response load (load effect) extrapolation for application to extreme response of wind turbines during operation. The extrapolation method is based on average conditional exceedance rates and is in the present implementation restricted...

  18. Size of Left Cardiac Chambers Correlates with Cerebral Microembolic Load in Open Heart Operations

    Directory of Open Access Journals (Sweden)

    Elena Z. Golukhova

    2010-01-01

    Full Text Available Background. Microemboli are a widely recognized etiological factor of cerebral complications in cardiac surgery patients. The present study was aimed to determine if size of left cardiac chambers relates to cerebral microembolic load in open heart operations. Methods. Thirty patients participated in the study. Echocardiography was performed in 2-3 days before surgery. A transcranial Doppler system was used for registering intraoperative microemboli. Results. Preoperative left atrium and left ventricular end-systolic and end-diastolic sizes significantly correlated with intraoperative microembolic load (s=0.48, 0.57 and 0.53, s≺.01, resp.. The associations between left ventricular diameters and number of cerebral microemboli remained significant when cardiopulmonary bypass time was included as a covariate into the analysis. Conclusions. The present results demonstrate that increased size of left heart chambers is an influential risk factor for elevated cerebral microembolic load during open heart operations. Mini-invasive surgery and carbon dioxide insufflation into wound cavity may be considered as neuroprotective approaches in patients with high risk of cerebral microembolism.

  19. Demonstration of an instrumental technique in the measurement of solution weight in the accountability vessels of a fuel reprocessing plant

    International Nuclear Information System (INIS)

    Nakajima, K.

    1977-04-01

    Load cells were installed on the input accountability vessel of a commercial reactor fuel reprocessing facility to determine if this proven principle of mass measurement is in fact applicable in such a severe radiation environment over a long period of time. Two other locations selected were the plutonium product nitrate solution accountability vessel and the plutonium product nitrate solution storage vessel. The latter two environments, while not severely radio-active, require a high degree of contamination control. All three vessels are of different geometrical configuration and capacity. Each vessel was carefully calibrated for volume measurements by adding controlled pre-measured increments of water. Measurements were made using the conventional dip-tube manometer system and the load cell - digital voltmeter. Standard deviation of the measurements on the input vessel and the plutonium storage vessel were in both cases 0.3%; for the plutonium accountability vessel 1.9%. Measurements taken of the input vessel during the ''cold run'' over a six-month period using solutions of unirradiated uranium showed a standard deviation of 0.4% and a bias of 0.8% in the summer months and 0.7% and 0.6% respectively in the winter months FINAL STOP CODE

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

    Energy Technology Data Exchange (ETDEWEB)

    Cooke, Conrad; Spann, Holger [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)

    2013-07-01

    Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle