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Sample records for bwr pressure vessel

  1. Crack growth tests on a ferritic reactor pressure vessel steel under the simultaneous influence of simulated BWR coolant and irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, H. [VGB PowerTech e.V., Essen (Germany); Huettner, F. [Hamburgische Electricitaets-Werke AG, Hamburg (Germany); Ilg, U. [EnBW Kraftwerke AG, Philippsburg (Germany); Wachter, O. [E.ON Kernkraft GmbH, Hannover(Germany); Widera, M. [RWE Power AG, Essen (Germany); Brozova, A.; Ernestova, M.; Kysela, J.; Vsolak, R. [Nuclear Research Institute Rez plc (Czech Republic)

    2004-07-01

    Crack growth tests under constant load with initial in-situ cycling were performed on the low alloy reactor pressure vessel (RPV) steel 22 NiMoCr 3 7 (A 508 Cl. 2) with the goal to determine crack growth rates of irradiated and non-irradiated steel under the simultaneous influence of simulated BWR coolant and irradiation. The tests were performed under conditions as near as possible to operational conditions in a commercial BWR reactor. The research results are summarized and are compared with international data. (orig.)

  2. Test of 6-in. -thick pressure vessels. Series 3: intermediate test vessel V-7. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    Merkle, J.G.; Robinson, G.C.; Holz, P.P.; Smith, J.E.; Bryan, R.H.

    1976-08-01

    The test of intermediate test vessel V-7 was a crack-initiation fracture test of a 152-mm-thick (6-in.), 990-mm-OD (39-in.) vessel of ASTM A533, grade B, class 1 steel plate with a sharp outside surface flaw 457 mm (18 in.) long and about 135 mm (5.3 in.) deep. The vessel was heated to 91/sup 0/C (196/sup 0/F) and pressurized hydraulically until leakage through the flaw terminated the test at a peak pressure of 147 MPa (21,350 psi). Fracture toughness data obtained by testing precracked Charpy-V and compact-tension specimens machined from a prolongation of the cylindrical test shell were used in pretest analyses of the flawed vessel. The vessel, as expected, did not burst. Upon depressurization, the ruptured ligament closed so as to maintain static pressure without leakage at about 129 MPa (18,700 psi).

  3. Studies of fragileness in steels of vessels of BWR reactors; Estudios de fragilizacion en aceros de vasija de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA MARK lll reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A{sup 2}. (Author)

  4. Description of a disposition line on the stress corrosion cracking behaviour of ferritic reactor pressure vessel steels under BWR-conditions; Beschreibung einer einhuellenden Risswachstumskurve zum Spannungsrisskorrosionsverhalten von ferritischen Reaktordruckbehaelter (RDB)-Staehlen unter Siedewasserreaktor (SWR)-Bedingungen

    Energy Technology Data Exchange (ETDEWEB)

    Bruemmer, G. [HEW, Hamburg (Germany); Hoffmann, H. [VGB-GS, Essen (Germany); Ilg, U. [EnBW Kraftwerke AG, Philippsburg (Germany); Wachter, O. [E.ON-Kernkraft, Hannover (Germany); Widera, M. [RWE Power, Essen (Germany); Roth, A. [Framatome ANP GmbH, Erlangen (Germany)

    2002-07-01

    The inner surface of the reactor pressure vessel of BWR reactors is lined with a welded, corrosion-resistant steel liner. In an assumed case of liner rupture down to the low-alloy ferritic base material, an integrity assessment of the pressure vesssel in consideration of the effects of reactor coolant is of utmost importance, and research in this field has been going on for more than ten years now. An analysis of the available data shows that it is now possible to describe a disposition line on the stress corrosion cracking behaviour of ferritic reactor pressure vessel steels in BWR conditions. Crack growth rates of a stress intensity factor corresponding to a T/4 wall defect (i.e. 25 percent of the wall thickness) are technically not relevant. This scientific finding is supported by measurements of about 450 reactor operation years of all German LWR reactor plants, none of which showed crack initiation in the reactor pressure vessel. [German] Die mediumberuehrte Innenoberflaeche des Reaktordruckbehaelters (RDB) von Siedewasserreaktoren (SWR) ist mit einer korrosionsbestaendigen austenitischen Schweissplattierung versehen. Fuer den unterstellten Fall einer bis auf den niedriglegierten, ferritischen Grundwerkstoff durchgerissenen Pattierung ist fuer die Beurteilung der Integritaet des RDB unter Beruecksichtigung der Einwirkung des Reaktorkuehlmittels die Klaerung der Frage eines korrosionsgestuetzten Risswachstums von grosser Bedeutung. Dieses Thema ist daher bereits seit mehr als 10 Jahren Gegenstand umfangreicher Forschungsaktivitaeten. Ende der 80er- und Anfang der 90er-Jahre wurden fuer ferritische RDB-Staehle von SWR-Anlagen Risswachstumsgeschwindigkeiten veroeffentlicht, die binnen weniger als einem Jahr zum Durchriss der drucktragenden Wand eines RDB gefuehrt haetten. Daraufhin wurden internationale Forschungsaktivitaeten zur Ermittlung zuverlaessiger und reproduzierbarer Risswachstumsdaten initiiert, deren Ergebnisse zusammenfassend dargestellt werden. Die

  5. Analysis of the microstructural evolution of the damage by neutron irradiation in the pressure vessel of a nuclear power reactor BWR; Analisis de la evolucion microestructural del dano por irradiacion neutronica en la vasija de presion de un reactor nuclear de potencia BWR

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel y R, M.

    2012-07-01

    Nuclear reactor pressure vessel type BWR, installed in Mexico and in many other countries, are made of an alloy of low carbon steel. The American Society for Testing and Materials (Astm) classifies this alloy as A533-B, class 1. Both the vessel and other internal structures are continuously exposed to the neutron flux from the reactions of fission in nuclear fuel. A large number of neutrons reach the vessel and penetrate certain depth depending on their energy. Its penetration in the neutron collides with the nuclei of the atoms out of their positions in the crystal lattice of steel, producing vacancies, interstitial, segregations, among other defects, capable of affecting its mechanical properties. Analyze the micro-structural damage to the vessel due to neutron irradiation, is essential for reasons of integrity of this enclosure and safety of any nuclear power plant. The objective of this thesis work is theoretical and experimentally determine the microstructural damage of a type nuclear reactor vessel steel BWR, due to neutron radiation from the reactor core, using microscopic and spectroscopic techniques as well as Monte Carlo simulation. Microscopy Optical, Scanning Electron Microscopy, Transmission Electron Microscopy, Energy Dispersion of X-rays Spectrometry and X-rays Diffractometry were the techniques used in this research. These techniques helped in the characterization of both the basis of design of pressure vessel steel and steel irradiated, after eight years of neutron irradiation on the vessel, allowing know the surface morphology and crystal structures of the previous steel and post-irradiation, analyze the change in the microstructure of the steel vessel, morphological damage to surface level in an irradiated sample, among which are cavities in the order of microns produced by Atomic displacements due to the impact of neutronic, above all in the first layers of thickness of the vessel, the effect of swelling, regions of greater damage and Atomic

  6. BWR/5 Pressure-Suppression Pool Response during an SBO

    Directory of Open Access Journals (Sweden)

    Javier Ortiz-Villafuerte

    2013-01-01

    Full Text Available RELAP/SCDAPSIM Mod 3.4 has been used to simulate a station blackout occurring at a BWR/5 power station. Further, a simplified model of a wet well and dry well has been added to the NSSS model to study the response of the primary containment during the evolution of this accident. The initial event leading to severe accident was considered to be a LOOP with simultaneous scram. The results show that RCIC alone can keep the core fully covered, but even in this case about 30% of the original liquid water inventory in the PSP is vaporized. During the SBO, without RCIC, this inventory is reduced about 5% more within six hours. Further, a significant pressure rise occurs in containment at about the time when a sharp increase of heat generation occurs in RPV due to cladding oxidation. Failure temperature of fuel clad is also reached at this point. As the accident progresses, conditions for containment venting can be reached in about nine hours, although there still exists considerable margin before reaching containment design pressure. Detailed information of accident progress in reactor vessel and containment is presented and discussed.

  7. Damage by radiation in structural materials of BWR reactor vessels; Dano por radiacion en materiales estructurales de vasijas de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [Departamento de Sintesis y Caracterizacion de Materiales, Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2002-07-01

    The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA Mark III Salazar reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A{sup 2}). (Author)

  8. 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 ...

  9. Generic BWR-4 degraded core in-vessel study. Status report

    Energy Technology Data Exchange (ETDEWEB)

    1984-11-01

    Original intent of this project was to produce a phenomenological study of the in-vessel degradation which occurs during the TQUX and TQUV sequences for a generic BWR-4 from the initiation of the FSAR Chapter 15 operational transient through core debris bed formation to the failure of the primary pressure boundary. Bounding calculations were to be performed for the two high pressure and low pressure non-LOCA scenarios to assess the uncertainties in the current state of knowledge regarding the source terms for containment integrity studies. Source terms as such were defined in terms of hydrogen generation, unreacted metal, and coolant inventroy, and in terms of the form, sequencing and mode of dispersal through the primary vessel boundary. Fission product release was not to be considered as part of this study. Premature termination of the project, however, led to the dicontinuation of work on an as is basis. Work on the in-core phase from the point of scram to core debris bed formation was largely completed. A preliminary scoping calculation on the debris bed phase had been initiated. This report documents the status of the study at termination.

  10. Latest experiences in inspecting the inside of BWR vessel shields

    Energy Technology Data Exchange (ETDEWEB)

    Alberdi, R.; Gonzalez, E.

    2001-07-01

    In the last few years, the owners of BWR nuclear power plants have been forced to address new fuel shield inspection requirements, TECNATOM has responded to this situation by launching the TEIDE projects, which include development of an inspection machine and the corresponding Non-Destructive Tests to examine the inside of this shield. With these projects, TECNATOM has performed more than 12 fuel shield inspections in different countries. This article describes the experience gained in the last three years. (Author)

  11. PRESSURE-RESISTANT VESSEL

    NARCIS (Netherlands)

    Beukers, A.; De Jong, T.

    1997-01-01

    Abstract of WO 9717570 (A1) The invention is directed to a wheel-shaped pressure-resistant vessel for gaseous, liquid or liquefied material having a substantially rigid shape, said vessel comprising a substantially continuous shell of a fiber-reinforced resin having a central opening, an inner l

  12. Identification and initial assessment of candidate BWR late-phase in-vessel accident management strategies

    Energy Technology Data Exchange (ETDEWEB)

    Hodge, S.A.

    1991-04-15

    Work sponsored by the United States Nuclear Regulatory Commission (USNRC) to identify and perform preliminary assessments of candidate BWR (boiling water reactor) in-vessel accident management strategies was completed at Oak Ridge National Laboratory (ORNL) during fiscal year 1990. Mitigative strategies for containment events have been the subject of a companion study at Brookhaven National Laboratory. The focus of this Oak Ridge effort was the development of new strategies for mitigation of the late phase events, that is, the events that would occur in-vessel after the onset of significant core damage. The work began with an investigation of the current status of BWR in-vessel accident management procedures and proceeded through a preliminary evaluation of several candidate new strategies. The steps leading to the identification of the candidate strategies are described. The four new candidate late-phase (in-vessel) accident mitigation strategies identified by this study and discussed in the report are: (1) keep the reactor vessel depressurized; (2) restore injection in a controlled manner; (3) inject boron if control blade damage has occurred; and (4) containment flooding to maintain core and structural debris in-vessel. Additional assessments of these strategies are proposed.

  13. Pressurized Vessel Slurry Pumping

    Energy Technology Data Exchange (ETDEWEB)

    Pound, C.R.

    2001-09-17

    This report summarizes testing of an alternate ''pressurized vessel slurry pumping'' apparatus. The principle is similar to rural domestic water systems and ''acid eggs'' used in chemical laboratories in that material is extruded by displacement with compressed air.

  14. GOLD PRESSURE VESSEL SEAL

    Science.gov (United States)

    Smith, A.E.

    1963-11-26

    An improved seal between the piston and die member of a piston-cylinder type pressure vessel is presented. A layer of gold, of sufficient thickness to provide an interference fit between the piston and die member, is plated on the contacting surface of at least one of the members. (AEC)

  15. Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code

    Energy Technology Data Exchange (ETDEWEB)

    Thakre, S.; Ma, W. [Royal Institute of Technology, KTH. Div. of Nuclear Power Safety, Stockholm (Sweden)

    2013-08-15

    Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different

  16. 46 CFR 169.249 - Pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Pressure vessels. 169.249 Section 169.249 Shipping COAST... and Certification Inspections § 169.249 Pressure vessels. Pressure vessels must meet the requirements of part 54 of this chapter. The inspection procedures for pressure vessels are contained in...

  17. 46 CFR 182.330 - Pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Pressure vessels. 182.330 Section 182.330 Shipping COAST...) MACHINERY INSTALLATION Auxiliary Machinery § 182.330 Pressure vessels. All unfired pressure vessels must be... unfired pressure vessels must meet the applicable requirements of subchapter F (Marine Engineering)...

  18. LOCA air-injection loads in BWR Mark II pressure suppression containment systems

    Energy Technology Data Exchange (ETDEWEB)

    Kukita, Y.; Shiba, M. (Japan Atomic Energy Research Inst., Tokai, Ibaraki); Namatame, K. (Institute of Nuclear Safety, Tokyo (Japan))

    1984-02-01

    Large-scale blowdown tests were conducted to investigate the thermal-hydrodynamic response of a boiling-water reactor (BWR) Mark II pressure suppression containment system to a postulated loss-of-coolant accident. This paper presents the test results on the early blowdown transients, where air in the drywell is injected into the pressure suppression pool and induces various hydrodynamic loads onto the containment pressure boundary and internal structures. The test data are compared to predictions by analytical models used for the licensing evaluation of the hydrodynamic loads to assess these models.

  19. 46 CFR 119.330 - Pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Pressure vessels. 119.330 Section 119.330 Shipping COAST... Machinery § 119.330 Pressure vessels. All unfired pressure vessels must be installed to the satisfaction of the cognizant OCMI. The design, construction, and original testing of such unfired pressure...

  20. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.P. [Paul Scherrer Institute, PSI, Villigen (Switzerland); Devrient, B.; Roth, A. [Framatome ANP GmbH, Erlangen (Germany); Ehrnsten, U. [VTT Industrial Systems, Espoo (Finland); Ernestova, M.; Zamboch, M. [Nuclear Research Institute, NRI, Rez (Czech Republic); Foehl, J.; Weissenberg, T. [Staatliche Materialpruefungsanstalt, MPA, Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, CIEMAT, Madrid (Spain)

    2004-07-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  1. 46 CFR 197.462 - Pressure vessels and pressure piping.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Pressure vessels and pressure piping. 197.462 Section... Diving Equipment § 197.462 Pressure vessels and pressure piping. (a) The diving supervisor shall ensure that each pressure vessel, including each volume tank, cylinder and PVHO, and each pressure...

  2. LOCA steam condensation loads in BWR Mark II pressure suppression containment system

    Energy Technology Data Exchange (ETDEWEB)

    Kukita, Y.; Namatame, K.; Takeshita, I.; Shiba, M.

    1987-06-01

    Hydrodynamic loads induced in the BWR Mark II pressure suppression containment system during a loss-of-coolant accident (LOCA) were investigated using a large scale test facility. The maximum-bounding loading conditions on the pressure suppression pool-boundary structures were defined by conducting experiments for a wide range of parameters. The maximum-bounding loads occurred when steam, with air concentration less than 1% in weight, was injected at moderate rates (approx. = 30 kg/m/sup 2/.s) into a low-temperature (below 310 K) pool. Such conditions are most likely to be encountered during LOCAs with intermediate break sizes.

  3. Effect of non-heterogeneous wetwell boundaries on pressure suppression system response. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    McCauley, E.W.; Holman, G.S.; Namatame, K.; Kukita, Y.; Shiba, M.

    1980-08-29

    The Full-Scale Mark II CRT (Containment Response Test) Program is in progress at the Tokai-Mura Establishment of the Japan Atomic Energy Research Institute (JAERI). The primary objective of the on-going CRT Program is to provide a data base for evaluation of the pressure suppression pool (wetwell) hydrodynamic loads associated with a postulated loss-of-coolant accident (LOCA) in the BWR Mark II containment system. The test facility is 1/18 of full scale in volume and has a wetwell which is a full-scale geometric replica of one 20/sup 0/-sector of a reference 1100MWe Mark II.

  4. Hydrogen storage in insulated pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Aceves, S.M.; Garcia-Villazana, O. [Lawrence Livermore National Lab., CA (United States)

    1998-08-01

    Insulated pressure vessels are cryogenic-capable pressure vessels that can be fueled with liquid hydrogen (LH{sub 2}) or ambient-temperature compressed hydrogen (CH{sub 2}). Insulated pressure vessels offer the advantages of liquid hydrogen tanks (low weight and volume), with reduced disadvantages (lower energy requirement for hydrogen liquefaction and reduced evaporative losses). This paper shows an evaluation of the applicability of the insulated pressure vessels for light-duty vehicles. The paper shows an evaluation of evaporative losses and insulation requirements and a description of the current analysis and experimental plans for testing insulated pressure vessels. The results show significant advantages to the use of insulated pressure vessels for light-duty vehicles.

  5. Neutron flux calculation and fluence in the encircling of the core and vessel of a reactor BWR; Calculo de flujo neutronico y fluencia en la envolvente del nucleo y la vasija de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez C, E.

    2011-07-01

    One of the main objectives related to the safety of any nuclear power plant, including the nuclear power plant of Laguna Verde is to ensure the structural integrity of reactor pressure vessel. To identify and quantify the damage caused by neutron irradiation in the vessel of any nuclear reactor, it is necessary to know both the neutron flux and the neutron fluence that the vessel has been receiving during its operation lifetime, and that the damage observed by mechanical testing are products of microstructural effects induced by neutron irradiation; therefore, it is important the study and prediction of the neutron flux in order to have a better understanding of the damage that these materials are receiving. The calculation here described uses the DORT code, which solves the neutron transport equation in discrete ordinates in two dimensions (x-y, r-{theta} and r-z), according to a regulatory guide, it should make an approximation of the neutron flux in three dimensions by the so called synthesis method. It is called in that way because it achieves a representation of 3 Dimensional neutron flux combining or summarizing the fluxes calculated by DORT r-{theta}, r-z and r. This work presents the application of synthesis method, according to Regulatory Guide 1190, to determine the 3 Dimensional fluxes in internal BWR reactor using three different spatial meshes. The results of the neutron flux and fluence, using three different meshes in the directions r, {theta} and z were compared with results reported in the literature obtaining a difference not larger than 9.61%, neutron flux reached its maximum, 1.58 E + 12 n/cm{sup 2}s, at a height H 4 (239.07 cm) and angle 32.236{sup o} in the core shroud and 4.00 E + 09 n/cm{sup 2}s at a height H 4 and angle 35.27{sup o} in the inner wall of the reactor vessel, positions that are consistent to within {+-}10% over the ones reported in the literature. (Author)

  6. Kuosheng BWR/6 containment pressure and temperature responses after recirculation line break using GOTHIC code

    Energy Technology Data Exchange (ETDEWEB)

    Lin, A.; Wang, J-R.; Chen, Y-S., E-mail: samuellin1999@iner.gov.tw, E-mail: jrwang@iner.gov.tw, E-mail: yschen@iner.gov.tw [Inst. of Nuclear Energy Research Atomic Energy Council (China); Shih, C., E-mail: ckshih@ess.nthu.edu.tw [National Tsing Hua Univ., Dept. of Engineering and System Science (China)

    2011-07-01

    In this study, we presented the calculated results of the containment P/T (pressure and temperature) response after the recirculation line break (RCLB) accident of a GE-designed twin-unit BWR/6 plant, which can be served as the design basis for the containment system. During the simulation, a power of SPU (stretch power uprate) range was used and a model of the Mark III type containment was built using the GOTHIC (Generation of Thermal-Hydraulic Information for Containments) code. The calculated results, similar to the FSAR (Final Safety Analysis Report) results, indicate the GOTHIC code has the capability to simulate the containment P/T response to the RCLB accident. (author)

  7. Organic fiber/epoxy pressure vessels

    Science.gov (United States)

    Chiao, T. T.; Hamstad, M. A.; Marcon, M. A.

    1974-01-01

    We evaluated the performance of an organic fiber in an epoxy matrix by winding 20-cm diam spherical and cylindrical pressure vessels of various designs. For the spherical vessels, we used soft aluminum liners 0.76 mm thick for a double boss design and 2 mm thick for a single boss design. For the cylindrical vessels, we used both 0.5-mm rubber liners and 0.76-mm soft aluminum liners. Vessels of both types were tested for burst pressure and cyclic fatigue at room temperature and liquid hydrogen temperature. The effects of temperature and vessel shape on the vessel performance factor were negligible. Our vessel fatigue data were marred by premature failure of the liners.

  8. High Toughness Light Weight Pressure Vessel Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Proposed is a pressure vessel with 25% better Fracture Strength over equal strength designed Fiberglass to help reduce 10 to 25% weight for aerospace use. Phase I is...

  9. Liquid Nitrogen Subcooler Pressure Vessel Engineering Note

    Energy Technology Data Exchange (ETDEWEB)

    Rucinski, R.; /Fermilab

    1997-04-24

    The normal operating pressure of this dewar is expected to be less than 15 psig. This vessel is open to atmospheric pressure thru a non-isolatable vent line. The backpressure in the vent line was calculated to be less than 1.5 psig at maximum anticipated flow rates.

  10. Estimate of radiation-induced steel embrittlement in the BWR core shroud and vessel wall from reactor-grade MOX/UOX fuel for the nuclear power plant at Laguna Verde, Veracruz, Mexico

    Science.gov (United States)

    Vickers, Lisa Rene

    The government of Mexico has expressed interest to utilize the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18--30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons. There is concern that a core with a fraction of MOX fuel (i.e., increased 239Pu wt%) would increase the radiation-induced steel embrittlement within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation-induced steel embrittlement within the core shroud and vessel wall is a concern because of the potentially adverse affect to plant and public safety, environment, and operating life of the reactor. This dissertation provides computational results of the neutron fluence, flux, energy spectrum, and radiation damage displacements per atom per second (dpa-s-1) in steel within the core shroud and vessel wall of the Laguna Verde Unit 1 BWR. The results were computed using the nuclear data processing code NJOY99 and the continuous energy Monte Carlo Neutral Particle transport code MCNP4B. The MCNP4B model of the reactor core was for maximum core loading fractions of ⅓ MOX and ⅔ UOX reactor-grade fuel in an equilibrium core. The primary conclusion of this dissertation was that the addition of the maximum fraction of ⅓ MOX fuel to the LV1 BWR core did significantly accelerate the radiation-induced steel embrittlement such that without mitigation of steel embrittlement by periodic thermal annealing or reduction in operating parameters such as, neutron fluence, core temperature and pressure, it posed a potentially adverse affect to the plant and public safety, environment, and operating life of the reactor.

  11. Blood vessels, circulation and blood pressure.

    Science.gov (United States)

    Hendry, Charles; Farley, Alistair; McLafferty, Ella

    This article, which forms part of the life sciences series, describes the vessels of the body's blood and lymphatic circulatory systems. Blood pressure and its regulatory systems are examined. The causes and management of hypertension are also explored. It is important that nurses and other healthcare professionals understand the various mechanisms involved in the regulation of blood pressure to prevent high blood pressure or ameliorate its damaging consequences.

  12. Thermal hydraulics characterization of the core and the reactor vessel type BWR; Caracterizacion termohidraulica del nucleo y de la vasija de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Zapata Y, M.; Lopez H, L.E. [CFE, Carretera Cardel-Nautla Km. 42.5, Municipio Alto Lucero, Veracruz (Mexico)]. e-mail: marxlenin.zapata@cfe.gob.mx

    2008-07-01

    The thermal hydraulics design of a reactor type BWR 5 as the employees in the nuclear power plant of Laguna Verde involves the coupling of at least six control volumes: Pumps jet region, Stratification region, Core region, Vapor dryer region, Humidity separator region and Reactor region. Except by the regions of the core and reactor, these control volumes only are used for design considerations and their importance as operative data source is limited. It is for that is fundamental to complement the thermal hydraulics relations to obtain major data that allow to determine the efficiency of internal components, such as pumps jet, humidity separator and vapor dryer. Like example of the previous thing, calculations are realized on the humidity of the principal vapor during starting, comparing it with the values at the moment incorporated in the data banks of the computers of process of both units. (Author)

  13. 46 CFR 50.30-15 - Class II pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Class II pressure vessels. 50.30-15 Section 50.30-15... Fabrication Inspection § 50.30-15 Class II pressure vessels. (a) Class II pressure vessels shall be subject to... pressure vessels shall be performed during the welding of the longitudinal joint. At this time the...

  14. 46 CFR 61.10-5 - Pressure vessels in service.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Pressure vessels in service. 61.10-5 Section 61.10-5... INSPECTIONS Tests and Inspections of Pressure Vessels § 61.10-5 Pressure vessels in service. (a) Basic requirements. Each pressure vessel must be examined or tested every 5 years. The extent of the test...

  15. 46 CFR 50.30-20 - Class III pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Class III pressure vessels. 50.30-20 Section 50.30-20... Fabrication Inspection § 50.30-20 Class III pressure vessels. (a) Class III pressure vessels shall be subject... specifically exempted by other regulations in this subchapter. (b) For Class III welded pressure vessels,...

  16. Characterizing Acoustic Sources in Pressure Vessels

    Institute of Scientific and Technical Information of China (English)

    李路明; 郑鹏; 刘时风; 施克仁

    2002-01-01

    The "dream" of acoustic emission (AE) testing is to get the acoustic source characteristics from AE signals, especially when evaluating aging pressure vessels. In this paper, the wavelet transform was used to analyze different AE signals from cracks (surface and inner), pencil-lead-breakage and leakage. These acoustic sources were applied on an actual pressure vessel. While the vessel experienced hydraulic pressure, their AE signals were acquired by a digital AE testing system with a wide frequency band transducer and a high speed A/D converter. Then, the digital signals were analyzed using the wavelet transform method. Correlation coefficients of the transformed data show that the different acoustic sources can be easily identified.

  17. Pressure vessel inspections using ultrasonic phased arrays

    Energy Technology Data Exchange (ETDEWEB)

    Moles, M.D.C. [R/D Tech Toronto, Toronto, ON (Canada); Jacques, F.; Dube, N. [R/D Tech Quebec PQ (Canada)

    2003-07-01

    Pressure vessels are used in several industries, including the petrochemical and petroleum industries. Welds in the pressure vessels often produce defects which propagate and fail with time. Many types of pressure vessel weld inspections can now be conducted using automated ultrasonics which offers several advantages over radiography. In terms of phased arrays, custom tailored time-of-flight diffraction (TOFD) and ASME code Section V raster scans, can successfully perform high speed inspections with minimal operator subjectivity. The phased array beams can be steered, scanned, swept and focused electronically. Beam steering can be used to map welds at appropriate angles to optimize probability of defect detection. The main challenge lies with the initial equipment cost, technology awareness, and availability of trained operators. Phased arrays offer great flexibility for different components, defect detection and tailored imaging. 8 refs., 10 figs.

  18. Curved and conformal high-pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Croteau, Paul F.; Kuczek, Andrzej E.; Zhao, Wenping

    2016-10-25

    A high-pressure vessel is provided. The high-pressure vessel may comprise a first chamber defined at least partially by a first wall, and a second chamber defined at least partially by the first wall. The first chamber and the second chamber may form a curved contour of the high-pressure vessel. A modular tank assembly is also provided, and may comprise a first mid tube having a convex geometry. The first mid tube may be defined by a first inner wall, a curved wall extending from the first inner wall, and a second inner wall extending from the curved wall. The first inner wall may be disposed at an angle relative to the second inner wall. The first mid tube may further be defined by a short curved wall opposite the curved wall and extending from the second inner wall to the first inner wall.

  19. Reactor Structural Materials: Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Chaouadi, R

    2000-07-01

    The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported.

  20. Kendall Analysis of Cannon Pressure Vessels

    Science.gov (United States)

    2012-04-11

    To) 4. TITLE AND SUBTITLE New PVD Technologies for New Ordnance Coatings 5a. CONTRACT NUMBER W911NF-11-D-0001 5b. GRANT NUMBER...yield pressure; autofrettage; fatigue life; cannon pressure vessels; residual stress; Bauschinger effect; 16. SECURITY CLASSIFICATION OF: 17...documents, enter the title classification in parentheses. 5a. CONTRACT NUMBER. Enter all contract numbers as they appear in the report, e.g. F33615-86

  1. 46 CFR 115.812 - Pressure vessels and boilers.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Pressure vessels and boilers. 115.812 Section 115.812 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) SMALL PASSENGER VESSELS CARRYING MORE... CERTIFICATION Material Inspections § 115.812 Pressure vessels and boilers. (a) Pressure vessels must be...

  2. (Irradiation embrittlement of reactor pressure vessels)

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, W.R.

    1990-09-24

    The traveler served as a member of the two-man US Nuclear Regulatory Commission sponsored team who visited the Prometey Complex in Leningrad to assess the potential for expanded cooperative research concerning integrity of the primary pressure boundary in commercial light-water reactors. The emphasis was on irradiation embrittlement, structural analysis, and fracture mechanics research for reactor pressure vessels. At the irradiation seminar in Cologne, presentations were made by German, French, Finnish, Russian, and US delegations concerning many aspects of irradiation of pressure vessel steels. The traveler made presentations on mechanisms of irradiation embrittlement and on important aspects of the Heavy-Section Steel Irradiation Program results of irradiated fracture mechanics tests.

  3. Preliminary Assessment of the Possible BWR Core/Vessel Damage States for Fukushima Daiichi Station Blackout Scenarios Using RELAP/SCDAPSIM

    Directory of Open Access Journals (Sweden)

    C. M. Allison

    2012-01-01

    Full Text Available Immediately after the accident at Fukushima Daiichi, Innovative Systems Software and other members of the international SCDAP Development and Training Program started an assessment of the possible core/vessel damage states of the Fukushima Daiichi Units 1–3. The assessment included a brief review of relevant severe accident experiments and a series of detailed calculations using RELAP/SCDAPSIM. The calculations used a detailed RELAP/SCDAPSIM model of the Laguna Verde BWR vessel and related reactor cooling systems. The Laguna Verde models were provided by the Comision Nacional de Seguridad Nuclear y Salvaguardias, the Mexican nuclear regulatory authority. The initial assessment was originally presented to the International Atomic Energy Agency on March 21 to support their emergency response team and later to our Japanese members to support their Fukushima Daiichi specific analysis and model development.

  4. Conformable pressure vessel for high pressure gas storage

    Energy Technology Data Exchange (ETDEWEB)

    Simmons, Kevin L.; Johnson, Kenneth I.; Lavender, Curt A.; Newhouse, Norman L.; Yeggy, Brian C.

    2016-01-12

    A non-cylindrical pressure vessel storage tank is disclosed. The storage tank includes an internal structure. The internal structure is coupled to at least one wall of the storage tank. The internal structure shapes and internally supports the storage tank. The pressure vessel storage tank has a conformability of about 0.8 to about 1.0. The internal structure can be, but is not limited to, a Schwarz-P structure, an egg-crate shaped structure, or carbon fiber ligament structure.

  5. 46 CFR 58.60-3 - Pressure vessel.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Pressure vessel. 58.60-3 Section 58.60-3 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING MAIN AND AUXILIARY MACHINERY AND... Pressure vessel. A pressure vessel that is a component in an industrial system under this subpart must...

  6. Computing the partial volume of pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Wiencke, Bent [Nestle USA, Corporate Engineering, 800 N. Brand Blvd, Glendale, CA 91203 (United States)

    2010-06-15

    The computation of the partial and total volume of pressure vessels with various type of head profiles requires detailed knowledge of the head profile geometry. Depending on the type of head profile the derivation of the equations can become very complex and the calculation process cumbersome. Certain head profiles require numerical methods to obtain the partial volume, which for most application is beyond the scope of practicability. This paper suggests a unique method that simplifies the calculation procedure for the various types of head profiles by using one common set of equations without the need for numerical or complex computation methods. For ease of use, all equations presented in this paper are summarized in a single table format for horizontal and vertical vessels. (author)

  7. Structural integrity of nuclear reactor pressure vessels

    Science.gov (United States)

    Knott, John F.

    2013-09-01

    The paper starts from concerns expressed by Sir Alan Cottrell, in the early 1970s, related to the safety of the pressurized water reactor (PWR) proposed at that time for the next phase of electrical power generation. It proceeds to describe the design and operation of nuclear generation plant and gives details of the manufacture of PWR reactor pressure vessels (RPVs). Attention is paid to stress-relief cracking and under-clad cracking, experienced with early RPVs, explaining the mechanisms for these forms of cracking and the means taken to avoid them. Particular note is made of the contribution of non-destructive inspection to structural integrity. Factors affecting brittle fracture in RPV steels are described: in particular, effects of neutron irradiation. The use of fracture mechanics to assess defect tolerance is explained, together with the failure assessment diagram embodied in the R6 procedure. There is discussion of the Master Curve and how it incorporates effects of irradiation on fracture toughness. Dangers associated with extrapolation of data to low probabilities are illustrated. The treatment of fatigue-crack growth is described, in the context of transients that may be experienced in the operation of PWR plant. Detailed attention is paid to the thermal shock associated with a large loss-of-coolant accident. The final section reviews the arguments advanced to justify 'Incredibility of Failure' and how these are incorporated in assessments of the integrity of existing plant and proposed 'new build' PWR pressure vessels.

  8. Pressure vessel calculations for VVER-440 reactors.

    Science.gov (United States)

    Hordósy, G; Hegyi, Gy; Keresztúri, A; Maráczy, Cs; Temesvári, E; Vértes, P; Zsolnay, E

    2005-01-01

    For the determination of the fast neutron load of the reactor pressure vessel a mixed calculational procedure was developed. The procedure was applied to the Unit II of Paks NPP, Hungary. The neutron source on the outer surfaces of the reactor was determined by a core design code, and the neutron transport calculations outside the core were performed by the Monte Carlo code MCNP. The reaction rate in the activation detectors at surveillance positions and at the cavity were calculated and compared with measurements. In most cases, fairly good agreement was found.

  9. Midland reactor pressure vessel flaw distribution

    Energy Technology Data Exchange (ETDEWEB)

    Foulds, J.R.; Kennedy, E.L. [Failure Analysis Associates, Inc., Menlo Park, CA (United States); Rosinski, S.T. [Sandia National Labs., Albuquerque, NM (United States)

    1993-12-01

    The results of laboratory nondestructive examination (NDE), and destructive cross-sectioning of selected weldment sections of the Midland reactor pressure vessel were analyzed per a previously developed methodology in order to develop a flaw distribution. The flaw distributions developed from the NDE results obtained by two different ultrasonic test (UT) inspections (Electric Power Research Institute NDE Center and Pacific Northwest Laboratories) were not statistically significantly different. However, the distribution developed from the NDE Center`s (destructive) cross-sectioning-based data was found to be significantly different than those obtained through the UT inspections. A fracture mechanics-based comparison of the flaw distributions showed that the cross-sectioning-based data, conservatively interpreted (all defects considered as flaws), gave a significantly lower vessel failure probability when compared with the failure probability values obtained using the UT-based distributions. Given that the cross-sectioning data were reportedly biased toward larger, more significant-appearing (by UT) indications, it is concluded that the nondestructive examinations produced definitively conservative results. In addition to the Midland vessel inspection-related analyses, a set of twenty-seven numerical simulations, designed to provide a preliminary quantitative assessment of the accuracy of the flaw distribution method used here, were conducted. The calculations showed that, in more than half the cases, the analysis produced reasonably accurate predictions.

  10. Digital implementation, simulation and tests in MATLAB of the models of Steam line, the turbines, the pressure regulator of a BWR type nucleo electric power plant; Implementacion digital, simulacion y pruebas en MATLAB de los modelos de la linea de vapor, las turbinas y el regulador de presion de una central Nucleoelectrica tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lopez R, A. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: andyskamx@yahoo.com.mx

    2004-07-01

    In this phase of the project they were carried out exhaustive tests to the models of the steam lines, turbines and pressure regulator of a BWR type nucleo electric central for to verify that their tendencies and behaviors are it more real possible. For it, it was necessary to also analyze the transfer functions of the different components along the steam line until the power generator. Such models define alone the dominant poles of the system, what is not limitation to reproduce a wide range of anticipated transitoriness of a power station operation. In the same manner, it was integrated and proved the integrated model form with the models of feeding water of the SUN-RAH, simulating the nuclear reactor starting from predetermined entrances of the prospective values of the vessel. Also it was coupled with the graphic interface developed with the libraries DirectX implementing a specific monitoring panel for this system. (Author)

  11. Reactor pressure vessel structural integrity research

    Energy Technology Data Exchange (ETDEWEB)

    Pennell, W.E.; Corwin, W.R. [Oak Ridge National Lab., TN (United States)

    1995-04-01

    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 shallows 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{sub 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 a 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.

  12. NDE and Stress Monitoring on Composite Overwrapped Pressure Vessels Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Damage caused by composite overwrapped pressure vessels (COPVs) failure can be catastrophic. Thus, monitoring condition and stress in the composite overwrap,...

  13. Radiation effects on reactor pressure vessel supports

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, R.E. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Technology; Lipinski, R.E. [Idaho National Engineering Lab., Rockville, MD (United States)

    1996-05-01

    The purpose of this report is to present the findings from the work done in accordance with the Task Action Plan developed to resolve the Nuclear Regulatory Commission (NRC) Generic Safety Issue No. 15, (GSI-15). GSI-15 was established to evaluate the potential for low-temperature, low-flux-level neutron irradiation to embrittle reactor pressure vessel (RPV) supports to the point of compromising plant safety. An evaluation of surveillance samples from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) had suggested that some materials used for RPV supports in pressurized-water reactors could exhibit higher than expected embrittlement rates. However, further tests designed to evaluate the applicability of the HFIR data to reactor RPV supports under operating conditions led to the conclusion that RPV supports could be evaluated using traditional method. It was found that the unique HFIR radiation environment allowed the gamma radiation to contribute significantly to the embrittlement. The shielding provided by the thick steel RPV shell ensures that degradation of RPV supports from gamma irradiation is improbable or minimal. The findings reported herein were used, in part, as the basis for technical resolution of the issue.

  14. Simulation of a scenario of total loss of external and internal power (Sbo) for different vent pressures of the containment of a BWR-5; Simulacion de un escenario de perdida total de potencia externa e interna (SBO) para distintas presiones de venteo de la contencion de un reactor BWR-5

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Mugica R, C. A.; Godinez S, V., E-mail: Jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2014-10-15

    The simulation of a Station Black Out (Sbo) was realized with intervention of the vent containment by means of a rigid vent coming from the dry-well and that discharges directly to the atmosphere, with the MELCOR code version 2.1. This scenario was carried out for a BWR-5 and containment type Mark II, with a thermal power of 2317 MWt similar to the reactor of nuclear power plant of Laguna Verde. For this scenario was considered as only available system for coolant injection to the reactor to the Reactor Core Isolation Cooling (Rcic), which remained operating 4 hours with batteries bank. The Security and Relief Valves (SR V) were considered functional (by simplicity) and that they mechanically do not exceed their capacity to liberate pressure due to the performances in their safety way. The operator maneuver to perform the SR V and to de pressurize the vessel until the pressure (13 kg/cm{sup 2}) to operate the low pressure systems was modeled. The results cover approximately 48 hours (172000 seconds), time in which was observed the behavior of the level and pressure in the vessel. Also the scenario evolution was analyzed to different vent pressures of the primary containment (2.0, 3.0, 4.5, 6.0, and 10.0 kg/cm{sup 2}), the temperature profiles of the dry-well, the hydrogen accumulation in the containment, the radio-nuclides liberation through rigid vent to the atmosphere and the inventory of these. In this work an analysis of the pressure behavior in the primary containment is presented, with the purpose of minimizing liberated fission products to the environment. (Author)

  15. 46 CFR 176.812 - Pressure vessels and boilers.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Pressure vessels and boilers. 176.812 Section 176.812... TONS) INSPECTION AND CERTIFICATION Material Inspections § 176.812 Pressure vessels and boilers. (a.... (b) Periodic inspection and testing requirements for boilers are contained in § 61.05 in subchapter...

  16. Fatigue life of organic fiber/epoxy pressure vessels

    Science.gov (United States)

    Hamstad, M. A.; Chiao, T. T.; Patterson, R. G.

    1975-01-01

    The cyclic fatigue life of 10.2-cm-diam cylindrical pressure vessels has been studied. The vessels were made of an organic fiber/epoxy composite. To determine the typical strength distribution of the vessels, 25 of them were internally pressurized until they burst. Twenty-five vessels were then tested under sinusoidal cycling at 1 Hz between 4% and 91% of the mean burst strength. An additional twenty-five vessels were tested between 4% and 91% with a rectangular pressure pulse at 1/3 Hz. A limited number of vessels were tested for stress rupture at the 91% level. Cyclic life was found to depend on time at peak load as well as the number of stress cycles.

  17. Evaluation of insulated pressure vessels for cryogenic hydrogen storage

    Energy Technology Data Exchange (ETDEWEB)

    Aceves, S M; Garcia-Villazana, O; Martinez-Frias, J

    1999-03-01

    This paper presents an analytical and experimental evaluation of the applicability of insulated pressure vessels for hydrogen-fueled light-duty vehicles. Insulated pressure vessels are cryogenic-capable pressure vessels that can be fueled with liquid hydrogen (LH?) or ambient-temperature compressed hydrogen (CH2). Insulated pressure vessels offer the advantages of liquid hydrogen tanks (low weight and volume), with reduced disadvantages (lower energy requirement for hydrogen liquefaction and reduced evaporative losses). The purpose of this work is to verify that commercially available aluminum-lined, fiber- wrapped vessels can be used for cryogenic hydrogen storage. The paper reports on previous and ongoing tests and analyses that have the purpose of improving the system design and assure its safety.

  18. ASME code ductile failure criteria for impulsively loaded pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Nickell, Robert E.; Duffey, T. A. (Thomas A.); Rodriguez, E. A. (Edward A.)

    2003-01-01

    Ductile failure criteria suitable for application to impulsively loaded high pressure vessels that are designed to the rules of the ASME Code Section VI11 Division 3 are described and justified. The criteria are based upon prevention of load instability and the associated global failure mechanisms, and on protection against progressive distortion for multiple-use vessels. The criteria are demonstrated by the design and analysis of vessels that contain high explosive charges.

  19. Quantification of Processing Effects on Filament Wound Pressure Vessels

    Science.gov (United States)

    Aiello, Robert A.; Chamis, Christos C.

    1999-01-01

    A computational simulation procedure is described which is designed specifically for the modeling and analysis of filament wound pressure vessels. Cylindrical vessels with spherical or elliptical end caps can be generated automatically. End caps other than spherical or elliptical may be modeled by varying circular sections along the x-axis according to the C C! end cap shape. The finite element model generated is composed of plate type quadrilateral shell elements on the entire vessel surface. This computational procedure can also be sued to generate grid, connectivity and material cards (bulk data) for component parts of a larger model. These bulk data are assigned to a user designated file for finite element structural/stress analysis of composite pressure vessels. The procedure accommodates filament would pressure vessels of all types of shells-of-revolution. It has provisions to readily evaluate initial stresses due to pretension in the winding filaments and residual stresses due to cure temperature.

  20. Quantification of Processing Effects on Filament Wound Pressure Vessels. Revision

    Science.gov (United States)

    Aiello, Robert A.; Chamis, Christos C.

    2002-01-01

    A computational simulation procedure is described which is designed specifically for the modeling and analysis of filament wound pressure vessels. Cylindrical vessels with spherical or elliptical end caps can be generated automatically. End caps other than spherical or elliptical may be modeled by varying circular sections along the x-axis according to the end cap shape. The finite element model generated is composed of plate type quadrilateral shell elements on the entire vessel surface. This computational procedure can also be used to generate grid, connectivity and material cards (bulk data) for component parts of a larger model. These bulk data are assigned to a user designated file for finite element structural/stress analysis of composite pressure vessels. The procedure accommodates filament wound pressure vessels of all types of shells-of -revolution. It has provisions to readily evaluate initial stresses due to pretension in the winding filaments and residual stresses due to cure temperature.

  1. Influence of residual stresses on failure pressure of cylindrical pressure vessels

    Institute of Scientific and Technical Information of China (English)

    M. Jeyakumar; T. Christopher

    2013-01-01

    The utilization of pressure vessels in aerospace applications is manifold. In this work, finite element analysis (FEA) has been carried out using ANSYS software package with 2D axisym-metric model to access the failure pressure of cylindrical pressure vessel made of ASTM A36 carbon steel having weld-induced residual stresses. To find out the effect of residual stresses on failure pressure, first an elasto-plastic analysis is performed to find out the failure pressure of pressure vessel not having residual stresses. Then a thermo-mechanical finite element analysis is performed to assess the residual stresses developed in the pressure vessel during welding. Finally one more elasto-plastic analysis is performed to assess the effect of residual stresses on failure pressure of the pressure vessel having residual stresses. This analysis indicates reduction in the failure pressure due to unfavorable residual stresses.

  2. Evaluation of pressure vessel fluence computation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C. H.; Whang, I. S.; Kim, T. G.; Lee, H. C. [Seoul National Univ., Seoul (Korea, Republic of); Jang, M. H.; Whang, H. R.; Park, W. S.; An, J. G. [Korea Atomic Energy Research Insitute, Daejeon (Korea, Republic of)

    1994-04-15

    This study was performed as follows: evaluation of neutron fluence calculational methodology through the analysis of benchmark problem, evaluation of calculational results of Yonggwang 3 and 4 reactor vessel fluence, examination of calculational results against the requirements by 10CFR 50.61 and/or standard review plan. The preservation of reactor vessel integrity throughout the reactor lifetime is directly related to the economical and safe operation of nuclear power plants. In this regard, it is very important to accurately predict and assess the neutron fluence which impacts directly upon the reactor vessel integrity. The accurate. prediction and assessment of the reactor vessel fluence require the use of accurate data as well as systematic methodology. However, it is felt that all of these prerequisites are not sufficient at the moment. It is, therefore, recommended to establish a systematic methodology with sufficient nuclear data library for the reliable licensing review of the reactor vessel safety, by performing R and D to resolve the problems presented in this study and by using the results of this study.

  3. Interpretation of Strain Measurements on Nuclear Pressure Vessels

    DEFF Research Database (Denmark)

    Andersen, Svend Ib Smidt; Engbæk, Preben

    1980-01-01

    Selected results from strain measurements on four 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, nozzles, internal vessel structure and flange bolts. The resu......Selected results from strain measurements on four 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, nozzles, 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 a detailed knowledge of the behaviour of the signal...... 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...

  4. Correlation and spectral measurements of fluctuating pressures and velocities in annular turbulent flow. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, R.J.; Jones, B.G.; Roy, R.P.

    1980-02-01

    An experimental study of the fluctuating velocity field, the fluctuating static wall pressure and the in-stream fluctuating static pressure in an annular turbulent air flow system with a radius ratio of 4.314 has been conducted. The study included direct measurements of the mean velocity profile, turbulent velocity field; fluctuating static wall pressure and in-stream fluctuating static pressure from which the statistical values of the turbulent intensity levels, power spectral densities of the turbulent quantities, the cross-correlation between the fluctuating static wall pressure and the fluctuating static pressure in the core region of the flow and the cross-correlation between the fluctuating static wall pressure and the fluctuating velocity field in the core region of the flow were obtained.

  5. Heritability of retinal vessel diameters and blood pressure

    DEFF Research Database (Denmark)

    Taarnhøj, Nina C B B; Larsen, Michael; Sander, Birgit

    2006-01-01

    PURPOSE: To assess the relative influence of genetic and environmental effects on retinal vessel diameters and blood pressure in healthy adults, as well as the possible genetic connection between these two characteristics. METHODS: In 55 monozygotic and 50 dizygotic same-sex healthy twin pairs......%-80%) for CRAE, 83% (95% CI: 73%-89%) for CRVE, and 61% (95% CI: 44%-73%) for mean arterial blood pressure (MABP). Retinal artery diameter decreased with increasing age and increasing arterial blood pressure. Mean vessel diameters in the population were 165.8 +/- 14.9 microm for CRAE, 246.2 +/- 17.7 microm...... and blood glucose, variations in retinal blood vessel diameters and blood pressure were predominantly attributable to genetic effects. A genetic influence may have a role in individual susceptibility to hypertension and other vascular diseases. The results suggest that retinal vessel diameters...

  6. Calculations of plastic collapse load of pressure vessel using FEA

    Institute of Scientific and Technical Information of China (English)

    Peng-fei LIU; Jin-yang ZHENG; Li MA; Cun-jian MIAO; Lin-lin WU

    2008-01-01

    This paper proposes a theoretical method using finite element analysis (FEA) to calculate the plastic collapse loads of pressure vessels under internal pressure, and compares the analytical methods according to three criteria stated in the ASME Boiler Pressure Vessel Code. First, a finite element technique using the arc-length algorithm and the restart analysis is developed to conduct the plastic collapse analysis of vessels, which includes the material and geometry non-linear properties of vessels. Second,as the mechanical properties of vessels are assumed to be elastic-perfectly plastic, the limit load analysis is performed by employing the Newton-Raphson algorithm, while the limit pressure of vessels is obtained by the twice-elastic-slope method and the tangent intersection method respectively to avoid excessive deformation. Finally, the elastic stress analysis under working pressure is conducted and the stress strength of vessels is checked by sorting the stress results. The results are compared with those obtained by experiments and other existing models. This work provides a reference for the selection of the failure criteria and the calculation of the plastic collapse load.

  7. Environmentally-Assisted Cracking of Low-Alloy Reactor Pressure Vessel Steels under Boiling Water Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2002-02-01

    The present report summarizes the experimental work performed by PSI on the environmentally-assisted cracking (EAC) of low-alloy steels (LAS) in the frame of the RIKORR-project during the period from January 2000 to August 2001. Within this project, the EAC crack growth behaviour of different low-alloy reactor pressure vessel (RPV) steels, weld filler and weld heat-affected zone materials is investigated under simulated transient and steady-state BWR/NWC power operation conditions. The EAC crack growth behaviour of different low-alloy RPV steels was characterized by slow rising load (SRL) / low-frequency corrosion fatigue (LFCF) and constant load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 C. These tests revealed the following important interim results: Under low-flow and highly oxidizing (ECP >= 100 mV SHE) conditions, the ASME XI 'wet' reference fatigue crack growth curve could be significantly exceeded by cyclic fatigue loading at low frequencies (<0.001 Hz), at high and low load-ratios R, and by ripple loading near to DKth fatigue thresholds. The BWR VIP 60 SCC disposition lines may be significantly or slightly exceeded (even in steels with a low sulphur content) in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 -250 C) in RPV materials, which show a distinct susceptibility to dynamic strain ageing (DSA). (author)

  8. High pressure gas vessels for neutron scattering experiments

    CERN Document Server

    Done, R; Evans, B E; Bowden, Z A

    2010-01-01

    The combination of high pressure techniques with neutron scattering proves to be a powerful tool for studying the phase transitions and physical properties of solids in terms of inter-atomic distances. In our report we are going to review a high pressure technique based on a gas medium compression. This technique covers the pressure range up to ~0.7GPa (in special cases 1.4GPa) and typically uses compressed helium gas as the pressure medium. We are going to look briefly at scientific areas where high pressure gas vessels are intensively used in neutron scattering experiments. After that we are going to describe the current situation in high pressure gas technology; specifically looking at materials of construction, designs of seals and pressure vessels and the equipment used for generating high pressure gas.

  9. 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.

  10. BWR suppression pool pressures during safety relief valve discharge. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, G.; Gay, R.R.

    1985-04-01

    An analytical understanding of the pool pressures measured during safety relief valve (SRV) discharge in BWRs equipped with x-quenchers has been developed and compared to experimental data. The local conditions inside the SRV discharge lines and inside of the x-quencher were modeled successfully with RELAP5. The measured pressure surges inside the quencher are successfully predicted by the code. In addition, the analytical predictions allow one to associate the peak pressure inside the quencher arm with the onset of air discharge into the suppression pool. A Rayleigh model of bubble dynamics successfully explained both the higher-frequency and the ensuing lower-frequency pressure oscillations that have been measured in suppression pools during SRV discharge tests. The higher-frequency oscillations are characteristic of an air bubble emanating from a single row of quencher holes. The lower-frequency pressure oscillations are characteristic of a large air bubble containing all the air expelled from one side of an x-quencher arm.

  11. Structural analysis of in-pool pressure vessel in CNS

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seok Hoon; Lee, K. H.; Lee, J. H.; Lee, H. Y

    2005-08-15

    The in-pool pressure vessel in the Cold Neutron Source consists of the moderator vessel and the vacuum tube which is inserted in vertical hole located in the reflector region. It is necessary that the moderator vessel is designed to minimize its thickness within the extent to satisfy the stress intensity allowable limits not to shrink the cold neutron flux. The vacuum tube should be designed to endure the high pressure loads by the fracture of the moderator vessel and the overpressure of cover gas. The stress calculation was performed to verify the design of the moderator vessel and vacuum tube. The loads taken into account in this analysis are pressure during normal operation, seismic events and thermal expansion. The detail analyses for the moderator vessel and the vacuum tube will be carried out after deciding the loads through the thermal hydraulic analysis. For the detail analysis, the loads such as failure of the moderator vessel and overpressure in cover gas should be considered by the accident analysis. The calculated stresses satisfied the ASME SC-1 component design and analysis rules. In the buckling analysis, the structural integrity was also verified in the vacuum tube such a long cylinder type.

  12. Heating of reactor pressure vessel bottom head and penetrations in a severe reactor accident; Reaktoripaineastian pohjan ja laepivientien kuumeneminen sydaemen sulamisonnettomuudessa

    Energy Technology Data Exchange (ETDEWEB)

    Ikonen, K. [VTT Energy, Espoo (Finland). Nuclear Energy

    1997-10-01

    The report describes the fundamentals of heat conductivity and convection and numerical methods like finite difference and control volume method for calculation of the thermal history of a reactor pressure vessel bottom head and penetrations. Phase changes from solids to liquids are considered. Time integration is performed by explicit or implicit method. Developed computer codes for thermal conductivity and convection analyses and codes for graphical visualization are described. The codes are applied to two practical cases. They deal with analyses of Swiss CORVIS-experiments and analyses of control rod and instrument penetrations in a BWR bottom head. A model for calculation of effective thermal conductivity of granular corium is developed. The work is also related to EU MVI-project (Core Melt-Pressure Vessel Interactions During a Light Water Reactor Severe Accident), whose coordinator is Prof. B. R. Sehgal at Royal Institute of Technology in Stockholm. (orig.) (11 refs.).

  13. 46 CFR 54.01-17 - Pressure vessel for human occupancy (PVHO).

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Pressure vessel for human occupancy (PVHO). 54.01-17... PRESSURE VESSELS General Requirements § 54.01-17 Pressure vessel for human occupancy (PVHO). Pressure vessels for human occupancy (PVHO's) must meet the requirements of subpart B (Commercial Diving...

  14. 30 CFR 56.13015 - Inspection of compressed-air receivers and other unfired pressure vessels.

    Science.gov (United States)

    2010-07-01

    ... other unfired pressure vessels. 56.13015 Section 56.13015 Mineral Resources MINE SAFETY AND HEALTH... and other unfired pressure vessels. (a) Compressed-air receivers and other unfired pressure vessels... applicable chapters of the National Board Inspection Code, a Manual for Boiler and Pressure Vessel...

  15. Expert system for evaluating the safety of pressure vessels

    Institute of Scientific and Technical Information of China (English)

    Dong Zhibo; Lu Yafeng; Wei Yanhong; Yang Yongfu; Ma Rui; Guo Ping

    2009-01-01

    With more application of welding technology in important structures more attention was paid to the evaluation of the safety of welded structures, the life prediction and decision to repair the welded structures. Based on material fiacture mechanism and Chinese standard of safety evaluations of pressure vessels, an expert system was developed to evaluate the safety of welded pressure vessels. The system can analyze the weld defects in a pressure vessel, convert different kinds of defects into equivalent cracks and obtain their equivalent sizes. Furthermore, the system can calculate the stress and strain in the positions of weld defects and make decision on whether the defects are tolerable or not according to the code. When it is tolerable, the system will calculate the safety margin. The fatigue life can be predicted if the defects undergo fatigue load too. Moreover, data bases are built for storing mechanical properties of material and evaluated results.

  16. Selection of materials for pressure vessels and chemical plants

    Energy Technology Data Exchange (ETDEWEB)

    Huppertz, P.H.; Retter, A. (Linde A.G., Hoellriegelskreuth (Germany, F.R.). Werksgruppe Tieftemperatur und Verfahrenstechnik)

    1980-04-01

    The selection of materials for pressure vessels and chemical plants depends on a number of factors such as operating, operating temperature, operating medium, regulations in force in the country of the plant user concerned and manufacturing possibilities. The essay clearly explains how the above specified factors individually influence the selection of materials. The article also deals with the ranges of application of certain material groups such as unalloyed and low-alloy steels, fine-grained steels, austenitic chromium-nickel steels, unalloyed ferritic chromium steels and other materials. The article closes with remarks on the operational safety of pressure vessels.

  17. Industrial safety of pressure vessels - structural integrity point of view

    Directory of Open Access Journals (Sweden)

    Sedmak Aleksandar

    2016-01-01

    Full Text Available This paper presents different aspects of pressure vessel safety in the scope of industrial safety, focused to the chemical industry. Quality assurance, including application of PED97/23 has been analysed first, followed shortly by the risk assessment and in details by the structural integrity approach, which has been illustrated with three case studies. One important conclusion, following such an approach, is that so-called water proof testing can actually jeopardize integrity of a pressure vessel instead of proving it. [Projekat Ministarstva nauke Republike Srbije, br. TR 174004 i br. TR 33044

  18. SMART composite high pressure vessels with integrated optical fiber sensors

    Science.gov (United States)

    Blazejewski, Wojciech; Czulak, Andrzej; Gasior, Pawel; Kaleta, Jerzy; Mech, Rafal

    2010-04-01

    In this paper application of integrated Optical Fiber Sensors for strain state monitoring of composite high pressure vessels is presented. The composite tanks find broad application in areas such as: automotive industry, aeronautics, rescue services, etc. In automotive application they are mainly used for gaseous fuels storage (like CNG or compressed Hydrogen). In comparison with standard steel vessels, composite ones have many advantages (i.e. high mechanical strength, significant weight reduction, etc). In the present work a novel technique of vessel manufacturing, according to this construction, was applied. It is called braiding technique, and can be used as an alternative to the winding method. During braiding process, between GFRC layers, two types of optical fiber sensors were installed: point sensors in the form of FBGs as well as interferometric sensors with long measuring arms (SOFO®). Integrated optical fiber sensors create the nervous system of the pressure vessel and are used for its structural health monitoring. OFS register deformation areas and detect construction damages in their early stage (ensure a high safety level for users). Applied sensor system also ensured a possibility of strain state monitoring even during the vessel manufacturing process. However the main application of OFS based monitoring system is to detect defects in the composite structure. An idea of such a SMART vessel with integrated sensor system as well as an algorithm of defect detection was presented.

  19. Threaded insert for compact cryogenic-capable pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Loza, Francisco; Ross, Timothy O.; Switzer, Vernon A.; Aceves, Salvador M.; Killingsworth, Nicholas J.; Ledesma-Orozco, Elias

    2015-06-16

    An insert for a cryogenic capable pressure vessel for storage of hydrogen or other cryogenic gases at high pressure. The insert provides the interface between a tank and internal and external components of the tank system. The insert can be used with tanks with any or all combinations of cryogenic, high pressure, and highly diffusive fluids. The insert can be threaded into the neck of a tank with an inner liner. The threads withstand the majority of the stress when the fluid inside the tank that is under pressure.

  20. Optimal design of pressure vessel using an improved genetic algorithm

    Institute of Scientific and Technical Information of China (English)

    Peng-fei LIU; Ping XU; Shu-xin HAN; Jin-yang ZHENG

    2008-01-01

    As the idea of simulated annealing(SA) is introduced into the fitness function,an improved genetic algorithm(GA) is proposed to perform the optimal design of a pressure vessel which aims to attain the minimum weight under burst pressure constraint.The actual burst pressure is calculated using the arc-length and restart analysis in finite element analysis(FEA).A penalty function in the fitness function is proposed to denl with the constrained problem.The erects of the population size and the number of generations in the GA on the weight and burst pressure of the vessel are explored.The optimization results using the proposed GA are also compared with those using the simple GA and the conventional Monte Carlo method.

  1. Reactor pressure vessels as type B transport containment boundaries

    Energy Technology Data Exchange (ETDEWEB)

    Nickell, R.E. [Applied Science and Technology, Inc., Poway, CA (United States); Griesbach, T.J. [ATI Consulting, Danville, CA (United States)

    1998-07-01

    Transportation risk and personnel exposure, as well as the cost of decommissioning nuclear power plants, can all be reduced significantly through the one-time use of the reactor pressure vessel as a containment boundary for shipping the activated internal components from the reactor site to a burial site. In order to help provide the technical basis for this end-use application, the ASME Board on Nuclear Codes and Standards, through its Subcommittee XI, has prepared a draft nuclear code case that contains requirements for any modifications to the vessel, including materials, design, fabrication, and examination. In particular, the requirements for evaluation of potential brittle fracture as the result of potentially low ambient shipping temperatures combined with hypothetical transportation accident loading are addressed. Existing ASME Code Section XI rules for linear elastic fracture mechanics evaluation of irradiated reactor pressure vessels have been adapted and included in the code case. (authors)

  2. USER SPECIFICATIONS FOR PRESSURE VESSELS AND TECHNICAL INTEGRITY

    Directory of Open Access Journals (Sweden)

    K.S. Johnston

    2012-01-01

    Full Text Available

    ENGLISH ABSTRACT: Specifications translated from user requirements are prescribed in an attempt to capture and incorporate best practices with regards to the design, fabrication, testing, and operation of pressure vessels. The question as to whether these requirements affect the technical integrity of pressure vessels is often a subjective matter. This paper examines typical user requirement specifications against technical integrity of pressure vessels.
    The paper draws on a survey of a convenience sample of practising engineers in a diversified petrochemical company. When compared with failures on selected pressure vessels recorded by Phillips and Warwick, the respondent feedback confirms the user specifications that have the highest impact on technical integrity.

    AFRIKAANSE OPSOMMING: Gebruikersbehoeftes word saamgevat in spesifikasies wat lei tot goeie praktyk vir ontwerp, vervaarding, toetsing en bedryf van drukvate. Subjektiwiteit van die gebruikersbehoeftes mag soms die tegniese integriteit van ‘n drukvat beinvloed.
    Die navorsing maak by wyse van monsterneming gebruik van die kennis van ingenieurs wat werk in ‘n gediversifiseerde petrochemiese bedryf. Die terugvoering bevestig dat bogenoemde spesifikasies inderdaad die grootste invloed het op tegniese integriteit.

  3. Status update of the BWR cask simulator

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximum thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations

  4. 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...

  5. 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...

  6. Circular cylinders and pressure vessels stress analysis and design

    CERN Document Server

    Vullo, Vincenzo

    2014-01-01

    This book provides comprehensive coverage of stress and strain analysis of circular cylinders and pressure vessels, one of the classic topics of machine design theory and methodology. Whereas other books offer only a partial treatment of the subject and frequently consider stress analysis solely in the elastic field, Circular Cylinders and Pressure Vessels broadens the design horizons, analyzing theoretically what happens at pressures that stress the material beyond its yield point and at thermal loads that give rise to creep. The consideration of both traditional and advanced topics ensures that the book will be of value for a broad spectrum of readers, including students in postgraduate, and doctoral programs and established researchers and design engineers. The relations provided will serve as a sound basis for the design of products that are safe, technologically sophisticated, and compliant with standards and codes and for the development of innovative applications.

  7. Reliability Considerations for Composite Overwrapped Pressure Vessels on Spacecraft

    Science.gov (United States)

    Murthy, Pappu L. N.; Gyekenyesi, John P.; Grimes-Ledesma, Lorie; Phoenix, S. L.

    2007-01-01

    Composite Overwrapped Pressure Vessels (COPVs) are used to store gases under high pressure onboard spacecraft. These are used for a variety of purposes such as propelling liquid fuel etc, Kevlar, glass, Carbon and other more recent fibers have all been in use to overwrap the vessels. COPVs usually have a thin metallic liner with the primary purpose of containing the gases and prevent any leakage. The liner is overwrapped with filament wound composite such as Kevlar, Carbon or Glass fiber. Although the liner is required to perform in the leak before break mode making the failure a relatively benign mode, the overwrap can fail catastrophically under sustained load due to stress rupture. It is this failure mode that is of major concern as the stored energy of such vessels is often great enough ta cause loss of crew and vehicle. The present paper addresses some of the reliability concerns associated specifically with Kevlar Composite Overwrapped Pressure Vessels. The primary focus of the paper is on how reliability of COPV's are established for the purpose of deciding in general their flight worthiness and continued use. Analytical models based on existing design data will be presented showing how to achieve the required reliability metric to the end of a specific period of performance. Uncertainties in the design parameters and how they affect reliability and confidence intervals will be addressed as well. Some trade studies showing how reliability changes with time during a program period will be presented.

  8. Fabrication of toroidal composite pressure vessels. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Dodge, W.G.; Escalona, A.

    1996-11-24

    A method for fabricating composite pressure vessels having toroidal geometry was evaluated. Eight units were fabricated using fibrous graphite material wrapped over a thin-walled aluminum liner. The material was wrapped using a machine designed for wrapping, the graphite material was impregnated with an epoxy resin that was subsequently thermally cured. The units were fabricated using various winding patterns. They were hydrostatically tested to determine their performance. The method of fabrication was demonstrated. However, the improvement in performance to weight ratio over that obtainable by an all metal vessel probably does not justify the extra cost of fabrication.

  9. Improved Attachment in a Hybrid Inflatable Pressure Vessel

    Science.gov (United States)

    Johnson, Christopher J.; Patterson, Ross; Spexarth, Gary R.

    2010-01-01

    The vessel is a hybrid that comprises an inflatable shell attached to a rigid structure. The inflatable shell is, itself, a hybrid that comprises (1) a pressure bladder restrained against expansion by (2) a restraint layer that comprises a web of straps made from high-strength polymeric fabrics. The present improvements are intended to overcome deficiencies in those aspects of the original design that pertain to attachment of the inflatable shell to the rigid structure. In a typical intended application, such attachment(s) would be made at one or more window or hatch frames to incorporate the windows or hatches as integral parts of the overall vessel.

  10. Simplified system for the pressure control of a Nucleo electric central of the BWR type; Sistema simplificado para el control de presion de una central Nucleoelectrica del tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez J, J. [FI-UNAM, DEPFI Campus Morelos, Jiutepec, Morelos (Mexico)

    2003-07-01

    One of the main preoccupations of the electric power generator stations is the appropriate operation of the same ones. The operators must be qualified to respond in an adequate way and to be able to take to these power stations to an optimal, sure and stable operation condition under any circumstance. The Laboratory of Analysis in Nuclear Reactors Engineering (LAIRN) of the Engineering Faculty of UNAM (Fl) in collaboration with the International Atomic Energy Agency (IAEA), it develops an interactive classroom simulator in which simulations of the phenomena which take place in a nuclear power station are executed. The classroom simulator bases its operation on specialized nuclear codes feeding interactive graphic unfolding with those that it is possible to make a monitoring, supervision and control of the behavior of the power station under any operation regime, either in normal operation, transitory events or postulated accident sequence. The development of this classroom simulator includes a modular and re configurable structure. Due to it is indispensable to count with a higher inter activity with the system it is included the simulation of the control system of the plant and inside the same, one of those more important it is the reactor pressure control system. The present work describes the conceptual design and the used methodology for the development and implementation in the simulator of a simplified model of the pressure control system for a BWR generic central. The reach of the development will allow to accomplish the necessary tests to demonstrate that this has an adequate performance according to the carried out simplifications. (Author)

  11. Estimation of ex-vessel steam explosion pressure loads

    Energy Technology Data Exchange (ETDEWEB)

    Leskovar, Matjaz [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)], E-mail: matjaz.leskovar@ijs.si; Ursic, Mitja [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

    2009-11-15

    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 this article, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is being developed for the simulation of fuel-coolant interactions. A parametric study was performed varying the location of the melt release (central, right and left 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 establish the influence of the varied parameters on the fuel-coolant interaction behaviour, to determine the most challenging cases and to estimate the expected pressure loadings on the cavity walls. For the most explosive central, right side and left side melt pour scenarios a detailed analysis of the explosion simulation results was performed. The study shows that for some ex-vessel steam explosion scenarios higher pressure loads are predicted than obtained in the OECD programme SERENA phase 1.

  12. Multivariable analysis of a failure event of pressure regulator in a BWR; Analisis multivariable de un evento de falla del regulador de presion en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Castillo D, R.; Ortiz V, J. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Calleros M, G. [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Carretera Cardel-Nautla, Km. 43.5, Veracruz (Mexico)], e-mail: rogelio.castillo@inin.gob.mx

    2009-10-15

    The boiling water reactors can experiment three types of instabilities: one caused by the controllers failure of plant, another renowned instability by reactivity and the last knew as thermal hydraulics instability. An event of pressure regulator failure of electro-hydraulic control of Unit 1 of nuclear power plant of Laguna Verde was analyzed, which caused power oscillations that were increasing their magnitude in the time course. The event has been analyzed using the Fourier transformation in short time for time-frequency analysis and for the frequency domain be employment the power spectral density. Both techniques reported a resonance to oscillation frequency of 0.055 Hz in the power spectrum, this frequency is of observed order of magnitude when fail the reactor control systems. However, these analysis did not allow to study the interrelation of event signals. Of the previous studies, were obtained power spectral densities containing picks and valleys related with the dynamic behaviour of reactor, which includes the control systems performance. For a pick or present valley to a specific frequency in the power spectrum for one of previous variables, can determine the influence of other variables on the pick or valley by relative contribution of power. This method was established in a developed program of name Noise, which uses a multivariable autoregressive model to obtain the autoregressive coefficients, and starting from them the relative contribution of power is determined. Basically two important results were obtained, the first is related with the influence of feed water flow on the other variables to the frequency of 0.055 Hz, the second is related with the instability by reactivity and confirms that this way was not excited during the event. (Author)

  13. Neutron activation analysis and activity in the vessel steel of a BWR reactor for their study without radiological risks in microscopy and spectrometry; Analisis de activacion neutronica y actividad en el acero de la vasija de un reactor nuclear tipo BWR para su estudio sin riesgos radiologicos en microscopia y espectrometria

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel, M.; Garcia B, A. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Fisica, Unidad Profesional Adolfo Lopez Mateos, Zacatenco, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: mmoranchel@ipn.mx [IAEA, Department of Technical Cooperation, Division for Latin America, Room B1109 Wagramerstrasse 5, PO Box 100, A-1400, Vienna (Austria)

    2012-07-01

    The vessel material of nuclear reactors is subject to irradiation damage induced by the bombardment of neutrons coming from the reactor core. Neutrons are classified as fast and thermal, which produce different effects. Fast neutrons cause damage to the material by dislocation or displacement of atoms in the crystal structure, while the effect of thermal neutrons is a nuclear transmutation that can significantly change the properties of the material. The type and intensity of damage is based on the characteristics of the material, the flow of neutrons and the modes of neutrons interaction with the atomic structures of the material, among others. This work, alluding to nuclear transmutation, makes an analysis of neutron activation of all isotopes in a steel boiling water nuclear reactor (BWR) vessel. An analytical expression is obtained in order to model activity of steel, on the basis of the weight percentage of its atomic components. Its activity is theoretically estimated in a witness sample of the same material as that of the vessel, placed within the nuclear reactor since the beginning of its commercial operation in April 1995, up to August 2010. It was theoretically determined that the witness sample, with a 0.56 g mass (1 x 1 x 0.07 cm{sup 3} dimensions or equivalent) does not present a radiological risks during the stage of preparation, observation and analysis of it in electron microscopy and X-ray diffraction equipment s. The theoretical results were checked experimentally by measuring the activity of the sample by means of gamma spectrometry, measurement of the exposure levels around the sample, as well as the induced level to whole body and limbs, using thermo-luminescent dosimetry (TLD). As a result of the theoretical analysis, new chemical elements are predicted, as a result of the activation phenomena and radioactive decay, whose presence can be a fundamental factor of change in the properties of the vessel. This work is a preamble to the

  14. ESTIMATE OF BURSTING PRESSURE OF MILD STEEL PRESSURE VESSEL AND PRESENTATION OF BURSTING FORMULA

    Institute of Scientific and Technical Information of China (English)

    ZHENG Chuanxiang

    2006-01-01

    In order to get more precise bursting pressure formula of mild steel, hundreds of bursting experiments of mild steel pressure vessels such as Q235(Gr.D) and 20R(1020) are done. Based on statistical data of bursting pressure and modification of Faupel formula, a more precise modified formula is given out according to the experimental data. It is proved to be more accurate after examining other bursting pressure value presented in many references. This bursting formula is very accurate in these experiments using pressure vessels with different diameter and shell thickness.Obviously, this modified bursting formula can be used in mild steel pressure vessels with different diameter and thickness of shell.

  15. 46 CFR 109.421 - Report of repairs to boilers and pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Report of repairs to boilers and pressure vessels. 109... Report of repairs to boilers and pressure vessels. Before making repairs, except normal repairs and maintenance such as replacement of valves or pressure seals, to boilers or unfired pressure vessels...

  16. 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.

  17. 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.

  18. Quality Testing of Gaseous Helium Pressure Vessels by Acoustic Emission

    CERN Document Server

    Barranco-Luque, M; Hervé, C; Margaroli, C; Sergo, V

    1998-01-01

    The resistance of pressure equipment is currently tested, before commissioning or at periodic maintenance, by means of normal pressure tests. Defects occurring inside materials during the execution of these tests or not seen by usual non-destructive techniques can remain as undetected potential sources of failure . The acoustic emission (AE) technique can detect and monitor the evolution of such failures. Industrial-size helium cryogenic systems employ cryogens often stored in gaseous form under pressure at ambient temperature. Standard initial and periodic pressure testing imposes operational constraints which other complementary testing methods, such as AE, could significantly alleviate. Recent reception testing of 250 m3 GHe storage vessels with a design pressure of 2.2 MPa for the LEP and LHC cryogenic systems has implemented AE with the above-mentioned aims.

  19. Design Considerations For Blast Loads In Pressure Vessels.

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, E. A. (Edward A.); Nickell, Robert E.; Pepin, J. E. (Jason E.)

    2007-01-01

    Los Alamos National Laboratory (LANL), under the auspices of the U.S. Department of Energy (DOE) and the National Nuclear Security Administration (NNSA), conducts confined detonation experiments utilizing large, spherical, steel pressure vessels to contain the reaction products and hazardous materials from high-explosive (HE) events. Structural design and analysis considerations include: (a) Blast loading phase (i.e., impulsive loading); (b) Dynamic structural response; (c) Fragment (i.e., shrapnel) generation and penetration; (d) Ductile and non-ductile fracture; and (e) Design Criteria to ASME Code Sec. VIII, Div. 3, Impulsively Loaded Vessels. These vessels are designed for one-time-use only, efficiently utilizing the significant plastic energy absorption capability of ductile vessel materials. Alternatively, vessels may be designed for multiple-detonation events, in which case the material response is restricted to elastic or near-elastic range. Code of Federal Regulations, Title 10 Part 50 provides requirements for commercial nuclear reactor licensing; specifically dealing with accidental combustible gases in containment structures that might cause extreme loadings. The design philosophy contained herein may be applied to extreme loading events postulated to occur in nuclear reactor and non-nuclear systems or containments.

  20. Stress Rupture Life Reliability Measures for Composite Overwrapped Pressure Vessels

    Science.gov (United States)

    Murthy, Pappu L. N.; Thesken, John C.; Phoenix, S. Leigh; Grimes-Ledesma, Lorie

    2007-01-01

    Composite Overwrapped Pressure Vessels (COPVs) are often used for storing pressurant gases onboard spacecraft. Kevlar (DuPont), glass, carbon and other more recent fibers have all been used as overwraps. Due to the fact that overwraps are subjected to sustained loads for an extended period during a mission, stress rupture failure is a major concern. It is therefore important to ascertain the reliability of these vessels by analysis, since the testing of each flight design cannot be completed on a practical time scale. The present paper examines specifically a Weibull statistics based stress rupture model and considers the various uncertainties associated with the model parameters. The paper also examines several reliability estimate measures that would be of use for the purpose of recertification and for qualifying flight worthiness of these vessels. Specifically, deterministic values for a point estimate, mean estimate and 90/95 percent confidence estimates of the reliability are all examined for a typical flight quality vessel under constant stress. The mean and the 90/95 percent confidence estimates are computed using Monte-Carlo simulation techniques by assuming distribution statistics of model parameters based also on simulation and on the available data, especially the sample sizes represented in the data. The data for the stress rupture model are obtained from the Lawrence Livermore National Laboratories (LLNL) stress rupture testing program, carried out for the past 35 years. Deterministic as well as probabilistic sensitivities are examined.

  1. The coolability limits of a reactor pressure vessel lower head

    Energy Technology Data Exchange (ETDEWEB)

    Theofanous, T.G.; Syri, S. [Univ. of California, Santa Barbara, CA (United States)

    1995-09-01

    Configuration II of the ULPU experimental facility is described, and from a comprehensive set of experiments are provided. The facility affords full-scale simulations of the boiling crisis phenomenon on the hemispherical lower head of a reactor pressure vessel submerged in water, and heated internally. Whereas Configuration I experiments (published previously) established the lower limits of coolability under low submergence, pool-boiling conditions, with Configuration II we investigate coolability under conditions more appropriate to practical interest in severe accident management; that is, heat flux shapes (as functions of angular position) representative of a core melt contained by the lower head, full submergence of the reactor pressure vessel, and natural circulation. Critical heat fluxes as a function of the angular position on the lower head are reported and related the observed two-phase flow regimes.

  2. Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL

    2010-08-01

    Materials behaviors caused by neutron irradiation under fission and/or fusion environments can be little understood without practical examination. Easily accessible material information system with large material database using effective computers is necessary for design of nuclear materials and analyses or simulations of the phenomena. The developed Embrittlement Data Base (EDB) at ORNL is this comprehensive collection of data. EDB database contains power reactor pressure vessel surveillance data, the material test reactor data, foreign reactor data (through bilateral agreements authorized by NRC), and the fracture toughness data. The lessons learned from building EDB program and the associated database management activity regarding Material Database Design Methodology, Architecture and the Embedded QA Protocol are described in this report. The development of IAEA International Database on Reactor Pressure Vessel Materials (IDRPVM) and the comparison of EDB database and IAEA IDRPVM database are provided in the report. The recommended database QA protocol and database infrastructure are also stated in the report.

  3. Development of PIE techniques for irradiated LWR pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Nishi, Masahiro; Kizaki, Minoru; Sukegawa, Tomohide [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-09-01

    For the evaluation of safety and integrity of light water reactors (LWRs), various post irradiation examinations (PIEs) of reactor pressure vessel (RPV) steels and fuel claddings have been carried out in the Research Hot Laboratory (RHL). In recent years, the instrumented Charpy impact testing machine was remodeled aiming at the improvement of accuracy and reliability. By this remodeling, absorbed energy and other useful information on impact properties can be delivered from the force-displacement curve for the evaluation of neutron irradiation embrittlement behavior of LWR-RPV steels at one-time striking. In addition, two advanced PIE technologies are now under development. One is the remote machining of mechanical test pieces from actual irradiated pressure vessel steels. The other is development of low-cycle and high-cycle fatigue test technology in order to clarify the post-irradiation fatigue characteristics of structural and fuel cladding materials. (author)

  4. Pressure distension in leg vessels as influenced by prolonged bed rest and a pressure habituation regimen.

    Science.gov (United States)

    Eiken, Ola; Mekjavic, Igor B; Kounalakis, Stylianos N; Kölegård, Roger

    2016-06-15

    Bed rest increases pressure distension in arteries, arterioles, and veins of the leg. We hypothesized that bed-rest-induced deconditioning of leg vessels is governed by the removal of the local increments in transmural pressure induced by assuming erect posture and, therefore, can be counteracted by intermittently increasing local transmural pressure during the bed rest. Ten men underwent 5 wk of horizontal bed rest. A subatmospheric pressure (-90 mmHg) was intermittently applied to one lower leg [pressure habituation (PH) leg]. Vascular pressure distension was investigated before and after the bed rest, both in the PH and control (CN) leg by increasing local distending pressure, stepwise up to +200 mmHg. Vessel diameter and blood flow were measured in the posterior tibial artery and vessel diameter in the posterior tibial vein. In the CN leg, bed rest led to 5-fold and 2.7-fold increments (P pressure-distension and flow responses, respectively, and to a 2-fold increase in tibial vein pressure distension. In the PH leg, arterial pressure-distension and flow responses were unaffected by bed rest, whereas bed rest led to a 1.5-fold increase in venous pressure distension. It thus appears that bed-rest-induced deconditioning of leg arteries, arterioles, and veins is caused by removal of gravity-dependent local pressure loads and may be abolished or alleviated by a local pressure-habituation regimen.

  5. Selection of material for building pressure vessels and chemical plants

    Energy Technology Data Exchange (ETDEWEB)

    Huppertz, P.H.; Retter, A.

    1979-06-01

    The authors give on extensive survey on the materials used in building pressure vessels and chemical plants for a temperature region of -200 to +1000/sup 0/C. The effect of various influences on the material behaviour is critically examined on the existing control plant, where the differences to foreign control are indicated. NE metals also come into consideration apart from steels, especially with low-temperature application.

  6. Fatigue Test of Domestic Manufactured Reactor Pressure Vessel Steel

    Institute of Scientific and Technical Information of China (English)

    ZHONG; Wei-hua; TONG; Zhen-feng; NING; Guang-sheng; YU; Bin-tao

    2013-01-01

    The CAP1400 will be built by our country,after the self-dependent innovation work on the imported technology of AP1000,which is a 3rd generation NPP.Now,the design of CAP1400 key equipment is ongoing,and the fatigue design of the domestic manufactured key equipment,such as reactor pressure vessel(RPV),is found to be a main problem in the design work,as the fatigue data is lacked.Thus the

  7. Thermally activated deformation of irradiated reactor pressure vessel steel

    Science.gov (United States)

    Böhmert, J.; Müller, G.

    2002-03-01

    Temperature and strain rate change tensile tests were performed on two VVER 1000-type reactor pressure vessel welds with different contents of nickel in unirradiated and irradiated conditions in order to determine the activation parameters of the contribution of the thermally activated deformation. There are no differences of the activation parameters in the unirradiated and the irradiated conditions as well as for the two different materials. This shows that irradiation hardening preferentially results from a friction hardening mechanism by long-range obstacles.

  8. Online Monitoring of Composite Overwrapped Pressure Vessels (COPV)

    DEFF Research Database (Denmark)

    Pereira, Gilmar Ferreira; Figueiredo, Joana; Faria, Hugo

    2015-01-01

    Composite overwrapped pressure vessels (COPV) have been increasingly pointed to as the most effective solution for high pressure storage of liquid and gaseous fluids. Reasonably high stiffness-to-weight ratios make them suitable for both static and mobile applications. However, higher operating...... (FEA) was made using the ABAQUS® platform. In this numerical analysis, accurate and realistic simulation of the different materials, geometry and loading conditions was approached. Particularly, the anisotropic nature of the wound laminate and the varying orientation of the fibers were attained...

  9. Neutron flux reduction programs for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, C.S. [Korea Atomic Energy Research Inst. KAERI, 150 Deogjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, B.C. [Korea Reactor Integrity Surveillance Technology KRIST, 150 Deogjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2011-07-01

    The objective of this work is to implement various fast neutron flux reduction programs on the belt-line region of the reactor pressure vessel to reduce the increasing rate of reference temperature for pressurized thermal shock (RT PTS) for Korea Nuclear Unit 1. A pressurized thermal shock (PTS) event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts in the belt-line region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Generally, the RT PTS value is continuously increasing according to the fast neutron irradiation during the reactor operation, and it can reach the screening criterion prior to the expiration of the operating license. To reduce the increasing rate of RT PTS, various neutron flux reduction programs can be implemented, which are focused on license renewal. In this paper, neutron flux reduction programs, such as low leakage loading pattern strategy, loading of neutron absorber rods, and dummy fuel assembly loading are considered for Korea Nuclear Unit 1, of which the RT PTS value of the leading material (circumferential weld) is going to reach the screening criterion in the near future. To evaluate the effects of the neutron flux reduction programs, plant and cycle specific forward neutron transport calculations for the various neutron flux reduction programs were carried out. For the analysis, all transport calculations were carried out by using the DORT 3.1 discrete ordinate code and BUGLE-96 cross-section library. (authors)

  10. STRESS ANALYSIS AND BURST PRESSURE DETERMINATION OF TWO LAYER COMPOUND PRESSURE VESSEL

    Directory of Open Access Journals (Sweden)

    HARERAM LOHAR

    2013-02-01

    Full Text Available Multilayer pressure vessel is designed to work under high-pressure condition. This paper introduces the stress analysis and the burst pressure calculation of a two-layer shrink fitted pressure vessel. In the shrink-fitting problems, considering long hollow cylinders, the plane strain hypothesis can be regarded as more natural. Generally hoops stress distribution is non-linear and sharply reduced toward the outer surface. By shrink fitting concentric shells towards the inner shells are placed in residual compression so that the initial compressive hoop stress must be relieved by internal pressure before hoop tensile stress are developed. Therefore the maximum hoop stress will be reduced, resulting more burst pressure. The analytical results of stress distribution and burst pressure is calculated and validated by ANSYS Workbench results.

  11. Prevention of organic iodide formation in BWR`s

    Energy Technology Data Exchange (ETDEWEB)

    Karjunen, T. [Finnish Centre for Radiation and Nuclear Safety, Helsinki (Finland); Laitinen, T.; Piippo, J.; Sirkiae, P. [VTT Manufacturing Technology (Finland)

    1996-12-01

    During an accident, many different forms of iodine may emerge. Organic iodides, such as methyl iodide and ethyl iodide, are relatively volatile, and thus their appearance leads to increased concentration of gaseous iodine. Since organic iodides are also relatively immune to most accident mitigation measures, such as sprays and filters, they can affect the accident source term significantly even when only a small portion of iodine is in organic form. Formation of organic iodides may not be limited by the amount of organic substances available. Excessive amounts of methane can be produced, for example, during oxidation of boron carbide, which is used in BWR`s as a neutron absorber material. Another important source is cable insulation. In a BWR, a large quantity of cables is placed below the pressure vessel. Thus a large quantity of pyrolyse gases will be produced, should the vessel fail. Organic iodides can be formed as a result of many different reactions, but at least in certain conditions the main reaction takes place between an organic radical produced by radiolysis and elemental iodine. A necessary requirement for prevention of organic iodide production is therefore that the pH in the containment water pools is kept high enough to eliminate formation of elemental iodine. In a typical BWR the suppression pool water is usually unbuffered. As a result, the pH may be dominated by chemicals introduced during an accident. If no system for adding basic chemicals is operable, the main factor affecting pool water pH may be hydrochloric acid released during cable degradation. Should this occur, the conditions could be very favorable for production of elemental iodine and, consequently, formation of organic iodides. Although high pH is necessary for iodine retention, it could have also adverse effects. High pH may, for example, accelerate corrosion of containment materials and alter the characteristics of the solid corrosion products. (author) 6 figs., 1 tab., 13 refs.

  12. Delivery of cold hydrogen in glass fiber composite pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Weisberg, Andrew H.; Aceves, Salvador M.; Espinosa-Loza, Francisco; Ledesma-Orozco, Elias; Myers, Blake [Lawrence Livermore National Laboratory, Engineering, 7000 East Avenue L-792, Livermore, CA 94551 (United States)

    2009-12-15

    We are proposing to minimize hydrogen delivery cost through utilization of glass fiber tube trailers at 200 K and 70 MPa to produce a synergistic combination of container characteristics with properties of hydrogen gas: (1) hydrogen cooled to 200 K is {proportional_to}35% more compact for a small increase in theoretical storage energy (exergy); and (2) these cold temperatures (200 K) strengthen glass fibers by as much as 50%, expanding trailer capacity without the use of much more costly carbon fiber composite vessels. Analyses based on US Department of Energy H2A cost and efficiency parameters and economic methodology indicate the potential for hydrogen delivery costs below $1/kg H{sub 2}. Dispensing cold hydrogen may also allow rapid refueling without overtemperatures and overpressures which are typically as high as 25%, simplifying automotive vessel design and improving safety while potentially reducing vessel weight and cost. Based on these results, we suggest hydrogen delivery by truck with trailers carrying hydrogen gas at pressures as high as 70 MPa, cooled to approximately 200 K in glass fiber vessels. (author)

  13. 46 CFR 167.25-1 - Boilers, pressure vessels, piping and appurtenances.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Boilers, pressure vessels, piping and appurtenances. 167... SCHOOLS PUBLIC NAUTICAL SCHOOL SHIPS Marine Engineering § 167.25-1 Boilers, pressure vessels, piping and... the following standards for boilers, pressure vessels, piping and appurtenances: (1)...

  14. 46 CFR 167.25-5 - Inspection of boilers, pressure vessels, piping and appurtenances.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Inspection of boilers, pressure vessels, piping and...) NAUTICAL SCHOOLS PUBLIC NAUTICAL SCHOOL SHIPS Marine Engineering § 167.25-5 Inspection of boilers, pressure vessels, piping and appurtenances. The inspection of boilers, pressure vessels, piping and...

  15. 46 CFR 50.05-5 - Existing boilers, pressure vessels or piping systems.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Existing boilers, pressure vessels or piping systems. 50... ENGINEERING GENERAL PROVISIONS Application § 50.05-5 Existing boilers, pressure vessels or piping systems. (a) Whenever doubt exists as to the safety of an existing boiler, pressure vessel, or piping system, the...

  16. 46 CFR 78.33-1 - Repairs of boiler and pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 3 2010-10-01 2010-10-01 false Repairs of boiler and pressure vessels. 78.33-1 Section... OPERATIONS Reports of Accidents, Repairs, and Unsafe Equipment § 78.33-1 Repairs of boiler and pressure vessels. (a) Before making any repairs to boilers or unfired pressure vessels, the chief engineer...

  17. 46 CFR 154.650 - Cargo tank and process pressure vessel welding.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Cargo tank and process pressure vessel welding. 154.650... Equipment Construction § 154.650 Cargo tank and process pressure vessel welding. (a) Cargo tank and process pressure vessel welding must meet Subpart 54.05 and Part 57 of this chapter. (b) Welding consumables...

  18. Jam proof closure assembly for lidded pressure vessels

    Science.gov (United States)

    Cioletti, Olisse C.

    1992-01-01

    An expendable closure assembly is provided for use (in multiple units) with a lockable pressure vessel cover along its rim, such as of an autoclave. This assembly is suited to variable compressive contact and locking with the vessel lid sealing gasket. The closure assembly consists of a thick walled sleeve insert for retention in the under bores fabricated in the cover periphery and the sleeve is provided with internal threading only. A snap serves as a retainer on the underside of the sleeve, locking it into an under bore retention channel. Finally, a standard elongate externally threaded bolt is sized for mating cooperation with the so positioned sleeve, whereby the location of the bolt shaft in the cover bore hole determines its compressive contact on the underlying gasket.

  19. Treating asphericity in fuel particle pressure vessel modeling

    Science.gov (United States)

    Miller, Gregory K.; Wadsworth, Derek C.

    1994-07-01

    The prototypical nuclear fuel of the New Production Modular High Temperature Gas-Cooled Reactor (NP-MHTGR) consists of spherical TRISO-coated particles suspended in graphite cylinders. The coating layers surrounding the fuel kernels in these particles consist of pyrolytic carbon layers and a silicon carbide layer. These coating layers act as a pressure vessel which retains fission product gases. In the operating conditions of the NP-MHTGR, a small percentage of these particles (pressure vessels) are expected to fail due to the pressure loading. The fuel particles of the NP-MHTGR deviate to some degree from a true spherical shape, which may have some effect on the failure percentages. A method is presented that treats the asphericity of the particles in predicting failure probabilities for particle samples. It utilizes a combination of finite element analysis and Monte Carlo sampling and is based on the Weibull statistical theory. The method is used here to assess the effects of asphericity in particles of two common geometric shapes, i.e. faceted particles and ellipsoidal particles. The method presented could be used to treat particle anomalies other than asphericity.

  20. Evaluation of Data-Logging Transducer to Passively Collect Pressure Vessel p/T History

    Science.gov (United States)

    Wnuk, Stephen P.; Le, Son; Loew, Raymond A.

    2013-01-01

    Pressure vessels owned and operated by NASA are required to be regularly certified per agency policy. Certification requires an assessment of damage mechanisms and an estimation of vessel remaining life. Since detail service histories are not typically available for most pressure vessels, a conservative estimate of vessel pressure/temperature excursions is typically used in assessing fatigue life. This paper details trial use of a data-logging transducer to passively obtain actual pressure and temperature service histories of pressure vessels. The approach was found to have some potential for cost savings and other benefits in certain cases.

  1. Structural Features and In-service Inspection of the LTNHR-200 Pressure Vessel

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    The pressure vessel of 200 MW low temperature nuclear heating reactor (LTNHR-200) is the main part of primary pressure boundary and its reasonable and reliable structural design is the key point to assure the safe operation of LTNHR-200. The double-shell pressure vessels were designed. LTNHR-200 pressure vessel meets the condition of Leak Before Break and has a relatively low failure probability. Metal containment (outer pressure vessel) has the similar features to LTNHR-200 pressure vessel. There exists no LOCA and core melting with the double vessel. The in-service inspection of the pressure vessel can be simplified greatly because of the safety and structural features of the reactor.

  2. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Science.gov (United States)

    Thiollay, Nicolas; Di Salvo, Jacques; Sandrin, Charlotte; Soldevila, Michel; Bourganel, Stéphane; Fausser, Clément; Destouches, Christophe; Blaise, Patrick; Domergue, Christophe; Philibert, Hervé; Bonora, Jonathan; Gruel, Adrien; Geslot, Benoit; Lamirand, Vincent; Pepino, Alexandra; Roche, Alain; Méplan, Olivier; Ramdhane, Mourad

    2016-02-01

    FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10-2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006-2007 in a geometry representative of 1300 MWe PWR.

  3. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Directory of Open Access Journals (Sweden)

    Thiollay Nicolas

    2016-01-01

    Full Text Available FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10−2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006–2007 in a geometry representative of 1300 MWe PWR.

  4. Retinal vessel diameter changes induced by transient high perfusion pressure

    Institute of Scientific and Technical Information of China (English)

    Yin-Ying; Zhao; Ping-Jun; Chang; Fang; Yu; Yun-E; Zhao

    2014-01-01

    ·AIM: To investigate the effects of transient high perfusion pressure on the retinal vessel diameter and retinal ganglion cells.·METHODS: The animals were divided into four groups according to different infusion pressure and infusion time(60 mm Hg-3min, 60 mm Hg-5min, 100 mm Hg-3min, 100 mm Hg-5min). Each group consisted of six rabbits. The left eye was used as the experimental eye and the right as a control. Retinal vascular diameters were evaluated before, during infusion, immediately after infusion, 5min, 10 min and 30 min after infusion based on the fundus photographs. Blood pressure was monitored during infusion. The eyes were removed after 24 h.Damage to retinal ganglion cell(RGC) was analyzed by histology.·RESULTS: Retina became whiten and papilla optic was pale during perfusion. Measurements showed significant decrease in retinal artery and vein diameter during perfusion in all of the four groups at the proximal of the edge of the optic disc. The changes were significant in the 100 mm Hg-3min group and 100 mm Hg-5min group compared with 60 mm Hg-3min group(P 1=0.025, P 2=0.000).The diameters in all the groups recovered completely after 30 min of reperfusion. The number of RGC)showed no significant changes at the IOP in 100 mm Hg with5 min compared with contralateral untreated eye(P >0.05).·CONCLUSION: Transient fluctuations during infusion lead to temporal changes of retinal vessels, which could affect the retinal blood circulation. The RGCs were not affected by this transient fluctuation. Further studies are necessary to evaluate the effect of pressure during realtime phacoemusification on retinal blood circulation.

  5. 30 CFR 57.13015 - Inspection of compressed-air receivers and other unfired pressure vessels.

    Science.gov (United States)

    2010-07-01

    ... other unfired pressure vessels. 57.13015 Section 57.13015 Mineral Resources MINE SAFETY AND HEALTH... receivers and other unfired pressure vessels. (a) Compressed-air receivers and other unfired pressure... the applicable chapters of the National Board Inspection Code, a Manual for Boiler and Pressure...

  6. Plastic Limit Load Analysis of Cylindrical Pressure Vessels with Different Nozzle Inclination

    Science.gov (United States)

    Prakash, Anupam; Raval, Harit Kishorchandra; Gandhi, Anish; Pawar, Dipak Bapu

    2016-04-01

    Sudden change in geometry of pressure vessel due to nozzle cutout, leads to local stress concentration and deformation, decreasing its strength. Elastic plastic analysis of cylindrical pressure vessels with different inclination angles of nozzle is important to estimate plastic limit load. In the present study, cylindrical pressure vessels with combined inclination of nozzles (i.e. in longitudinal and radial plane) are considered for elastic plastic limit load analysis. Three dimensional static nonlinear finite element analyses of cylindrical pressure vessels with nozzle are performed for incremental pressure loading. The von Mises stress distribution on pressure vessel shows higher stress zones at shell-nozzle junction. Approximate plastic limit load is obtained by twice elastic slope method. Variation in limit pressure with different combined inclination angle of nozzle is analyzed and found to be distinct in nature. Reported results can be helpful in optimizing pressure vessel design.

  7. Research on reasonable winding angle of ribbons of Flat Steel Ribbon Wound Pressure Vessel

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    Flat Steel Ribbon Wound Pressure Vessels (FSRWPVs) are used in many important industry areas. There is no such kind of pressure vessel exploding on operation for its reasonable structure design. Many explosion experiments on Flat Steel Ribbon Wound Pressure Vessel showed that their limited load pressure is related to the winding angle of the steel ribbons.FSRWPVs with reasonable winding angle have better security and lower cost. Reasonable angels given at the end of this paper facilitate engineering design.

  8. Certification Testing and Demonstration of Insulated Pressure Vessels for Vehicular Hydrogen and Natural Gas Storage

    Energy Technology Data Exchange (ETDEWEB)

    Aceves, S M; Martinez-Frias, J; Espinosa-Loza, F; Schaffer, R; Clapper, W

    2002-05-22

    We are working on developing an alternative technology for storage of hydrogen or natural gas on light-duty vehicles. This technology has been titled insulated pressure vessels. Insulated pressure vessels are cryogenic-capable pressure vessels that can accept either liquid fuel or ambient-temperature compressed fuel. Insulated pressure vessels offer the advantages of cryogenic liquid fuel tanks (low weight and volume), with reduced disadvantages (fuel flexibility, lower energy requirement for fuel liquefaction and reduced evaporative losses). The work described in this paper is directed at verifying that commercially available pressure vessels can be safely used to store liquid hydrogen or LNG. The use of commercially available pressure vessels significantly reduces the cost and complexity of the insulated pressure vessel development effort. This paper describes a series of tests that have been done with aluminum-lined, fiber-wrapped vessels to evaluate the damage caused by low temperature operation. All analysis and experiments to date indicate that no significant damage has resulted. Future activities include a demonstration project in which the insulated pressure vessels will be installed and tested on two vehicles. A draft standard will also be generated for obtaining insulated pressure vessel certification.

  9. PRESSURE AND PRESSURE GRADIENT IN AN AXISYMMETRIC RIGID VESSEL WITH STENOSIS

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    Based on an improvement of the Karman-Pohlhausen's method, using nonlinear polynomial fitting and numerical integral, the axial distributions of pressure and its gradient in an axisymmetric rigid vessel with stenosis were obtained, and the distributions related to Reynolds number and the geometry of stenotic vessel were discussed. It shows that with the increasing of stenotic degree or Reynolds number, the fluctuation of pressure and its gradient in stenotic area is intense rapidly, and negative pressure occurs subsequently in the diverging part of stenotic area. Especially when the axial range of stenosis extends, the flow of blood in the diverging part will be more obviously changed.In higher Reynolds number or heavy stenosis, theoretical calculation is mainly in accordance with past experiments.

  10. Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

    Energy Technology Data Exchange (ETDEWEB)

    Chakraborty, Pritam [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Biner, Suleyman Bulent [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Spencer, Benjamin Whiting [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-07-01

    The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

  11. Multipurpose Pressure Vessel Scanner and Photon Doppler Velocimetry

    Science.gov (United States)

    Ellis, Tayera

    2015-01-01

    Critical flight hardware typically undergoes a series of nondestructive evaluation methods to screen for defects before it is integrated into the flight system. Conventionally, pressure vessels have been inspected for flaws using a technique known as fluorescent dye penetrant, which is biased to inspector interpretation. An alternate method known as eddy current is automated and can detect small cracks better than dye penetrant. A new multipurpose pressure vessel scanner has been developed to perform internal and external eddy current scanning, laser profilometry, and thickness mapping on pressure vessels. Before this system can be implemented throughout industry, a probability of detection (POD) study needs to be performed to validate the system’s eddy current crack/flaw capabilities. The POD sample set will consist of 6 flight-like metal pressure vessel liners with defects of known size. Preparation for the POD includes sample set fabrication, system operation, procedure development, and eddy current settings optimization. For this, collaborating with subject matter experts was required. This technical paper details the preparation activities leading up to the POD study currently scheduled for winter 2015/2016. Once validated, this system will be a proven innovation for increasing the safety and reliability of necessary flight hardware.Additionally, testing of frangible joint requires Photon Doppler Velocimetry (PDV) and Digital Image Correlation instrumentation. There is often noise associated with PDV data, which necessitates a frequency modulation (FM) signal-to-noise pre-test. Generally, FM radio works by varying the carrier frequency and mixing it with a fixed frequency source, creating a beat frequency which is represented by audio frequency that can be heard between about 20 to 20,000 Hz. Similarly, PDV reflects a shifted frequency (a phenomenon known as the Doppler Effect) from a moving source and mixes it with a fixed source frequency, which results in

  12. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-01

    The embrittlement trend curve development project for HFIR reactor pressure vessel (RPV) steels was carried out with three major tasks. Which are (1) data collection to match that used in HFIR steel embrittlement trend published in 1994 Journal Nuclear Material by Remec et. al, (2) new embrittlement data of A212B steel that are not included in earlier HFIR RPV trend curve, and (3) the adjustment of nil-ductility-transition temperature (NDTT) shift data with the consideration of the irradiation temperature effect. An updated HFIR RPV steel embrittlement trend curve was developed, as described below. NDTT( C) = 23.85 log(x) + 203.3 log (x) + 434.7, with 2- uncertainty of 34.6 C, where parameter x is referred to total dpa. The developed update HFIR RPV embrittlement trend curve has higher embrittlement rate compared to that of the trend curve developed in 1994.

  13. ACS Algorithm in Discrete Ordinates for Pressure Vessel Dosimetry

    Directory of Open Access Journals (Sweden)

    Walters William

    2016-01-01

    Full Text Available The Adaptive Collision Source (ACS method can solve the Linear Boltzmann Equation (LBE more efficiently by adaptation of the angular quadrature order. This is similar to, and essentially an extension of, the first collision source method. Previously, the ACS methodology has been implemented into the TITAN discrete ordinates code, and has shown speedups of 2–4 on a simple test problem, with very little loss of accuracy (within a provided adaptive tolerance. This work examines the use of the ACS method for a more realistic problem: pressure vessel dosimetry with the VENUS-2 MOX-fuelled reactor dosimetry benchmark. The ACS method proved to be able to obtain accurate results while being approximately twice as efficient as using a constant quadrature in a standard source iteration scheme.

  14. D-Zero Central Calorimeter Pressure Vessel and Vacuum Vessel Safety Notes

    Energy Technology Data Exchange (ETDEWEB)

    Rucinski, R.; Luther, R.; /Fermilab

    1990-10-25

    The relief valve and relief piping capacity was calculated to be 908 sefm air. This exceeds all relieving conditions. The vessel also has a rupture disc with a 2640 scfm air stamped capacity. In order to significantly decrease the amount of time required to fill the cryostats, it is desired to raise the setpoint of the 'operating' relief valve on the argon storage dewar to 20 psig from its existing 16 psig setting. This additional pressure increases the flow to the cryostats and will overwhelm the relief capacity if the temperature of the modules within these vessels is warm enough. Using some conservative assumptions and simple calculations within this note, the maximum average temperature that the modules within each cryostat can be at prior to filling from the storage dewar with liquid argon is at least 290 K. The average temperature of the module mass for any of the three cryostats can be as high as 290 K prior to filling that particular cryostat. This should not be confused with the average temperature of a single type or location which is useful in protecting the modules-not necessarily the vessel itself. A few modules of each type and at different elevations should be used in an average which would account for the different weights of each module. Note that at 290 K, the actual flow of argon through the relief valve and the rupture disk was under the maximum theoretical flows for each relief device. This means that the bulk temperature could actually have been raised to flow argon through the reliefs at their maximum capacity. Therefore, the temperature of 290 K is a conservative value for the calculated flow rate of 12.3 gpm. Safeguards in addition to and used in conjunction with operating procedures shall be implemented in such a way so that the above temperature limitation is not exceeded and such that it is exclusive of the programmable logic controller (PLC). One suggestion is using a toggle switch for each cryostat mounted in the PLC I/O box

  15. Performance Evaluation Tests of Insulated Pressure Vessels for Vehicular Hydrogen Storage

    Energy Technology Data Exchange (ETDEWEB)

    Aceves, S M; Martinez-Frias, J; Espinoza-Loza, F

    2002-03-01

    Insulated pressure vessels are cryogenic-capable pressure vessels that can be fueled with liquid hydrogen or ambient-temperature compressed hydrogen. This flexibility results in multiple advantages with respect to compressed hydrogen tanks or low-pressure liquid hydrogen tanks. Our work is directed at verifying that commercially available aluminum-lined, fiber-wrapped pressure vessels can be safely used to store liquid hydrogen. A series of tests have been conducted, and the results indicate that no significant vessel damage has resulted from cryogenic operation. Future activities include a demonstration project in which the insulated pressure vessels will be installed and tested on two vehicles. A draft standard will also be generated for certification of insulated pressure vessels.

  16. Certification Testing and Demonstration of Insulated Pressure Vessels for Vehicular Hydrogen Storage

    Energy Technology Data Exchange (ETDEWEB)

    Aceves, S M; Martinez-Frias, J; Espinosa-Loza, F

    2002-05-22

    Insulated pressure vessels are cryogenic-capable pressure vessels that can be fueled with liquid hydrogen or ambient-temperature compressed hydrogen. This flexibility results in multiple advantages with respect to compressed hydrogen tanks or low-pressure liquid hydrogen tanks. Our work is directed at verifying that commercially available aluminum-lined, fiber-wrapped pressure vessels can be safely used to store liquid hydrogen. A series of tests have been conducted, and the results indicate that no significant vessel damage has resulted from cryogenic operation. Future activities include a demonstration project in which the insulated pressure vessels will be installed and tested on two vehicles. A draft standard will also be generated for certification of insulated pressure vessels.

  17. Performance and Certification Testing of Insulated Pressure Vessels for Vehicular Hydrogen Storage

    Energy Technology Data Exchange (ETDEWEB)

    Aceves, S M; Martinez-Frias, J; Garcia-Villazana, O; Espinosa-Loza, F

    2001-06-03

    Insulated pressure vessels are cryogenic-capable pressure vessels that can be fueled with liquid hydrogen (LH2) or ambient-temperature compressed hydrogen (CH2). Insulated pressure vessels offer the advantages of liquid hydrogen tanks (low weight and volume), with reduced disadvantages (fuel flexibility, lower energy requirement for hydrogen liquefaction and reduced evaporative losses). The work described here is directed at verifying that commercially available pressure vessels can be safely used to store liquid hydrogen. The use of commercially available pressure vessels significantly reduces the cost and complexity of the insulated pressure vessel development effort. This paper describes a series of tests that have been done with aluminum-lined, fiber-wrapped vessels to evaluate the damage caused by low temperature operation. All analysis and experiments to date indicate that no significant damage has resulted. Required future tests are described that will prove that no technical barriers exist to the safe use of aluminum-fiber vessels at cryogenic temperatures. Future activities also include a demonstration project in which the insulated pressure vessels will be installed and tested on two vehicles. A draft standard will also be generated for obtaining certification for insulated pressure vessels.

  18. Insulated Pressure Vessels for Vehicular Hydrogen Storage: Analysis and Performance Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Aceves, S M; Martinez-Frias, J; Garcia-Villazana, O; Espinosa-Loza, F

    2001-06-26

    Insulated pressure vessels are cryogenic-capable pressure vessels that can be fueled with liquid hydrogen (LH{sub 2}) or ambient-temperature compressed hydrogen (CH{sub 2}). Insulated pressure vessels offer the advantages of liquid hydrogen tanks (low weight and volume), with reduced disadvantages (fuel flexibility, lower energy requirement for hydrogen liquefaction and reduced evaporative losses). The work described here is directed at verifying that commercially available pressure vessels can be safely used to store liquid hydrogen. The use of commercially available pressure vessels significantly reduces the cost and complexity of the insulated pressure vessel development effort. This paper describes a series of tests that have been done with aluminum-lined, fiber-wrapped vessels to evaluate the damage caused by low temperature operation. All analysis and experiments to date indicate that no significant damage has resulted. Required future tests are described that will prove that no technical barriers exist to the safe use of aluminum-fiber vessels at cryogenic temperatures. Future activities also include a demonstration project in which the insulated pressure vessels will be installed and tested on two vehicles. A draft standard will also be generated for obtaining certification for insulated pressure vessels.

  19. SAFT inspections for developing empirical database of fabrication flaws in nuclear reactor pressure vessels

    Science.gov (United States)

    Doctor, Steven R.; Schuster, George J.; Pardini, Allan F.

    1998-03-01

    The Pacific Northwest National Laboratory (PNNL) is developing a methodology for estimating the size and density distribution of fabrication flaws in U.S. nuclear reactor pressure vessels. This involves the nondestructive evaluation (NDE) of reactor pressure vessel materials and the destructive validation of the flaws found. NDE has been performed on reactor pressure vessel material made by Babcock & Wilcox and Combustion Engineering. A metallographic analysis is being performed to validate the flaw density and size distributions estimated from the 2500 indications of fabrication flaws that were detected and characterized in the very sensitive SAFT-UT (synthetic aperture focusing technique for ultrasonic testing) inspection data from the Pressure Vessel Research User Facility (PVRUF) vessel at Oak Ridge National Laboratory. Research plans are also described for expanding the work to include other reactor pressure vessel materials.

  20. Creep of A508/533 Pressure Vessel Steel

    Energy Technology Data Exchange (ETDEWEB)

    Richard Wright

    2014-08-01

    ABSTRACT Evaluation of potential Reactor Pressure Vessel (RPV) steels has been carried out as part of the pre-conceptual Very High Temperature Reactor (VHTR) design studies. These design studies have generally focused on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Initially, three candidate materials were identified by this process: conventional light water reactor (LWR) RPV steels A508 and A533, 2¼Cr-1Mo in the annealed condition, and Grade 91 steel. The low strength of 2¼Cr-1Mo at elevated temperature has eliminated this steel from serious consideration as the VHTR RPV candidate material. Discussions with the very few vendors that can potentially produce large forgings for nuclear pressure vessels indicate a strong preference for conventional LWR steels. This preference is based in part on extensive experience with forging these steels for nuclear components. It is also based on the inability to cast large ingots of the Grade 91 steel due to segregation during ingot solidification, thus restricting the possible mass of forging components and increasing the amount of welding required for completion of the RPV. Grade 91 steel is also prone to weld cracking and must be post-weld heat treated to ensure adequate high-temperature strength. There are also questions about the ability to produce, and very importantly, verify the through thickness properties of thick sections of Grade 91 material. The availability of large components, ease of fabrication, and nuclear service experience with the A508 and A533 steels strongly favor their use in the RPV for the VHTR. Lowering the gas outlet temperature for the VHTR to 750°C from 950 to 1000°C, proposed in early concept studies, further strengthens the justification for this material selection. This steel is allowed in the ASME Boiler and Pressure Vessel Code for nuclear service up to 371°C (700°F); certain excursions above that temperature are

  1. A methodology for the control of the residual lifetimes of carbon fibre reinforced composite pressure vessels

    OpenAIRE

    Bunsell, Anthony R.; Blassiau, Sébastien; Thionnet, Alain

    2005-01-01

    International audience; Pressure vessels must be periodically proof tested. Traditional techniques for metal vessels are inapplicable for composite vessels as the latter do not break by crack propagation so that the reasoning behind the traditional testing procedures is not appropriate. Damage accumulation leading to the degradation of a composite vessel is by fibre failure. Fibres show a wide distribution in strengths and loading a composite inevitably breaks some. The method which has been ...

  2. Modelling of the dynamics of the vessel and circuits of recirculation of a BWR type nucleo electric as part of the SUN-RAH university simulator; Modelado de la dinamica de la vasija y circuitos de recirculacion de una nucleoelectrica tipo BWR como parte del simulador universitario SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez S, R.A. [DEPFI, Campus Morelos, en IMTA, Jiutepec, Morelos (Mexico)]. e-mail: rsanchez_15@yahoo.com.mx

    2003-07-01

    In the present project, the development of a model for the dynamics of the process of energy transport generated in the nuclear fuel until the main steam lines of a nucleo electric central with BWR type nuclear reactor, using mathematical models of reduced order is presented. These models present the main characteristics of the reactor vessel and of the recirculation system, defined by the main phenomena that intervene in those physical processes. Likewise, the objective of the general project of the one University student nucleo electric simulator with Boiling Water Reactor (SUN-RAH) for later on to establish the modeling equations for each part of the nuclear reactor as well as of the load pursuit system. Also, its were described the graphic interfaces implemented in an three layers architecture in which the different measuring variables are presented in the monitor. It fits signalize that the advantage presented by the University student nucleo electric simulator is the possibility to carry out changes in the magnitudes of those different variables that intervene in the physical processes made in the one reactor and in the recirculation system in execution time of the same one. Of same way, the creation of a graphic intuitive interface, friendly, and designed with the same technology with the one that the video games are programmed in the present time. Besides all the above mentioned, the pending goals inside of the project are exposed, as well as the developments in construction process or conceptualized to be included in future versions of the simulator. Finally its are thinking about possible scenarios of applications of SUN-RAH, as well as their reaches. (Author)

  3. A simplified approach for assessing the leak-before-break for the flawed pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Kannan, P. [Ramagundam Super Thermal Power Station, NTPC Ltd, Jyothinagar 505215 (India); Amirthagadeswaran, K.S. [Faculty of Mechanical Engineering, Government College of Technology, Coimbatore 641013 (India); Christopher, T. [Faculty of Mechanical Engineering, Government College of Engineering, Tirunelveli 627007 (India); Nageswara Rao, B., E-mail: bnrao52@rediffmail.com [Faculty of Mechanical Engineering, School of Mechanical and Civil Sciences, K L University, Green Fields, Vaddeswaram, Guntur 522502 (India)

    2016-06-15

    Surface cracks or embedded cracks in pressure vessels under service may grow and form stable through-thickness cracks causing leak prior to failure. If this leak-before-break phenomenon takes place, then there is a possibility of preventing the vessel failure. This paper presents a simplified approach for assessing the leak-before-break or failure of the flawed pressure vessels. This approach is validated through comparison of existing test data.

  4. Full system decontamination experience in BWR

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, N.; Sugai, K.; Katayouse, N.; Fujimori, A.; Iida, K.; Hayashi, K. [Tokyo Electric Power Company, Tokyo (Japan); Kanasaki, T.; Inami, I. [Toshiba Corporation, Yokohama (Japan); Strohmer, F. [Framatome ANP Gmbh, Eelangen (Germany)

    2002-07-01

    At the Fukushima Daiichi Nuclear Power Station unit 3, unit 2, unit 5 and unit 1 of Tokyo Electric Power Company (TEPCO), the replacement of the core shroud and internals has been conducted since 1997 in this order. The welded core internals in operating BWR plants were replaced to improve stress corrosion cracking (SCC) resistance. At present these units are operating smoothly. The developed technology concept is to restore those internals in open air inside the reactor pressure vessel (RPV). To reduce the radiation dose rate inside the RPV, not only a shielding method was applied to cut the radiation from the irradiated structures but also a chemical decontamination method was applied to dissolve the radioactive crud deposited on the surface by using chemical agents. The calculated decontamination factor (DF) at the RPV bottom reached 35-117. As result, the dose rate decreased to approximately 0.1 mSv/h under water. Before and after the installation of the in-vessel shielding, a mechanical cleaning was extensively applied inside the RPV to remove the residual crud as well as the swarf, chips from cutting. As a result, the dose rate at the RPV bottom decreased to ranging from 0.2 to 0.4 mSv/h in air. A working environment for human access, which was better than expected, was established inside the RPV, resulting in 70, 140, 50 and 70 man-Sv (estimated) saving respectively at unit 3 (1F-3), unit 2(1F-2), unit 5(1F-5) and unit 1(1F-1). All four full system decontamination (FSDs) contributed to the successful realization of the core shroud replacement project under the dry condition in RPV.

  5. Coordinated sensing and autonomous repair of pressure vessels and structures

    Science.gov (United States)

    Huston, Dryver R.; Hurley, David A.; Gollins, Kenneth; Gervais, Anthony

    2010-04-01

    Self-repairing structural systems can potentially improve performance ranges and lifetimes compared to those of conventional systems without self-healing capability. Self-healing materials have been used in automotive and aeronautical applications for over a century. The bulk of these systems operate by using the damage to directly initiate the repair response without any supervisory coordination. Integrating sensing and supervisory control technologies with self-healing may improve the safety and reliability of critical components and structures. This project used laboratory scale test beds to illustrate the benefit of an integrated sensing, control and self-healing system. A thermal healing polymer embedded with resistive heating wires acted as the sensing-healing material. Sensing duties were performed using an impedance, capacitance, and resistance testing device and a PC acted as the controller. As damage occurs to the polymer it is detected, located, and characterized. Based on the sensor signal, a decision is made as to whether to execute a repair and then to subsequently monitor the repair process to ensure completeness. The second demonstration was a self-sealing pressure vessel with integrated sensing and healing capability. These proof-of-concept prototypes can likely be expanded and improved with alternative sensor options, sensing-healing materials, and system architecture.

  6. Development of automatic Ultrasonic testing equipment for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kor R.; Kim, Jae H.; Lee, Jae C.

    1996-06-01

    The selected weld areas of a reactor pressure vessel and adjacent piping are examined by the remote mechanized ultrasonic testing (MUT) equipment. Since the MUT equipment was purchased from southwest Research Institute (SwRI) in April 1985, 15 inservice inspections and 5 preservice inspections are performed with this MUT equipment. However due to the old age of the equipment and frequent movements to plant sites, the reliability of examination was recently decreased rapidly and it is very difficult to keep spare parts. In order to resolve these problems and to meet the strong request from plant sites, we intend to develop a new 3-axis control system including hardware and software. With this control system, we expect more efficient and reliable examination of the nozzle to shell weld areas, which is specified in ASME Code Section XI. The new 3-axis control system hardware and software were designed and development of our own control system, the advanced technologies of computer control mechanism were established and examination reliability of the nozzle to shell weld area was improved. With the development of our 3-axis control system for PaR ISI-2 computer control system, the reliability of nozzle to shell weld area examination has been improved. The established technologies from the development and detailed analysis of existing control system, are expected to be applied to the similar control systems in nuclear power plants. (author). 12 refs., 4 tabs., 33 figs.

  7. Miniaturized Charpy test for reactor pressure vessel embrittlement characterization

    Energy Technology Data Exchange (ETDEWEB)

    Manahan, M.P. Sr. [MPM Research and Consulting, Lemont, PA (United States)

    1999-10-01

    Modifications were made to a conventional Charpy machine to accommodate the miniaturized Charpy V-Notch (MCVN) specimens which were fabricated from an archived reactor pressure vessel (RPV) steel. Over 100 dynamic MCVN tests were performed and compared to the results from conventional Charpy V-Notch (CVN) tests to demonstrate the efficacy of the miniature specimen test. The optimized sidegrooved MCVN specimens exhibit transitional fracture behavior over essentially the same temperature range as the CVN specimens which indicates that the stress fields in the MCVN specimens reasonably simulate those of the CVN specimens and this fact has been observed in finite element calculations. This result demonstrates a significant breakthrough since it is now possible to measure the ductile-brittle transition temperature (DBTT) using miniature specimens with only small correction factors, and for some materials as in the present study, without the need for any correction factor at all. This development simplifies data interpretation and will facilitate future regulatory acceptance. The non-sidegrooved specimens yield energy-temperature data which is significantly shifted downward in temperature (non-conservative) as a result of the loss of constraint which accompanies size reduction.

  8. 77 FR 59408 - Finding of Equivalence; Alternate Pressure Relief Valve Settings on Certain Vessels Carrying...

    Science.gov (United States)

    2012-09-27

    ... SECURITY Coast Guard Finding of Equivalence; Alternate Pressure Relief Valve Settings on Certain Vessels... Coast Guard announces the availability of CG-ENG Policy Letter 04-12, ``Alternative Pressure Relief Valve Settings on Vessels Carrying Liquefied Gases in Bulk in Independent Type B and Type C...

  9. OPTIMUM AUTOFRETTAGE PRESSURE AND SHRINK-FIT COMBINATION FOR MINIMUM STRESS IN MULTILAYER PRESSURE VESSEL

    Directory of Open Access Journals (Sweden)

    S.K. Acharyaa

    2011-05-01

    Full Text Available In present work, more effective ways of decreasing the net working stress in multilayer vessel is brought into focus. Analysis of combined effect of autofrettage and shrink-fit in multi-layeredvessel is carried out. Possible sequences of assembly of autofrettage and shrink-fit in multilayered vessel have been discussed and effective sequence has been sorted out. With the increase in number of layers, sequential order of assembly increases. Optimization of thickness of each vessel for 3-layered vessel, percentage of autofrettage and amount of radial shrink-fit is carried out for all the possible sequences with the help of Genetic Algorithm. While performing optimization, thickness of each layers, autofrettage percentage and radial interference for shrink-fit is considered as design variables, whereas hoop stress throughout the thickness isobjective function. Apart from this, calculation of fatigue life for each case is studied. It is observed that all the possibilities of assembly gives approximately same behaviour under same working pressure with some exceptions.

  10. Vulnerability analysis of a pressurized aluminum composite vessel against hypervelocity impacts

    Science.gov (United States)

    Hereil, Pierre-Louis; Plassard, Fabien; Mespoulet, Jérôme

    2015-09-01

    Vulnerability of high pressure vessels subjected to high velocity impact of space debris is analyzed with the response of pressurized vessels to hypervelocity impact of aluminum sphere. Investigated tanks are CFRP (carbon fiber reinforced plastics) overwrapped Al vessels. Explored internal pressure of nitrogen ranges from 1 bar to 300 bar and impact velocity are around 4400 m/s. Data obtained from Xrays radiographies and particle velocity measurements show the evolution of debris cloud and shock wave propagation in pressurized nitrogen. Observation of recovered vessels leads to the damage pattern and to its evolution as a function of the internal pressure. It is shown that the rupture mode is not a bursting mode but rather a catastrophic damage of the external carbon composite part of the vessel.

  11. Vulnerability analysis of a pressurized aluminum composite vessel against hypervelocity impacts

    Directory of Open Access Journals (Sweden)

    Hereil Pierre-Louis

    2015-01-01

    Full Text Available Vulnerability of high pressure vessels subjected to high velocity impact of space debris is analyzed with the response of pressurized vessels to hypervelocity impact of aluminum sphere. Investigated tanks are CFRP (carbon fiber reinforced plastics overwrapped Al vessels. Explored internal pressure of nitrogen ranges from 1 bar to 300 bar and impact velocity are around 4400 m/s. Data obtained from Xrays radiographies and particle velocity measurements show the evolution of debris cloud and shock wave propagation in pressurized nitrogen. Observation of recovered vessels leads to the damage pattern and to its evolution as a function of the internal pressure. It is shown that the rupture mode is not a bursting mode but rather a catastrophic damage of the external carbon composite part of the vessel.

  12. Determination of the critical buckling pressure of blood vessels using the energy approach.

    Science.gov (United States)

    Han, Hai-Chao

    2011-03-01

    The stability of blood vessels under lumen blood pressure is essential to the maintenance of normal vascular function. Differential buckling equations have been established recently for linear and nonlinear elastic artery models. However, the strain energy in bent buckling and the corresponding energy method have not been investigated for blood vessels under lumen pressure. The purpose of this study was to establish the energy equation for blood vessel buckling under internal pressure. A buckling equation was established to determine the critical pressure based on the potential energy. The critical pressures of blood vessels with small tapering along their axis were estimated using the energy approach. It was demonstrated that the energy approach yields both the same differential equation and critical pressure for cylindrical blood vessel buckling as obtained previously using the adjacent equilibrium approach. Tapering reduced the critical pressure of blood vessels compared to the cylindrical ones. This energy approach provides a useful tool for studying blood vessel buckling and will be useful in dealing with various imperfections of the vessel wall.

  13. Optimized clearing work concept for the BWR containment; Optimiertes Raeumungskonzept fuer SWR-Sicherheitsbehaelter

    Energy Technology Data Exchange (ETDEWEB)

    Kraps, Uwe [AREVA NP GmbH (Germany)

    2012-11-01

    Based on the experiences of reactor dismantling in the NPPs Wuergasse, Obrigheim and Stade an optimized clearing work concept for the BWR containment including the reactor pressure vessel and the biological shield was developed. The concept is focused on the safety objective, the reduction of the collective dose and the reduction of the execution time. Precondition for the decommissioning license was up to now the removal of fuel elements from the reactor; due to the significantly increased period until fulfillment of this premises concepts are developed that can be performed with simultaneous reduction of the radiological inventories and the fire loads. The most important step of the guideline of the concept is the transition from hot to cold. The in-situ disassembling of the reactor internals can be performed with decreased water level in the reactor pressure vessel, with following water treatment and complete shutdown of operational systems. This status allows an accelerated further dismantling of the plant.

  14. 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)

  15. ADDITIONAL STRESS AND FRACTURE MECHANICS ANALYSES OF PRESSURIZED WATER REACTOR PRESSURE VESSEL NOZZLES

    Energy Technology Data Exchange (ETDEWEB)

    Walter, Matthew [Structural Integrity Associates, Inc.; Yin, Shengjun [ORNL; Stevens, Gary [U.S. Nuclear Regulatory Commission; Sommerville, Daniel [Structural Integrity Associates, Inc.; Palm, Nathan [Westinghouse Electric Company, Cranberry Township, PA; Heinecke, Carol [Westinghouse Electric Company, Cranberry Township, PA

    2012-01-01

    In past years, the authors have undertaken various studies of nozzles in both boiling water reactors (BWRs) and pressurized water reactors (PWRs) located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Those studies described stress and fracture mechanics analyses performed to assess various RPV nozzle geometries, which were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-life (EOL) to require evaluation of embrittlement as part of the RPV analyses associated with pressure-temperature (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, Fracture Toughness Requirements, and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP

  16. Some Observations on Damage Tolerance Analyses in Pressure Vessels

    Science.gov (United States)

    Raju, Ivatury S.; Dawicke, David S.; Hampton, Roy W.

    2017-01-01

    AIAA standards S080 and S081 are applicable for certification of metallic pressure vessels (PV) and composite overwrap pressure vessels (COPV), respectively. These standards require damage tolerance analyses with a minimum reliable detectible flaw/crack and demonstration of safe life four times the service life with these cracks at the worst-case location in the PVs and oriented perpendicular to the maximum principal tensile stress. The standards require consideration of semi-elliptical surface cracks in the range of aspect ratios (crack depth a to half of the surface length c, i.e., (a/c) of 0.2 to 1). NASA-STD-5009 provides the minimum reliably detectible standard crack sizes (90/95 probability of detection (POD) for several non-destructive evaluation (NDE) methods (eddy current (ET), penetrant (PT), radiography (RT) and ultrasonic (UT)) for the two limits of the aspect ratio range required by the AIAA standards. This paper tries to answer the questions: can the safe life analysis consider only the life for the crack sizes at the two required limits, or endpoints, of the (a/c) range for the NDE method used or does the analysis need to consider values within that range? What would be an appropriate method to interpolate 90/95 POD crack sizes at intermediate (a/c) values? Several procedures to develop combinations of a and c within the specified range are explored. A simple linear relationship between a and c is chosen to compare the effects of seven different approaches to determine combinations of aj and cj that are between the (a/c) endpoints. Two of the seven are selected for evaluation: Approach I, the simple linear relationship, and a more conservative option, Approach III. For each of these two Approaches, the lives are computed for initial semi-elliptic crack configurations in a plate subjected to remote tensile fatigue loading with an R-ratio of 0.1, for an assumed material evaluated using NASGRO (registered 4) version 8.1. These calculations demonstrate

  17. Reactor Pressure Vessel Fracture Analysis Capabilities in Grizzly

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Backman, Marie [Univ. of Tennessee, Knoxville, TN (United States); Chakraborty, Pritam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hoffman, William [Univ. of Idaho, Moscow, ID (United States)

    2015-03-01

    Efforts have been underway to develop fracture mechanics capabilities in the Grizzly code to enable it to be used to perform deterministic fracture assessments of degraded reactor pressure vessels (RPVs). Development in prior years has resulted a capability to calculate -integrals. For this application, these are used to calculate stress intensity factors for cracks to be used in deterministic linear elastic fracture mechanics (LEFM) assessments of fracture in degraded RPVs. The -integral can only be used to evaluate stress intensity factors for axis-aligned flaws because it can only be used to obtain the stress intensity factor for pure Mode I loading. Off-axis flaws will be subjected to mixed-mode loading. For this reason, work has continued to expand the set of fracture mechanics capabilities to permit it to evaluate off-axis flaws. This report documents the following work to enhance Grizzly’s engineering fracture mechanics capabilities for RPVs: • Interaction Integral and -stress: To obtain mixed-mode stress intensity factors, a capability to evaluate interaction integrals for 2D or 3D flaws has been developed. A -stress evaluation capability has been developed to evaluate the constraint at crack tips in 2D or 3D. Initial verification testing of these capabilities is documented here. • Benchmarking for axis-aligned flaws: Grizzly’s capabilities to evaluate stress intensity factors for axis-aligned flaws have been benchmarked against calculations for the same conditions in FAVOR. • Off-axis flaw demonstration: The newly-developed interaction integral capabilities are demon- strated in an application to calculate the mixed-mode stress intensity factors for off-axis flaws. • Other code enhancements: Other enhancements to the thermomechanics capabilities that relate to the solution of the engineering RPV fracture problem are documented here.

  18. Embrittlement recovery due to annealing of reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Eason, E.D.; Wright, J.E.; Nelson, E.E. [Modeling and Computing Services, Boulder, CO (United States); Odette, G.R.; Mader, E.V. [Univ. of California, Santa Barbara, CA (United States)

    1996-03-01

    Embrittlement of reactor pressure vessels (RPVs) can be reduced by thermal annealing at temperatures higher than the normal operating conditions. Although such an annealing process has not been applied to any commercial plants in the United States, one US Army reactor, the BR3 plant in Belgium, and several plants in eastern Europe have been successfully annealed. All available Charpy annealing data were collected and analyzed in this project to develop quantitative models for estimating the recovery in 30 ft-lb (41 J) Charpy transition temperature and Charpy upper shelf energy over a range of potential annealing conditions. Pattern recognition, transformation analysis, residual studies, and the current understanding of the mechanisms involved in the annealing process were used to guide the selection of the most sensitive variables and correlating parameters and to determine the optimal functional forms for fitting the data. The resulting models were fitted by nonlinear least squares. The use of advanced tools, the larger data base now available, and insight from surrogate hardness data produced improved models for quantitative evaluation of the effects of annealing. The quality of models fitted in this project was evaluated by considering both the Charpy annealing data used for fitting and the surrogate hardness data base. The standard errors of the resulting recovery models relative to calibration data are comparable to the uncertainty in unirradiated Charpy data. This work also demonstrates that microhardness recovery is a good surrogate for transition temperature shift recovery and that there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes.

  19. The evaluation of pressure effects on the ex-vessel cooling for KNGR with MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Hwa; Park, Soo Yong; Kim, Dong Ha

    2001-03-01

    In this report, the effect of external vessel cooling on debris coolability and vessel integrity for the KNGR were examined from the two typical pressure range of high(170 bar) and low(5 bar)case using the lower plenum model in MELCOR1.8.4. As the conditions of these calculations, 80 ton of debris was relocated simultaneously into the lower vessel head and the debris relocation temperature from the core region was 2700 K. The decay heat has been assumed to be that of one hour after reactor shutdown. The creep failure of the vessel wall was simulated with 1-D model, which can consider the rapid temperature gradient over the wall thickness during the ex-vessel cooling. From the calculation results, both the coolant temperature and the total amount of coolant mass injected into the cavity are known to be the important factors in determining the time period to keep the external vessel cool. Therefore, a long-term strategy to keep the coolant temperature subcooled throughout the transient is suggested to sustain or prolong the effect of external vessel cooling. Also, it is expected that to keep the primary side at low pressure and to perform the ex-vessel flooding be the essential conditions to sustain the vessel integrity. From MELCOR, the penetration failure always occurs after relocation regardless of the RCS pressure or availability of the external vessel cooling. Therefore, It is expected that the improvement of the model for the penetration tube failure will be necessary.

  20. Photoacoustic sample vessel and method of elevated pressure operation

    Science.gov (United States)

    Autrey, Tom; Yonker, Clement R.

    2004-05-04

    An improved photoacoustic vessel and method of photoacoustic analysis. The photoacoustic sample vessel comprises an acoustic detector, an acoustic couplant, and an acoustic coupler having a chamber for holding the acoustic couplant and a sample. The acoustic couplant is selected from the group consisting of liquid, solid, and combinations thereof. Passing electromagnetic energy through the sample generates an acoustic signal within the sample, whereby the acoustic signal propagates through the sample to and through the acoustic couplant to the acoustic detector.

  1. Pre-study of dynamic loads on the internals caused by a large pipe break in a BWR; Foerstudie av stroemningsinducerade laster paa interndelar vid brott i huvudcirkulationskretsarna i BWR

    Energy Technology Data Exchange (ETDEWEB)

    Marcinkiewicz, Jerzy; Lindgren, Anders [Det Norske Veritas Nuclear Technology AB, Stockholm (Sweden)

    2002-12-01

    Det Norske Veritas Nuclear Technology has performed a literature study of dynamic load on a BWR (Boiling Water Reactor) internals caused by a large pipe break. The goal of the study was to improve the knowledge about the physics of phenomena occurring in the RPV (Reactor Pressure Vessel) after pipe break in the main circulation system and also to make a review of calculation methods, models and computer programs including their capabilities when calculating the dynamic loads. The report presents description of relevant parts of a BWR, initial and boundary conditions, and phenomena determining the loads - rapid depressurization and propagation of pressure wave (including none-equilibrium). Furthermore, the report generally describes possible methodologies for calculating the dynamic loads on internals after the pipe break and the experiences from calculations the dynamic loads with different methods (computer programs) including comparisons with experimental data. Fluid-Structure Interaction methodology and its importance for calculation of dynamic loads on reactor internals is discussed based on experimental data. A very intensive research program for studying and calculating the dynamic loads on internals after pipe breaks has been performed in USA and Germany during the seventies and the eighties. Several computer programs have been developed and a number of large-scale experiments have been performed to calibrate the calculation methods. In spite of the fact that all experiments were performed for PWR several experiences should be valid also for BWR. These experiences, connected mainly to capabilities of computer programs calculating dynamic loads, are discussed in the report.

  2. Review of pressurized thermal shock studies of large scale reactor pressure vessels in Hungary

    Directory of Open Access Journals (Sweden)

    Tamás Fekete

    2016-03-01

    Full Text Available In Hungary, four nuclear power units were constructed more than 30 years ago; they are operating to this day. In every unit, VVER-440 V213-type light-water cooled, light-water moderated, ressurized water reactors are in operation. Since the mid-1980s, numerous researches in the field of Pressurized Thermal Shock (PTS analyses of Reactor Pressure Vessels (RPVs have been conducted in Hungary; in all of them, the concept of structural integrity was the basis of research and development. During this time, four large PTS studies with industrial relevance have been completed in Hungary. Each used different objectives and guides, and the analysis methodology was also changing. This paper gives a comparative review of the methodologies used in these large PTS Structural Integrity Analysis projects, presenting the latest results as well

  3. A quick guide to API 510 certified pressure vessel inspector syllabus example questions and worked answers

    CERN Document Server

    Matthews, Clifford

    2010-01-01

    The API Individual Certification Programs (ICPs) are well established worldwide in the oil, gas, and petroleum industries. This Quick Guide is unique in providing simple, accessible and well-structured guidance for anyone studying the API 510 Certified Pressure Vessel Inspector syllabus by summarizing and helping them through the syllabus and providing multiple example questions and worked answers.Technical standards are referenced from the API 'body of knowledge' for the examination, i.e. API 510 Pressure vessel inspection, alteration, rerating; API 572 Pressure vessel inspection; API

  4. Designing of a Fleet-Leader Program for Carbon Composite Overwrapped Pressure Vessels

    Science.gov (United States)

    Murthy, Pappu L.N.; Phoenix, S. Leigh

    2009-01-01

    Composite Overwrapped Pressure Vessels (COPVs) are often used for storing pressurant gases on board spacecraft when mass saving is a prime requirement. Substantial weight savings can be achieved compared to all metallic pressure vessels. For example, on the space shuttle, replacement of all metallic pressure vessels with Kevlar COPVs resulted in a weight savings of about 30 percent. Mass critical space applications such as the Ares and Orion vehicles are currently being planned to use as many COPVs as possible in place of all-metallic pressure vessels to minimize the overall mass of the vehicle. Due to the fact that overwraps are subjected to sustained loads during long periods of a mission, stress rupture failure is a major concern. It is, therefore, important to ascertain the reliability of these vessels by analysis, since it is practically impossible to show by experimental testing the reliability of flight quality vessels. Also, it is a common practice to set aside flight quality vessels as "fleet leaders" in a test program where these vessels are subjected to slightly accelerated operating conditions so that they lead the actual flight vessels both in time and load. The intention of fleet leaders is to provide advanced warning if there is a serious design flaw in the vessels so that a major disaster in the flight vessels can be averted with advance warning. On the other hand, the accelerating conditions must be not so severe as to be prone to false alarms. The primary focus of the present paper is to provide an analytical basis for designing a viable fleet leader program for carbon COPVs. The analysis is based on a stress rupture behavior model incorporating Weibull statistics and power-law sensitivity of life to fiber stress level.

  5. DEVELOPMENT OF ASME SECTION X CODE RULES FOR HIGH PRESSURE COMPOSITE HYDROGEN PRESSURE VESSELS WITH NON-LOAD SHARING LINERS

    Energy Technology Data Exchange (ETDEWEB)

    Rawls, G.; Newhouse, N.; Rana, M.; Shelley, B.; Gorman, M.

    2010-04-13

    The Boiler and Pressure Vessel Project Team on Hydrogen Tanks was formed in 2004 to develop Code rules to address the various needs that had been identified for the design and construction of up to 15000 psi hydrogen storage vessel. One of these needs was the development of Code rules for high pressure composite vessels with non-load sharing liners for stationary applications. In 2009, ASME approved new Appendix 8, for Section X Code which contains the rules for these vessels. These vessels are designated as Class III vessels with design pressure ranging from 20.7 MPa (3,000 ps)i to 103.4 MPa (15,000 psi) and maximum allowable outside liner diameter of 2.54 m (100 inches). The maximum design life of these vessels is limited to 20 years. Design, fabrication, and examination requirements have been specified, included Acoustic Emission testing at time of manufacture. The Code rules include the design qualification testing of prototype vessels. Qualification includes proof, expansion, burst, cyclic fatigue, creep, flaw, permeability, torque, penetration, and environmental testing.

  6. 46 CFR 38.05-3 - Design and construction of pressure vessel type cargo tanks-TB/ALL.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 1 2010-10-01 2010-10-01 false Design and construction of pressure vessel type cargo... LIQUEFIED FLAMMABLE GASES Design and Installation § 38.05-3 Design and construction of pressure vessel type cargo tanks—TB/ALL. (a) Cargo tanks of pressure vessel configuration (e.g. cylindrical, spherical,...

  7. Evaluation of fast neutron fluence for Kori Unit 2 pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Young Kyou; Lim, Mi Joung; Kim, Byoung Chul; Kim, Kyung Sik [Korea Reactor Integrity Surveillance Technology, Daejeon (Korea, Republic of)

    2011-10-15

    Unit 2 at Kori reactor has been put into operation in 1983. During 24 cycle operation, five surveillance capsules at inner vessel and three ex-vessel dosimeter at cavity both are taken out for evaluation to neutron fluence. The evaluations following the surveillance program of Kori 2 unit which are required detect and prevent degradation of safety-related structures and components of the vessel. The fast (E > 1.0 MeV) neutron fluencies are necessary to estimate the fracture toughness of the pressure vessel materials. The determination of the pressure vessel neutron fluence is based on both calculations and measurements. The fluence prediction is made with a calculation, and the measurements are used to qualify the calculational methodology. Measurement-to-calculation comparisons are used to identify biases in the calculations and to provide reliable estimates of the fluence uncertainties As shown in Fig. 1, the Kori unit 2 reactor vessel surveillance programs includes the analysis of flux dosimeters contained in capsules located on the inner vessel wall at the Beltline region (0., 15., 30. and 40. Azimuth) and Ex vessel dosimeter capsules located on the cavity at connected bid chain. In this paper, the methodologies used to perform neutron transport calculations and dosimetry evaluations are described, the results of the plant specific transport calculations are given for the beltline region of Kori Unit 2 pressure vessel and the comparisons of calculations and measurements are discussed

  8. 30 CFR 56.13001 - General requirements for boilers and pressure vessels.

    Science.gov (United States)

    2010-07-01

    ... vessels. 56.13001 Section 56.13001 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL MINE SAFETY AND HEALTH SAFETY AND HEALTH STANDARDS-SURFACE METAL AND NONMETAL... standards and specifications of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code....

  9. 77 FR 23513 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

    Science.gov (United States)

    2012-04-19

    ... COMMISSION Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water... Management Criteria for PWR Reactor Vessel Internal Components.'' The original notice provided the ADAMS... published a notice requesting public comments on draft LR-ISG-2011-04, ``Updated Aging Management...

  10. Summary of Activities for Health Monitoring of Composite Overwrapped Pressure Vessels Updated January 2014

    Science.gov (United States)

    Skow, Miles G.

    2014-01-01

    This three year project (FY12-14) will design and demonstrate the ability of new Magnetic Stress Gages for the measurement of stresses on the inner diameter of a Composite Overwrapped Pressure Vessel overwrap.

  11. Workbook for predicting pressure wave and fragment effects of exploding propellant tanks and gas storage vessels

    Science.gov (United States)

    Baker, W. E.; Kulesz, J. J.; Ricker, R. E.; Bessey, R. L.; Westine, P. S.; Parr, V. B.; Oldham, G. A.

    1975-01-01

    Technology needed to predict damage and hazards from explosions of propellant tanks and bursts of pressure vessels, both near and far from these explosions is introduced. Data are summarized in graphs, tables, and nomographs.

  12. Analysis and Design of Cryogenic Pressure Vessels for Automotive Hydrogen Storage

    Science.gov (United States)

    Espinosa-Loza, Francisco Javier

    Cryogenic pressure vessels maximize hydrogen storage density by combining the high pressure (350-700 bar) typical of today's composite pressure vessels with the cryogenic temperature (as low as 25 K) typical of low pressure liquid hydrogen vessels. Cryogenic pressure vessels comprise a high-pressure inner vessel made of carbon fiber-coated metal (similar to those used for storage of compressed gas), a vacuum space filled with numerous sheets of highly reflective metalized plastic (for high performance thermal insulation), and a metallic outer jacket. High density of hydrogen storage is key to practical hydrogen-fueled transportation by enabling (1) long-range (500+ km) transportation with high capacity vessels that fit within available spaces in the vehicle, and (2) reduced cost per kilogram of hydrogen stored through reduced need for expensive structural material (carbon fiber composite) necessary to make the vessel. Low temperature of storage also leads to reduced expansion energy (by an order of magnitude or more vs. ambient temperature compressed gas storage), potentially providing important safety advantages. All this is accomplished while simultaneously avoiding fuel venting typical of cryogenic vessels for all practical use scenarios. This dissertation describes the work necessary for developing and demonstrating successive generations of cryogenic pressure vessels demonstrated at Lawrence Livermore National Laboratory. The work included (1) conceptual design, (2) detailed system design (3) structural analysis of cryogenic pressure vessels, (4) thermal analysis of heat transfer through cryogenic supports and vacuum multilayer insulation, and (5) experimental demonstration. Aside from succeeding in demonstrating a hydrogen storage approach that has established all the world records for hydrogen storage on vehicles (longest driving range, maximum hydrogen storage density, and maximum containment of cryogenic hydrogen without venting), the work also

  13. Development of a Numerical Model of Hypervelocity Impact into a Pressurized Composite Overwrapped Pressure Vessel

    Science.gov (United States)

    Garcia, M. A.; Davis, B. A.; Miller, J. E.

    2017-01-01

    As the outlook for space exploration becomes more ambitious and spacecraft travel deeper into space than ever before, it is increasingly important that propulsion systems perform reliably within the space environment. The increased reliability compels designers to increase design margin at the expense of system mass, which contrasts with the need to limit vehicle mass to maximize payload. Such are the factors that motivate the integration of high specific strength composite materials in the construction of pressure vessels commonly referred to as composite overwrapped pressure vessels (COPV). The COPV consists of a metallic liner for the inner shell of the COPV that is stiff, negates fluid permeation and serves as the anchor for composite laminates or filaments, but the liner itself cannot contain the stresses from the pressurant it contains. The compo-site-fiber reinforced polymer (CFRP) is wound around the liner using a combination of hoop (circumferential) and helical orientations. Careful consideration of wrap orientation allows the composite to evenly bear structural loading and creates the COPV's characteristic high strength to weight ratio. As the CFRP overwrap carries most of the stresses induced by pressurization, damage to the overwrap can affect mission duration, mission success and potentially cause loss-of-vehicle/loss-of-crew. For this reason, it is critical to establish a fundamental understanding of the mechanisms involved in the failure of a stressed composite such as that of the COPV. One of the greatest external threats to the integrity of a spacecraft's COPV is an impact from the meteoroid and orbital debris environments (MMOD). These impacts, even from submillimeter particles, generate extremely high stress states in the CFRP that can damage numerous fibers. As a result of this possibility, initial assumptions in survivability analysis for some human-rated NASA space-craft have assumed that any alteration of the vessel due to impact is

  14. French PWR 900 MWe pressure vessel surveillance neutron field characteristics TRIPOLI-3 calculations and experimental determination

    Energy Technology Data Exchange (ETDEWEB)

    Nimal, J.C.; Bourdet, L.; Zheng, S.H.; Vergnaud, T.; Kodeli, I. [CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie; Lloret, R.; Bevilacqua, A. [CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. des Reacteurs Experimentaux; Lefebvre, J.C. [Electricite de France (EDF), 69 - Villeurbanne (France)

    1994-12-31

    This paper presents an overview of the studies performed by CEA and EDF in the scope of the pressure vessel surveillance of the French nuclear power plants. The power plants are equipped with surveillance capsules, attached to the thermal shield. They contain the dosimeters and vessel material specimens for monitoring the effects of irradiation on the pressure vessel material. The Monte Carlo code TRIPOLI-3 is used with two nuclear data libraries to calculate the neutron flux, the steel damage and the dosimeter reaction rates, and takes into account the results of sensitivity/uncertainty calculations. 2 figs., 7 tabs., 10 refs.

  15. Analytical and computational methodology to assess the over pressures generated by a potential catastrophic failure of a cryogenic pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Zamora, I.; Fradera, J.; Jaskiewicz, F.; Lopez, D.; Hermosa, B.; Aleman, A.; Izquierdo, J.; Buskop, J.

    2014-07-01

    Idom has participated in the risk evaluation of Safety Important Class (SIC) structures due to over pressures generated by a catastrophic failure of a cryogenic pressure vessel at ITER plant site. The evaluation implements both analytical and computational methodologies achieving consistent and robust results. (Author)

  16. 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.

  17. Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Kori 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; Kim, Kwan Hyun; Hong, Joon Wha

    2007-02-15

    This report describes a neutron fluence assessment performed for the Kori Unit 1 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. After Cycle 22 of reactor operation, 2nd Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 1 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 23.

  18. 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.

  19. Final report for the 3rd Ex-Vessel Neutron Dosimetry Installations and Evaluations for Kori Unit 1 Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai (and others)

    2008-03-15

    This report describes a neutron fluence assessment performed for the Kori 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. After Cycle 23 of reactor operation, 3rd Ex-Vessel Neutron Dosimetry Program was instituted at Kori 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 24.

  20. 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.

  1. 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.

  2. Final report of the 1st ex-vessel neutron dosimetry installation and evaluations for Yonggwang 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

    2006-09-15

    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 15 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 15.

  3. 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.

  4. Detecting leaks in gas-filled pressure vessels using acoustic resonances

    Science.gov (United States)

    Gillis, K. A.; Moldover, M. R.; Mehl, J. B.

    2016-05-01

    We demonstrate that a leak from a large, unthermostatted pressure vessel into ambient air can be detected an order of magnitude more effectively by measuring the time dependence of the ratio p/f2 than by measuring the ratio p/T. Here f is the resonance frequency of an acoustic mode of the gas inside the pressure vessel, p is the pressure of the gas, and T is the kelvin temperature measured at one point in the gas. In general, the resonance frequencies are determined by a mode-dependent, weighted average of the square of the speed-of-sound throughout the volume of the gas. However, the weighting usually has a weak dependence on likely temperature gradients in the gas inside a large pressure vessel. Using the ratio p/f2, we measured a gas leak (dM/dt)/M ≈ - 1.3 × 10-5 h-1 = - 0.11 yr-1 from a 300-liter pressure vessel filled with argon at 450 kPa that was exposed to sunshine-driven temperature and pressure fluctuations as large as (dT/dt)/T ≈ (dp/dt)/p ≈ 5 × 10-2 h-1 using a 24-hour data record. This leak could not be detected in a 72-hour record of p/T. (Here M is the mass of the gas in the vessel and t is the time.)

  5. 29 CFR 1915.172 - Portable air receivers and other unfired pressure vessels.

    Science.gov (United States)

    2010-07-01

    .... 1915.172 Section 1915.172 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR (CONTINUED) OCCUPATIONAL SAFETY AND HEALTH STANDARDS FOR... Engineers Boiler and Pressure Vessel Code, Section VIII, Rules for Construction of Unfired Pressure...

  6. Could Nano-Structured Materials Enable the Improved Pressure Vessels for Deep Atmospheric Probes?

    Science.gov (United States)

    Srivastava, D.; Fuentes, A.; Bienstock, B.; Arnold, J. O.

    2005-01-01

    A viewgraph presentation on the use of Nano-Structured Materials to enable pressure vessel structures for deep atmospheric probes is shown. The topics include: 1) High Temperature/Pressure in Key X-Environments; 2) The Case for Use of Nano-Structured Materials Pressure Vessel Design; 3) Carbon based Nanomaterials; 4) Nanotube production & purification; 5) Nanomechanics of Carbon Nanotubes; 6) CNT-composites: Example (Polymer); 7) Effect of Loading sequence on Composite with 8% by volume; 8) Models for Particulate Reinforced Composites; 9) Fullerene/Ti Composite for High Strength-Insulating Layer; 10) Fullerene/Epoxy Composite for High Strength-Insulating Layer; 11) Models for Continuous Fiber Reinforced Composites; 12) Tensile Strength for Discontinuous Fiber Composite; 13) Ti + SWNT Composites: Thermal/Mechanical; 14) Ti + SWNT Composites: Tensile Strength; and 15) Nano-structured Shell for Pressure Vessels.

  7. The inclusion of weld residual stress in fracture margin assessments of embrittled nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    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{sub 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.

  8. Evaluation on Safety and Reliability of High-Pressure Vessel in Missile System

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    Through theoretical analysis of reliability and simulation analysis of dispersivi of da/dN based on Monte Carlo method, the distribution function of n and c was set up.Meanwhile, the distribution of critical opening displacement ( COD ) δ, was defined by the use of coherent coefficient method, and the probabilistic model of defects assessment of military special vessel was built. Thereby the theoretical and practical fundamental research on evaluation of reliability of military high-pressure vessels was carried out.

  9. Research Progress of Irradiation Embrittlement Behavior and Prediction Technology of Reactor Pressure Vessel Steel

    Institute of Scientific and Technical Information of China (English)

    YANG; Wen; TONG; Zhen-feng; NING; Guang-sheng; ZHANG; Chang-yi; BAI; Bing

    2015-01-01

    The reactor pressure vessel(RPV)is the core of the most important equipment in pressurized water reactor,and is the key equipment that cannot be replaced in nuclear power plant.The service life of RPV determines the use of nuclear power plant,and directly affects the safety and economy of nuclear power plant.Because of high temperature,high pressure and high-energy

  10. Managing Pressure Vessel Equipment as a Capital Asset.

    Science.gov (United States)

    Robinson, Glenn; Trombley, Robert; Shultes, Kenneth

    1999-01-01

    Argues the importance of treating facility pressure equipment as capital assets and discusses three steps in their management process. The following steps are discussed: understanding the condition of all major equipment; altering maintenance practices and procedures; and developing a long-term equipment strategy such as increased monitoring,…

  11. Progress in understanding the mechanical behavior of pressure-vessel materials at elevated temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Swindeman, R.W.; Brinkman, C.R.

    1981-01-01

    Progress during the 1970's on the production of high-temperature mechanical properties data for pressure vessel materials was reviewed. The direction of the research was toward satisfying new data requirements to implement advances in high-temperature inelastic design methods. To meet these needs, servo-controlled testing machines and high-resolution extensometry were developed to gain more information on the essential behavioral features of high-temperature alloys. The similarities and differences in the mechanical response of various pressure vessel materials were identified. High-temperature pressure vessel materials that have received the most attention included Type 304 stainless steel, Type 316 stainless steel, 2 1/4 Cr-1 Mo steel, alloy 800H, and Hastelloy X.

  12. 3D TRANSIENT COUPLED THERMO-ELASTIC-PLASTIC CONTACT SEALING ANALYSIS OF REACTOR PRESSURE VESSEL

    Institute of Scientific and Technical Information of China (English)

    Du Xuesong; Li Runfang; Lin Tengjiao

    2005-01-01

    Sealing analysis of sealing system in reactor pressure vessels is relevant with multiple nonlinear coupled-field effects, so even large-scale commercial finite element software cannot finish the complicated analysis. A fmite element method of 3D transient coupled thermo-elastic-plastic contact sealing analysis for reactor pressure vessels is presented, in which the surface nonlinearity,material nonlinearity, transient heat transfer nonlinearity and multiple coupled effect are taken into account and the sealing equation is coupling solved in iterative procedure. At the same time, a computational analysis program is developed, which is applied in the sealing analysis of experimental reactor pressure vessel, and the numerical results are in good coincidence with the experimental results. This program is also successful in analyzing the practical problem in engineering.

  13. The deterministic 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; Lu, Feng; Wang, Rongshan; Huang, Ping; Liu, Xiangbin; Zhang, Guodong; Xu, Chaoliang

    2015-07-15

    Highlights: • The conservative and non-conservative assumptions in the codes were shown. • The influence of different loads on the SM was given. • The unloading effect of the cladding was studied. • A concentrated reflection of the safety was shown based on 3-D FE analyses. - Abstract: The deterministic structural integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. While the nil-ductility-transition temperature (RT{sub NDT}) parameter is widely used, the influence of fluence and temperature distributions along the thickness of the base metal wall cannot be reflected in the comparative analysis. This paper introduces the method using a structure safety margin (SM) parameter which is based on a comparison between the material toughness (the fracture initiation toughness K{sub IC} or fracture arrest toughness K{sub Ia}) and the stress intensity factor (SIF) along the crack front for the integrity analysis of a RPV subjected to PTS transients. A 3-D finite element model is used to perform fracture mechanics analyses considering both crack initiation assessment and arrest assessment. The results show that the critical part along the crack front is always the clad-base metal interface point (IP) rather than the deepest point (DP) for either crack initiation assessment or crack arrest assessment under the thermal load. It is shown that the requirement in Regulatory Guide 1.154 that ‘axial flaws with depths less than 20% of the wall thickness and all circumferential flaws should be modeled in infinite length’ may be non-conservative. As the assessment result is often poor universal for a given material, crack and transient, caution is recommended in the safety assessment, especially for the IP. The SIF reduces under the thermal or pressure load if the map cracking (MC) effect is considered. Therefore, the assumption in the ASME and RCCM codes that the cladding should be taken into account in

  14. Summary of Activities for Health Monitoring of Composite Overwrapped Pressure Vessels

    Science.gov (United States)

    Russell, Rick; Skow, Miles

    2013-01-01

    This three-year project (FY12-14) will design and demonstrate the ability of new Magnetic Stress Gages for the measurement of stresses on the inner diameter of a Composite Overwrapped Pressure Vessel overwrap. The sensors are being tested at White Sands Testing Facility (WSTF) where the results will be correlated with a known nondestructive technique acoustic emission. The gages will be produced utilizing Meandering Winding Magnetometer (MWM) and/or MWM array eddy current technology. The ultimate goal is to utilize this technology for the health monitoring of Composite Overwrapped Pressure Vessels for all future flight programs. The first full-scale pressurization test was performed at WSTF in June 2012. The goals of this test were to determine adaptations of the magnetic stress gauge instrumentation that would be necessary to allow multiple sensors to monitor the vessel's condition simultaneously and to determine how the sensor response changes with sensor selection and orientation. The second full scale pressurization test was performed at WSTF in August 2012. The goals of this test were to monitor the vessel's condition with multiple sensors simultaneously, to determine the viability of the multiplexing units (MUX) for the application, and to determine if the sensor responses in different orientations are repeatable. For both sets of tests the vessel was pressured up to 6,000 psi to simulate maximum operating pressure. Acoustic events were observed during the first pressurization cycle. This suggested that the extended storage period prior to use of this bottle led to a relaxation of the residual stresses imparted during auto-frettage. The pressurization tests successfully demonstrated the use of multiplexers with multiple MWM arrays to monitor a vessel. It was discovered that depending upon the sensor orientation, the frequencies, and the sense element, the MWM arrays can provide a variety of complementary information about the composite overwrapped pressure

  15. New Developments in Nickel-Hydrogen Dependent Pressure Vessel (DPV) Cell and Battery Design

    Science.gov (United States)

    Caldwell, Dwight B.; Fox, Chris L.; Miller, Lee E.

    1997-01-01

    THe Dependent Pressure Vessel (DPV) Nickel-Hydrogen (NiH2) design is being developed as an advanced battery for military and commercial, aerospace and terrestrial applications. The DPV cell design offers high specific energy and energy density as well as reduced cost, while retaining the established Individual Pressure Vessel (IPV) technology flight heritage and database. This advanced DPV design also offers a more efficient mechanical, electrical and thermal cell and battery configuration and a reduced part count. The DPV battery design promotes compact, minimum volume packaging and weight efficiency, and delivers cost and weight savings with minimal design risk.

  16. Performance features of 22-cell, 19Ah single pressure vessel nickel hydrogen battery

    Science.gov (United States)

    Rao, Gopalakrishna M.; Vaidyanathan, Hari

    1996-01-01

    Two 22-cells 19Ah Nickel-Hydrogen (Ni-H2) Single Pressure Vessel (SPV) Qual batteries, one each from EPI/Joplin and EPI/Butler, were designed and procured. The two batteries differ in the cell encapsulation technology, stack preload, and activation procedure. Both the Butler and Joplin batteries met the specified requirements when subjected to qualification testing and completed 2100 and 1300 LEO cycles respectively, with nominal performance. This paper discusses advantages, design features, testing procedures, and results of the two single pressure vessel Ni-H2 batteries.

  17. Collaborative investigations of in-service irradiated material from the Japan Power Demonstration Reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, W.R.; Broadhead, B.L. [Oak Ridge National Lab., TN (United States); Suzuki, M.; Kohsaka, A. [Japan Atomic Energy Research Institute, Tokai (Japan)

    1997-02-01

    There is a need to validate the results of irradiation effects research by the examination of material taken directly from the wall of a pressure vessel that has been irradiated during normal service. Just such an evaluation is currently being conducted on material from the wall of the pressure vessel from the Japan Power Demonstration Reactor (JPDR). The research is being jointly performed at the Tokai Research Establishment of the Japan Atomic Energy Research Institute (JAERI) and by the Nuclear Regulatory Commission (NRC)-funded Heavy-Section Steel Irradiation Program at the Oak Ridge National Laboratory (ORNL).

  18. Mechanical properties of reactor pressure vessel steels studied by static and dynamic torsion tests

    Science.gov (United States)

    Munier, A.; Maamouri, M.; Schaller, R.; Mercier, O.

    1993-06-01

    Internal friction measurements and torsional plastic deformation tests have been performed in reactor pressure vessel steels (unirradiated, irradiated and irradiated/annealed specimens). The results of these experiments have been interpreted with help of transmission electron microscopy observations (conventional and in situ). It is shown how the interactions between screw dislocations and obstacles (Peierls valleys, impurities and precipitates) could explain the low temperature hardening and the irradiation embrittlement of ferritic steels. In addition, it appears that the nondestructive internal friction technique could be used advantageously to follow the evolution of the material properties under irradiation, as for instance the irradiation embrittlement of the reactor pressure vessel steels.

  19. Numerical simulation of premixed Hydrogen/air combustion pressure in a spherical vessel

    Directory of Open Access Journals (Sweden)

    Guo Han-yu

    2016-01-01

    Full Text Available In order to study the development process of hydrogen combustion in a closed vessel, an on-line chemical equilibrium calculator and a numerical simulation method would be used to analysis the combustion pressure and flame front of mixed gas, which based on 20L H2/air explosion experiments in spherical vessel (Crowl and Jo,2009. The results showed that, the turbulent model could reflect the process of combustion, and the error of combustion pressure by simulation is smaller than the Chemical Equilibrium Calculation. The heat loss and incomplete combustion are the main reason to cause the error.

  20. 46 CFR 50.30-10 - Class I, I-L and II-L pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Class I, I-L and II-L pressure vessels. 50.30-10 Section... PROVISIONS Fabrication Inspection § 50.30-10 Class I, I-L and II-L pressure vessels. (a) Classes I, I-L and II-L pressure vessels shall be subject to shop inspection at the plant where they are...

  1. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Unit Nuclear Energy, Netherlands Energy Research Foundation ECN, Petten (Netherlands)); Hoogenboorm, J.E.; De Leege, P.F.A. (International Reactor Institute IRI, University of Leiden, Leiden (Netherlands)); Van de Voet, J.; Verhagen, F.C.M. (KEMA NV, Arnhem (Netherlands))

    1992-01-01

    In order to carry out reliable reactor core calculations for a boiled water reactor (BWR) or a pressurized water reactor (PWR) first reactivity calculations have to be carried out for which several calculation programs are available. The purpose of the title project is to exchange experiences to improve the knowledge of this reactivity calculations. In a large number of institutes reactivity calculations of PWR and BWR pin cells were executed by means of available computer codes. Results are compared. It is concluded that the variations in the calculated results are problem dependent. Part of the results is satisfactory. However, further research is necessary.

  2. Evaluation of the performance of ultrasonic and eddy current testing of austenitic claddings of reactor pressure vessels; Bewertung der Aussagefaehigkeit von Ultraschall- und Wirbelstrompruefung austenitischer Plattierungen von Reaktordruckbehaeltern

    Energy Technology Data Exchange (ETDEWEB)

    Waidele, H.; Maier, H.J.; Just, T.; Seidenkranz, T.; Seydel, O.; Weiss, R.

    2000-12-01

    In the scope of this project, non-destructive testing methods were carried out on specimens with defects intentionally manufactured in the region of the cladding. The aim of these trials is an evaluation of the performance of ultrasonic and eddy current examinations of austenitic claddings of reactor pressure vessels. A review of the non-destructive testing of claddings showed that the majority of the investigations have been carried out on specimens with artificial defects (notches, holes). Therefore, for the realisation of this project MPA Stuttgart produced specimens with natural defects in the cladding. In detail these are specimens with intergranular stress-corrosion cracking, hot cracks and welding defects in the cladding as well as specimens with underclad cracks. The thickness of the specimens is about 150 mm (BWR-RPV), so that in addition to the testing from the ID (PWR, ultrasonic, eddy current) also the testing from the OD (BWR, ultrasonic) could be examined. The measurements show that most of the cladding defects can be detected with the standard ultrasonic test methods, however, in some cases generate only low echo amplitudes. Favourable results were obtained from the ID testing by means of a phased array probe, in particular in connection with the eddy current technique. Investigations on specimens containing defects not known to the inspection teams (blind tests), which will allow a further evaluation of the performance of non-destructive testing methods under realistic conditions, will be carried out in Phase II of the project. (orig.) [German] Im Rahmen dieses Vorhabens wurden an Testkoerpern mit im Plattierungsbereich gezielt eingebrachten Fehlerbildungen zerstoerungsfreie Pruefungen durchgefuehrt. Ziel der Untersuchungen ist eine Bewertung der Aussagefaehigkeit von Ultraschall- und Wirbelstrompruefungen an austenitischen Plattierungen von Reaktordruckbehaeltern. Bei einer Bestandsaufnahme zur zerstoerungsfreien Pruefung von Plattierungen zeigte

  3. The sensitivity of the burst performance of impact damaged pressure vessels to material strength properties

    Science.gov (United States)

    Lasn, K.; Vedvik, N. P.; Echtermeyer, A. T.

    2016-07-01

    This numerical study is carried out to improve the understanding of short-term residual strength of impacted composite pressure vessels. The relationship between the impact, created damage and residual strength is predicted by finite element (FE) analysis. The burst predictions depend largely on the strength properties used in the material models. However, it is typically not possible to measure all laminate properties on filament wound structures. Reasonable testing efforts are concentrated on critical properties, while obtaining other less sensitive parameters from e.g. literature. A parametric FE model is hereby employed to identify the critical strength properties, focusing on the cylindrical section of the pressure vessel. The model simulates an impactor strike on an empty vessel, which is subsequently pressurized until burst. Monte Carlo Simulations (MCS) are employed to investigate the correlations between strength related material parameters and the burst pressure. The simulations indicate the fracture toughness of the composite, hoop layer tensile strength and the yield stress of the PE liner as the most influential parameters for current vessel and impact configurations. In addition, the conservative variation in strength parameters is shown to have a rather moderate effect (COV ca. 7%) on residual burst pressures.

  4. Development of Improved Composite Pressure Vessels for Hydrogen Storage

    Energy Technology Data Exchange (ETDEWEB)

    Newhouse, Norman L. [Hexagon Lincoln, Lincoln, NE (United States)

    2016-04-29

    Hexagon Lincoln started this DOE project as part of the Hydrogen Storage Engineering Center of Excellence (HSECoE) contract on 1 February 2009. The purpose of the HSECoE was the research and development of viable material based hydrogen storage systems for on-board vehicular applications to meet DOE performance and cost targets. A baseline design was established in Phase 1. Studies were then conducted to evaluate potential improvements, such as alternate fiber, resin, and boss materials. The most promising concepts were selected such that potential improvements, compared with the baseline Hexagon Lincoln tank, resulted in a projected weight reduction of 11 percent, volume increase of 4 percent, and cost reduction of 10 percent. The baseline design was updated in Phase 2 to reflect design improvements and changes in operating conditions specified by HSECoE Partners. Evaluation of potential improvements continued during Phase 2. Subscale prototype cylinders were designed and fabricated for HSECoE Partners’ use in demonstrating their components and systems. Risk mitigation studies were conducted in Phase 3 that focused on damage tolerance of the composite reinforcement. Updated subscale prototype cylinders were designed and manufactured to better address the HSECoE Partners’ requirements for system demonstration. Subscale Type 1, Type 3, and Type 4 tanks were designed, fabricated and tested. Laboratory tests were conducted to evaluate vacuum insulated systems for cooling the tanks during fill, and maintaining low temperatures during service. Full scale designs were prepared based on results from the studies of this program. The operating conditions that developed during the program addressed adsorbent systems operating at cold temperatures. A Type 4 tank would provide the lowest cost and lightest weight, particularly at higher pressures, as long as issues with liner compatibility and damage tolerance could be resolved. A Type 1 tank might be the choice if the

  5. Application of Closed Vessel Technique for the Evaluation of Burning Rates of Propellants at Low Pressures

    Directory of Open Access Journals (Sweden)

    D. Vittal

    1977-04-01

    Full Text Available Closed vessel technique has been well established for the evaluation of burning characteristics of gun, mortar and small arms propellants at high pressures of about 750 kg/cm/sup 2/ - 3000 kg/cm/sup 2/ propellants in the pressure range up to about 200 kg/cm/sup 2/ (19.6 MPa. One of the modern trends in armaments technology is development of short range, high efficiency rockets and rocket assisted projectiles where the chamber pressure are in the range of 100 kg/cm/sup 2/ - 800 kg/cm/sup 2/ (9.8 MPa-78.5 MPa. An extension of the closed vessel technique is now presented for the measurement of rates of burning of propellants in this pressure range and a few experimental results on some conventional propellants are given.

  6. Preliminary results of steel containment vessel model test

    Energy Technology Data Exchange (ETDEWEB)

    Luk, V.K.; Hessheimer, M.F. [Sandia National Labs., Albuquerque, NM (United States); Matsumoto, T.; Komine, K.; Arai, S. [Nuclear Power Engineering Corp., Tokyo (Japan); Costello, J.F. [Nuclear Regulatory Commission, Washington, DC (United States)

    1998-04-01

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

  7. Joining dissimilar stainless steels for pressure vessel components

    Science.gov (United States)

    Sun, Zheng; Han, Huai-Yue

    1994-03-01

    A series of studies was carried out to examine the weldability and properties of dissimilar steel joints between martensitic and austenitic stainless steels - F6NM (OCr13Ni4Mo) and AISI 347, respectively. Such joints are important parts in, e.g. the primary circuit of a pressurized water reactor (PWR). This kind of joint requires both good mechanical properties, corrosion resistance and a stable magnetic permeability besides good weldability. The weldability tests included weld thermal simulation of the martensitic steel for investigating the influence of weld thermal cycles and post-weld heat treatment (PWHT) on the mechanical properties of the heat-affected zone (HAZ); implant testing for examining the tendency for cold cracking of martensitic steel; rigid restraint testing for determining hot crack susceptibility of the multi-pass dissimilar steel joints. The joints were subjected to various mechanical tests including a tensile test, bending test and impact test at various temperatures, as well as slow strain-rate test for examining the stress corrosion cracking tendency in the simulated environment of a primary circuit of a PWR. The results of various tests indicated that the quality of the tube/tube joints is satisfactory for meeting all the design requirements.

  8. Nuclear Technology. Course 30: Mechanical Inspection. Module 30-7, Pressure Vessel Inspection.

    Science.gov (United States)

    Kupiec, Chet; Espy, John

    This seventh in a series of eight modules for a course titled Mechanical Inspection is devoted to the design and fabrication of the reactor pressure vessel. The module follows a typical format that includes the following sections: (1) introduction, (2) module prerequisites, (3) objectives, (4) notes to instructor/student, (5) subject matter, (6)…

  9. 46 CFR 54.01-10 - Steam-generating pressure vessels (modifies U-1(g)).

    Science.gov (United States)

    2010-10-01

    ...)). (a) Pressure vessels in which steam is generated are classed as “Unfired Steam Boilers” except as required otherwise by paragraph (b) of this section. Unfired steam boilers must be fitted with an efficient... § 54.15-15. Unfired steam boilers must be constructed in accordance with this part other than when...

  10. 77 FR 16270 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

    Science.gov (United States)

    2012-03-20

    ... COMMISSION Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water... license renewal interim staff guidance (LR-ISG), LR-ISG-2011-04, ``Updated Aging Management Criteria for... Aging Lessons Learned (GALL) Report for the aging management of stainless steel structures...

  11. Issues of intergranular embrittlement of VVER-type nuclear reactors pressure vessel materials

    Science.gov (United States)

    Zabusov, O.

    2016-04-01

    In light of worldwide tendency to extension of service life of operating nuclear power plants - VVER-type in the first place - recently a special attention is concentrated on phenomena taking place in reactor pressure vessel materials that are able to lead to increased level of mechanical characteristics degradation (resistibility to brittle fracture) during long term of operation. Formerly the hardening mechanism of degradation (increase in the yield strength under influence of irradiation) mainly had been taken into consideration to assess pressure vessel service life limitations, but when extending the service life up to 60 years and more the non-hardening mechanism (intergranular embrittlement of the steels) must be taken into account as well. In this connection NRC “Kurchatov Institute” has initiated a number of works on investigations of this mechanism contribution to the total embrittlement of reactor pressure vessel steels. The main results of these investigations are described in this article. Results of grain boundary phosphorus concentration measurements in specimens made of first generation of VVER-type pressure vessels materials as well as VVER-1000 surveillance specimens are presented. An assessment of non-hardening mechanism contribution to the total ductile-to- brittle transition temperature shift is given.

  12. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Science.gov (United States)

    2010-01-01

    ... Requirements for thermal annealing of the reactor pressure vessel. (a) For those light water nuclear power... life of these components. (B) The effects of localized high temperatures on degradation of the concrete... thermal annealing or to operate the nuclear power reactor following the annealing must be identified....

  13. 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

  14. High-density automotive hydrogen storage with cryogenic capable pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Aceves, Salvador M.; Espinosa-Loza, Francisco; Ledesma-Orozco, Elias; Ross, Timothy O.; Weisberg, Andrew H. [Lawrence Livermore National Laboratory, P.O. Box 808, L-792, Livermore, CA 94551 (United States); Brunner, Tobias C.; Kircher, Oliver [BMW Group, Knorrstr. 147, 80788 Munich (Germany)

    2010-02-15

    LLNL is developing cryogenic capable pressure vessels with thermal endurance 5-10 times greater than conventional liquid hydrogen (LH{sub 2}) tanks that can eliminate evaporative losses in routine usage of (L)H{sub 2} automobiles. In a joint effort BMW is working on a proof of concept for a first automotive cryo-compressed hydrogen storage system that can fulfill automotive requirements on system performance, life cycle, safety and cost. Cryogenic pressure vessels can be fueled with ambient temperature compressed gaseous hydrogen (CGH{sub 2}), LH{sub 2} or cryogenic hydrogen at elevated supercritical pressure (cryo-compressed hydrogen, CcH{sub 2}). When filled with LH{sub 2} or CcH{sub 2}, these vessels contain 2-3 times more fuel than conventional ambient temperature compressed H{sub 2} vessels. LLNL has demonstrated fueling with LH{sub 2} onboard two vehicles. The generation 2 vessel, installed onboard an H{sub 2}-powered Toyota Prius and fueled with LH{sub 2} demonstrated the longest unrefueled driving distance and the longest cryogenic H{sub 2} hold time without evaporative losses. A third generation vessel will be installed, reducing weight and volume by minimizing insulation thickness while still providing acceptable thermal endurance. Based on its long experience with cryogenic hydrogen storage, BMW has developed its cryo-compressed hydrogen storage concept, which is now undergoing a thorough system and component validation to prove compliance with automotive requirements before it can be demonstrated in a BMW test vehicle. (author)

  15. Non-destructive Evaluation of Composite Pressure Vessel by Using FBG Sensors

    Institute of Scientific and Technical Information of China (English)

    HAO Jun-cai; LENG Jin-song; WEI Zhang

    2007-01-01

    In recent years, advanced composite structures are used extensively in many industries such as aerospace, aircraft, automobile,pipeline and civil engineering. Reliability and safety are crucial requirements posed by them to the advanced composite structures because of their harsh working conditions. Therefore, as a very important measure, structural health monitoring (SHM) in-service is definitely demanded for ensuring their safe working in-situ. In this paper, fiber Bragg grating (FBG) sensors are surface-mounted on the hoop and in the axial directions of a FRP pressure vessel to monitor the strain status during its pressurization. The experimental results show that the FBG sensors could be used to monitor the strain development and determine the ultimate failure strain of the composite pressure vessel.

  16. Sensitivity coefficients for the stochastic estimation of the radiation damage to the reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Alvarez, C.M.; Hernandez Valle, S. [Centro de Investigaciones Tecnologicas, Nucleares y Ambientales, La Habana (Cuba). E-mail: calvarez@ctn.isctn.edu.cu; svalle@ctn.isctn.edu.cu

    2000-07-01

    The construction of the sensitivity matrix in the case of the vessel radiation damage estimation by Monte Carlo techniques poses new problems related to the uncertainties of the obtained responses. In the case of deterministic calculations, the sensitivity coefficient obtention is a straightforward procedure based on the perturbation formalism through the calculation of the adjoint fluxes. In the paper an alternative procedure implementation based on the differential operator method is described with the modifications needed to the used HEXANN-EVALU code for the response estimations in the VVER-440 pressure vessel. (author)

  17. Dosimetry analyses of the Ringhals 3 and 4 reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Kulesza, J.A.; Fero, A.H. [Westinghouse Electric Company, Cranberry Township, PA 16066 (United States); Rouden, J.; Green, E.L. [Vattenfall/Ringhals AB, 432 85 Vaeroebacka (Sweden)

    2011-07-01

    A comprehensive series of neutron dosimetry measurements consisting of surveillance capsules, reactor pressure vessel cladding samples, and ex-vessel neutron dosimetry has been analyzed and compared to the results of three-dimensional, cycle-specific neutron transport calculations for the Ringhals Unit 3 and Unit 4 reactors in Sweden. The comparisons show excellent agreement between calculations and measurements. The measurements also demonstrate that it is possible to perform retrospective dosimetry measurements using the {sup 93}Nb (n,n') {sup 93m}Nb reaction on samples of 18-8 austenitic stainless steel with only trace amounts of elemental niobium. (authors)

  18. Uncertainties in risk assessment of hydrogen discharges from pressurized storage vessels at low temperatures

    DEFF Research Database (Denmark)

    Markert, Frank; Melideo, D.; Baraldi, D.

    2013-01-01

    Evaluations of the uncertainties resulting from risk assessment tools to predict releases from the various hydrogen storage types are important to support risk informed safety management. The tools have to predict releases from a wide range of storage pressures (up to 80 MPa) and temperatures (at...... 20K) e.g. the cryogenic compressed gas storage covers pressures up to 35 MPa and temperatures between 33K and 338 K. Accurate calculations of high pressure releases require real gas EOS. This paper compares a number of EOS to predict hydrogen properties typical in different storage types. The vessel...

  19. Fracture assessment of the Oskarshamn 1 reactor pressure vessel under cold over-pressurization

    Energy Technology Data Exchange (ETDEWEB)

    Sattari-Far, I. [DNV Technical Consulting AB, Stockholm (Sweden)

    2001-03-01

    The major motivation of this study was to develop a methodology for fracture assessment of surface defects in the 01 reactor pressure vessel under cold loading scenarios, particularly the cold over-pressurization event. According to a previous study, the FENIX project, the cold over-pressurization of the O1 reactor is a limiting loading case, as the ductile/brittle transition temperature (RT{sub NDT}) of certain welds in the O1 beltline region may be over 100 deg C at the-end-of-life condition. The FENIX project gave values of the acceptable and critical crack depth to be equal to the thickness of the cladding layer (about 6 mm) under this load case using the ASME K{sub Ic} reference curve methodology. This study is aimed to develop a methodology to give a more precise fracture assessment of the O1 reactor under cold loading scenarios. Some of the main objectives of this study have been as below: To prepare a material which can simulate the mechanical properties and RT{sub NDT} of the O1 reactor at the end-of-life conditions. To conduct a fracture mechanics test program to cover the essential influencing factors, such as crack geometry (shallow and deep cracks) and loading condition (uniaxial and biaxial) on the cleavage fracture toughness. To perform fracture mechanics analyses to identify a suitable methodology for assessment of the experimental results. To study the responses of engineering fracture assessment methods to the experimental results from the clad specimens. To propose a fracture assessment procedure for determination of the acceptable and critical flaw sizes in the 01 reactor under the cold loading events. A test program consisted of experiments on standard SEN(B) specimens and clad beams, containing surface cracks was conducted during the course of this project. A total of nine clad beams and clad cruciform specimens were tested under uniaxial and biaxial loading. The test material is reactor steel of type A 508 Grade B, which is specially heat

  20. 46 CFR 54.01-2 - Adoption of division 1 of section VIII of the ASME Boiler and Pressure Vessel Code.

    Science.gov (United States)

    2010-10-01

    ... Boiler and Pressure Vessel Code. 54.01-2 Section 54.01-2 Shipping COAST GUARD, DEPARTMENT OF HOMELAND... division 1 of section VIII of the ASME Boiler and Pressure Vessel Code. (a) Pressure vessels shall be designed, constructed, and inspected in accordance with section VIII of the ASME Boiler and Pressure...

  1. Evaluation of Agency Non-Code Layered Pressure Vessels (LPVs). Corrected Copy, Aug. 25, 2014

    Science.gov (United States)

    Prosser, William H.

    2014-01-01

    In coordination with the Office of Safety and Mission Assurance and the respective Center Pressure System Managers (PSMs), the NASA Engineering and Safety Center (NESC) was requested to formulate a consensus draft proposal for the development of additional testing and analysis methods to establish the technical validity, and any limitation thereof, for the continued safe operation of facility non-code layered pressure vessels. The PSMs from each NASA Center were asked to participate as part of the assessment team by providing, collecting, and reviewing data regarding current operations of these vessels. This report contains the outcome of the assessment and the findings, observations, and NESC recommendations to the Agency and individual NASA Centers.

  2. Analysis of deterministic and statistical approaches to fatigue crack growth in pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Francisco, Alexandre S.; Melo, P.F. Frutuoso e [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear. E-mail: frutuoso@lmn.con.ufrj.br

    2000-07-01

    This work presents three approaches to the fatigue crack growth process in steel pressure vessels as applied to failure probability calculation. In the Thomson's methodology, the crack growth is the term that represents the mechanical behavior which along the time will take the pressure vessel to a structural failure. The first result of failure probability will be obtained considering a deterministic approach, since the crack growth laws are of a deterministic nature. This approach will provide a reference value. Next, two statistical approaches will be performed based on the fact that fatigue crack growth is a random phenomenon. One of them takes into account only the variability of experimental data, proposing a distribution function to represent the failure process. The other, the stochastic approach, considers the random nature of crack growth along time, by performing the randomization of a crack growth law. The solution of this stochastic equation is a transition distribution function fitted to experimental data. (author)

  3. IAEA international studies on irradiation embrittlement of reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Brumovsky, M. [Nuclear Research Institute Rez plc (Czech Republic); Steele, L.E. [Chief Scientific Investigator of the Programme, Springfield, VA (United States)

    1997-02-01

    In last 25 years, three phases a Co-operative Research Programme on Irradiation Embrittlement of Reactor Pressure Vessel Steels has been organized by the International Atomic Energy Agency. This programme started with eight countries in 1971 and finally 16 countries took part in phase III of the Programme in 1983. Several main efforts were put into preparation of the programme, but the principal task was concentrated on an international comparison of radiation damage characterization by different laboratories for steels of {open_quotes}old{close_quotes} (with high impurity contents) and {open_quotes}advanced{close_quotes} (with low impurity contents) types as well as on development of small scale fracture mechanics procedures applicable to reactor pressure vessel surveillance programmes. This year, a new programme has been opened, concentrated mostly on small scale fracture mechanics testing.

  4. Dosimetry assessments for the reactor pressure vessel and core barrel in UK PWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Thornton, D.A.; Allen, D.A.; Huggon, A.P.; Picton, D.J.; Robinson, A.T.; Steadman, R.J. [Serco, Rutherford House, Quedgeley, Gloucester, Gl2 4NF (United Kingdom); Seren, T.; Lipponen, M.; Kekki, T. [VTT, Technical Research Centre of Finland, Otakaari 3 K, P.O. BOX 1000, Espoo, FI-02044 (Finland)

    2011-07-01

    Specimens for the Sizewell B reactor pressure vessel (RPV) inservice steels surveillance program are irradiated inside eight capsules located within the reactor pressure vessel and loaded prior to commissioning. The periodic removal of these capsules and testing of their contents provides material properties data at intervals during the lifetime of the plant. Neutron activation measurements and radiation transport calculations play an essential role in assessing the neutron exposure of the specimens and RPV. Following the most recent withdrawal, seven capsules have now been removed covering nine cycles of reactor operation. This paper summarizes the dosimetry results of the Sizewell B surveillance program obtained to date. In addition to an overview of the calculational methodology it includes a review of the measurements. Finally, it describes an extension of the methodology to provide dosimetry recommendations for the core barrel and briefly discusses the results that were obtained. (authors)

  5. Retrospective Dosimetry of Vver 440 Reactor Pressure Vessel at the 3RD Unit of Dukovany Npp

    Science.gov (United States)

    Marek, M.; Viererbl, L.; Sus, F.; Klupak, V.; Rataj, J.; Hogel, J.

    2009-08-01

    Reactor pressure vessel (RPV) residual lifetime of the Czech VVER-440 is currently monitored under Surveillance Specimens Programs (SSP) focused on reactor pressure vessel materials. Neutron fluence in the samples and its distribution in the RPV are determined by a combination of calculation results and the experimental data coming from the reactor dosimetry measurements both in the specimen containers and in the reactor cavity. The direct experimental assessment of the neutron flux density incident onto RPV and neutron fluence for the entire period of nuclear power plant unit operation can be based on the evaluation of the samples taken from the inner RPV cladding. The Retrospective Dosimetry was also used at Dukovany NPP at its 3rd unit after the 18th cycle. The paper describes methodology, experimental setup for sample extraction, measurement of activities, and the determination of the neutron flux and fluence averaged over the samples.

  6. Evaluation of Acoustic Emission NDE of Kevlar Composite Over Wrapped Pressure Vessels

    Science.gov (United States)

    Horne, Michael R.; Madaras, Eric I.

    2008-01-01

    Pressurization and failure tests of small Kevlar/epoxy COPV bottles were conducted during 2006 and 2007 by Texas Research Institute Austin, Inc., at TRI facilities. This is a report of the analysis of the Acoustic Emission (AE) data collected during those tests. Results of some of the tests indicate a possibility that AE can be used to track the stress-rupture degradation of COPV vessels.

  7. Ex-vessel Steam Explosion Analysis for Pressurized Water Reactor and Boiling Water Reactor

    OpenAIRE

    Matjaž Leskovar; Mitja Uršič

    2016-01-01

    A steam explosion may occur during a severe accident, when the molten core comes into contact with water. The pressurized water reactor and boiling water reactor ex-vessel steam explosion study, which was carried out with the multicomponent three-dimensional Eulerian fuel–coolant interaction code under the conditions of the Organisation for Economic Co-operation and Development (OECD) Steam Explosion Resolution for Nuclear Applications project reactor exercise, is presented and discussed. In ...

  8. Validation and Application of the Thermal Hydraulic System Code TRACE for Analysis of BWR Transients

    Directory of Open Access Journals (Sweden)

    V. H. Sánchez

    2012-01-01

    Full Text Available The Karlsruhe Institute of Technology (KIT is participating on (Code Applications and Maintenance Program CAMP of the US Nuclear Regulatory Commission (NRC to validate TRACE code for LWR transient analysis. The application of TRACE for the safety assessment of BWR requires a throughout verification and validation using experimental data from separate effect and integral tests but also using plant data. The validation process is normally focused on safety-relevant phenomena for example, pressure drop, void fraction, heat transfer, and critical power models. The purpose of this paper is to validate selected BWR-relevant TRACE-models using both data of bundle tests such as the (Boiling Water Reactor Full-Size Fine-Mesh Bundle Test BFBT and plant data recorded during a turbine trip event (TUSA occurred in a Type-72 German BWR plant. For the validation, TRACE models of the BFBT bundle and of the BWR plant were developed. The performed investigations have shown that the TRACE code is appropriate to describe main BWR-safety-relevant phenomena (pressure drop, void fraction, and critical power with acceptable accuracy. The comparison of the predicted global BWR plant parameters for the TUSA event with the measured plant data indicates that the code predictions are following the main trends of the measured parameters such as dome pressure and reactor power.

  9. Environmental crack-growth behavior of high strength pressure vessel alloys

    Science.gov (United States)

    Forman, R. G.

    1975-01-01

    Results of sustained-load environmental crack growth threshold tests performed on six spacecraft pressure vessel alloys are presented. The alloys were Inconel 718, 6Al-4V titanium, A-286 steel, AM-350 stainless steel, cryoformed AISI 301 stainless steel; and cryoformed AISI 304L steel. The test environments for the program were air, pressurized gases of hydrogen, oxygen, nitrogen, and carbon dioxide, and liquid environments of distilled water, sea water, nitrogen tetroxide, hydrazine, aerozine 50, monomethyl hydrazine, and hydrogen peroxide. Surface flaw type specimens were used with flaws located in both base metal and weld metal.

  10. Robinson 2 reactor vessel: pressurized thermal shock analysis for a small-break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Marston, T.; Griesbach, T.; Chao, J.; Chexal, B.; Norris, D.; Nickell, B.; Layman, B.

    1984-08-01

    A best-estimate Pressurized Thermal Shock (PTS) analysis was performed for a three-inch diameter hot-leg small-break loss-of-coolant accident for the Robinson 2 plant. This plant specific analysis was performed using EPRI's linked set of codes for PTS analysis. The analysis shows that with the H.B. Robinson 2 reactor pressure vessel, a hot-leg small-break loss-of-coolant accident does not pose a significant health or safety concern to the public for at least 40 years of operation.

  11. NUMERICAL PREDICTION OF HIGHER SELF-PRESSURIZATION RATES IN A TYPICAL STORAGE VESSEL

    Directory of Open Access Journals (Sweden)

    HARI KRISHNA RAJ

    2012-07-01

    Full Text Available Self-pressurization, as a result of vaporization can occur in many scientific and technical applications like cryogenic storage tanks, pressurized water reactors etc. Predictions of both the pressurization and vaporization rates are vital in defining design requirements conforming to the tank’s maximum working pressure andexpected liquid losses. Predicting precisely the highly transient interface phenomenon due to mass transfer coupled with phase change due to evaporation is the major challenge encountered in modeling selfpressurization. The recent improvements of the multiphase flow modeling in the ANSYS FLUENT code make it now possible to simulate these mechanisms in detail without the need of user defined functions. The volume-of-fluid (VOF method in conjunction with evaporation–condensation mass transfer model has been used here. In this paper we are extending the proven capability of VOF model for predicting higher selfpressurization rates due to phase change in storage vessels.

  12. Pressure vessels dossier restoration according to NR-13 requirements; Enquadramento de vasos de pressao a norma NR-13

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Jose L. [PETROBRAS Transporte S.A. (TRANSPETRO), Rio de Janeiro, RJ (Brazil); Goncalves, Osorio C. [PETROBRAS Transporte S.A. (TRANSPETRO), Rio de Janeiro, RJ (Brazil)

    2005-07-01

    Pressure vessels are static pressurized equipment typical in oil industry facilities. In TRANSPETRO terminals and stations as well as in the whole PETROBRAS, these equipment can be found in the form of condenser accumulators, separators, heat exchangers, storage spheres and others. Because they work sustaining pressure and, many times flammable fluids, pressure vessels have a reasonable potential for hazard. For this reason, the NR-13 regulation was created. It deals with the safety in maintenance, operation and inspection of pressure vessels and boilers. During the compliance to the NR- 13 rules, a problem usually found is the lack of documents for different reasons. In this case, the NR-13 obligates the owner to recreate the vessel documentation under the responsibility of a chartered professional. This paper presents a case study where NR-13 rules were conformed by tasks involving documentation reconstruction based on information collected by means of inspection and tests performed on the field. (author)

  13. Numerical investigations on pressurized AL-composite vessel response to hypervelocity impacts: Comparison between experimental works and a numerical code

    Directory of Open Access Journals (Sweden)

    Mespoulet Jérôme

    2015-01-01

    Full Text Available Response of pressurized composite-Al vessels to hypervelocity impact of aluminum spheres have been numerically investigated to evaluate the influence of initial pressure on the vulnerability of these vessels. Investigated tanks are carbon-fiber overwrapped prestressed Al vessels. Explored internal air pressure ranges from 1 bar to 300 bar and impact velocity are around 4400 m/s. Data obtained from experiments (Xray radiographies, particle velocity measurement and post-mortem vessels have been compared to numerical results given from LS-DYNA ALE-Lagrange-SPH full coupling models. Simulations exhibit an under estimation in term of debris cloud evolution and shock wave propagation in pressurized air but main modes of damage/rupture on the vessels given by simulations are coherent with post-mortem recovered vessels from experiments. First results of this numerical work are promising and further simulation investigations with additional experimental data will be done to increase the reliability of the simulation model. The final aim of this crossed work is to numerically explore a wide range of impact conditions (impact angle, projectile weight, impact velocity, initial pressure that cannot be explore experimentally. Those whole results will define a rule of thumbs for the definition of a vulnerability analytical model for a given pressurized vessel.

  14. Numerical investigations on pressurized AL-composite vessel response to hypervelocity impacts: Comparison between experimental works and a numerical code

    Science.gov (United States)

    Mespoulet, Jérôme; Plassard, Fabien; Hereil, Pierre-Louis

    2015-09-01

    Response of pressurized composite-Al vessels to hypervelocity impact of aluminum spheres have been numerically investigated to evaluate the influence of initial pressure on the vulnerability of these vessels. Investigated tanks are carbon-fiber overwrapped prestressed Al vessels. Explored internal air pressure ranges from 1 bar to 300 bar and impact velocity are around 4400 m/s. Data obtained from experiments (Xray radiographies, particle velocity measurement and post-mortem vessels) have been compared to numerical results given from LS-DYNA ALE-Lagrange-SPH full coupling models. Simulations exhibit an under estimation in term of debris cloud evolution and shock wave propagation in pressurized air but main modes of damage/rupture on the vessels given by simulations are coherent with post-mortem recovered vessels from experiments. First results of this numerical work are promising and further simulation investigations with additional experimental data will be done to increase the reliability of the simulation model. The final aim of this crossed work is to numerically explore a wide range of impact conditions (impact angle, projectile weight, impact velocity, initial pressure) that cannot be explore experimentally. Those whole results will define a rule of thumbs for the definition of a vulnerability analytical model for a given pressurized vessel.

  15. Cleavage Fracture Modeling of Pressure Vessels under Transient Thermo-Mechanical Loading

    Energy Technology Data Exchange (ETDEWEB)

    Qian, Xudong [National University of Singapore; Dodds, Robert [University of Illinois; Yin, Shengjun [ORNL; Bass, Bennett Richard [ORNL

    2008-02-01

    The next generation of fracture assessment procedures for nuclear reactor pressure vessels (RPVs) will combine nonlinear analyses of crack-front response with stochastic treatments of crack size, shape, orientation, location, material properties and thermal-pressure transients. The projected computational demands needed to support stochastic approaches with detailed 3-D, nonlinear stress analyses of vessels containing defects appear well beyond current and near-term capabilities. In the interim, 2-D models become appealing to approximate certain classes of critical flaws in RPVs, and have computational demands within reach for stochastic frameworks. The present work focuses on the capability of 2-D models to provide values for the Weibull stress fracture parameter with accuracy comparable to those from very detailed 3-D models. Weibull stress approaches provide one route to connect nonlinear vessel response with fracture toughness values measured using small laboratory specimens. The embedded axial flaw located in the RPV wall near the cladding-vessel interface emerges from current linear-elastic, stochastic investigations as a critical contributor to the conditional probability of initiation. Three different types of 2-D models reflecting this configuration are subjected to a thermal-pressure transient characteristic of a critical pressurized thermal shock event. The plane-strain, 2-D models include: the modified boundary layer (MBL) model, the middle tension (M(T)) model, and the 2-D RPV model. The 2-D MBL model provides a high quality estimate for the Weibull stress but only in crack-front regions with a positive T-stress. For crack-front locations with low constraint (T-stress < 0), the M(T) specimen provides very accurate Weibull stress values but only for pressure load acting alone on the RPV. For RPVs under a combined thermal-pressure transient, Weibull stresses computed from the 2-D RPV model demonstrate close agreement with those computed from the

  16. Cleavage Fracture Modeling of Pressure Vessels Under Transient Thermo-Mechanical Loading

    Energy Technology Data Exchange (ETDEWEB)

    Qian, Xudong [National University of Singapore; Dodds, Robert [University of Illinois; Yin, Shengjun [ORNL; Bass, Bennett Richard [ORNL

    2008-01-01

    Abstract The next generation of fracture assessment procedures for nuclear reactor pressure vessels (RPVs) will combine nonlinear analyses of crack-front response with stochastic treatments of crack size, shape, orientation, location, material properties and thermal-pressure transients. The projected computational demands needed to support stochastic approaches with detailed 3-D, nonlinear stress analyses of vessels containing defects appear well beyond current and near-term capabilities. In the interim, 2-D models be-come appealing to approximate certain classes of critical flaws in RPVs, and have computational demands within reach for stochastic frameworks. The present work focuses on the capability of 2-D models to provide values for the Weibull stress fracture parameter with accuracy comparable to those from very detailed 3-D models. Weibull stress approaches provide one route to connect nonlinear vessel response with fracture toughness values measured using small laboratory specimens. The embedded axial flaw located in the RPV wall near the cladding-vessel interface emerges from current linear-elastic, stochastic investigations as a critical contributor to the conditional probability of initiation. Three different types of 2-D models reflecting this configuration are subjected to a thermal-pressure transient characteristic of a critical pressurized thermal shock event. The plane-strain, 2-D models include: the modified boundary layer (MBL) model, the middle tension (M(T)) model, and the 2-D RPV model. The 2-D MBL model provides a high quality estimate for the Weibull stress but only in crack-front regions with a positive T-stress. For crack-front locations with low constraint (T-stress < 0), the M(T) specimen provides very accurate Weibull stress values but only for pressure load acting alone on the RPV. For RPVs under a combined thermal-pressure transient, Weibull stresses computed from the 2-D RPV model demonstrate close agreement with those computed from

  17. Pressure vessel deformation under in-vessel retention condition%熔融物堆内滞留条件下压力容器变形

    Institute of Scientific and Technical Information of China (English)

    温爽; 李铁萍; 李聪新; 高新力

    2016-01-01

    熔融物堆内滞留(In-Vessel Retention, IVR)已经成为第三代反应堆一项关键的严重事故缓解策略,而压力容器外部冷却(External Reactor Vessel Cooling, ERVC)技术则是保证IVR得以成功实施的关键。当发生堆芯熔化时,高温熔融物对压力容器(Reactor Pressure Vessel, RPV)下封头的热冲击会导致RPV壁面和由其构成的外部冷却通道的形状发生变化,使局部传热恶化,进而造成IVR的失效。因此,有必要对IVR条件下RPV壁面的变形进行研究。本文利用有限元软件ANSYS对RPV进行了几何建模、温度场分析和力学场分析。结果表明,在RPV外部实现冷却、内部实现泄压的前提下,壁面变形为13.85−18.75 mm。在1 MPa内压的作用下,高温蠕变会使壁面变形随时间增大,但其增量有限。热膨胀是造成壁面变形的主要因素。%Background: In-vessel retention (IVR) has become an important severe accident mitigation strategy for advanced light water reactor in recent years. The successful implementation of IVR depends on the external reactor vessel cooling (ERVC) technique. In case of core melt, the bottom head of reactor pressure vessel (RPV) becomes deformed due to the thermal impacts of high temperature, and causes the narrowing of external coolant channel which is the gap between pressure vessel outer wall and insulation layer. This phenomenon could lead to local heat transfer deterioration and then causes the failure of IVR.Purpose: The aim of this paper is to analyze the deformation of reactor pressure vessel under IVR condition.Methods: The thermal and mechanical calculations of reactor pressure vessel are performed by using the finite element methods. This work can be divided into two steps. The first step is the evaluation of the thermal field of RPV, and the second step is the calculation of stress and displacement of RPV based on its temperature fields.Results: The result shows that the maximum vertical

  18. Fabrication Flaw Density and Distribution In Repairs to Reactor Pressure Vessel and Piping Welds

    Energy Technology Data Exchange (ETDEWEB)

    GJ Schuster, FA Simonen, SR Doctor

    2008-04-01

    The Pacific Northwest National Laboratory is developing a generalized fabrication flaw distribution for the population of nuclear reactor pressure vessels and for piping welds in U.S. operating reactors. The purpose of the generalized flaw distribution is to predict component-specific flaw densities. The estimates of fabrication flaws are intended for use in fracture mechanics structural integrity assessments. Structural integrity assessments, such as estimating the frequency of loss-of-coolant accidents, are performed by computer codes that require, as input, accurate estimates of flaw densities. Welds from four different reactor pressure vessels and a collection of archived pipes have been studied to develop empirical estimates of fabrication flaw densities. This report describes the fabrication flaw distribution and characterization in the repair weld metal of vessels and piping. This work indicates that large flaws occur in these repairs. These results show that repair flaws are complex in composition and sometimes include cracks on the ends of the repair cavities. Parametric analysis using an exponential fit is performed on the data. The relevance of construction records is established for describing fabrication processes and product forms. An analysis of these records shows there was a significant change in repair frequency over the years when these components were fabricated. A description of repair flaw morphology is provided with a discussion of fracture mechanics significance. Fabrication flaws in repairs are characterized using optimized-access, high-sensitivity nondestructive ultrasonic testing. Flaw characterizations are then validated by other nondestructive evaluation techniques and complemented by destructive testing.

  19. Monitoring Composite Material Pressure Vessels with a Fiber-Optic/Microelectronic Sensor System

    Science.gov (United States)

    Klimcak, C.; Jaduszliwer, B.

    1995-01-01

    We discuss the concept of an integrated, fiber-optic/microelectronic distributed sensor system that can monitor composite material pressure vessels for Air Force space systems to provide assessments of the overall health and integrity of the vessel throughout its entire operating history from birth to end of life. The fiber optic component would include either a semiconductor light emitting diode or diode laser and a multiplexed fiber optic sensing network incorporating Bragg grating sensors capable of detecting internal temperature and strain. The microelectronic components include a power source, a pulsed laser driver, time domain data acquisition hardware, a microprocessor, a data storage device, and a communication interface. The sensing system would be incorporated within the composite during its manufacture. The microelectronic data acquisition and logging system would record the environmental conditions to which the vessel has been subjected to during its storage and transit, e.g., the history of thermal excursions, pressure loading data, the occurrence of mechanical impacts, the presence of changing internal strain due to aging, delamination, material decomposition, etc. Data would be maintained din non-volatile memory for subsequent readout through a microcomputer interface.

  20. RADIATION DOSIMETRY OF THE PRESSURE VESSEL INTERNALS OF THE HIGH FLUX BEAM REACTOR.

    Energy Technology Data Exchange (ETDEWEB)

    HOLDEN,N.E.; RECINIELLO,R.N.; HU,J.P.; RORER,D.C.

    2002-08-18

    In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, both measurements and calculations of the decay gamma-ray dose rate have been performed for the reactor pressure vessel and vessel internal structures which included the upper and lower thermal shields, the transition plate, and the control rod blades. The measurements were made using Red Perspex{trademark} polymethyl methacrylate high-level film dosimeters, a Radcal ''peanut'' ion chamber, and Eberline's high-range ion chamber. To compare with measured gamma-ray dose rate, the Monte Carlo MCNP code and geometric progressive Microshield code were used to model the gamma transport and dose buildup.

  1. Creep crack growth in a reactor pressure vessel steel at 360 deg C

    Energy Technology Data Exchange (ETDEWEB)

    Rui Wu; Seitisleam, F.; Sandstroem, R. [Swedish Institute for Metals Research, Stockholm (Sweden)

    1998-12-31

    Plain creep (PC) and creep crack growth (CCG) tests at 360 deg C and post metallography were carried out on a low alloy reactor pressure vessel steel (ASTM A508 class 2) with different microstructures. Lives for the CCG tests were shorter than those for the PC tests and this is more pronounced for simulated heat affected zone microstructure than for the parent metal at longer lives. For the CCG tests, after initiation, the cracks grew constantly and intergranularly before they accelerated to approach rupture. The creep crack growth rate is well described by C*. The relations between reference stress, failure time and steady crack growth rate are presented for the CCG tests. It is demonstrated that the failure stress due to CCG is considerably lower than the yield stress at 360 deg C. Consequently, the CCG will control the static strength of a reactor vessel. (orig.) 17 refs.

  2. UT digitized data processing for in service inspection of pressurized water reactor vessels

    Energy Technology Data Exchange (ETDEWEB)

    Lasserre, F.; Hernandez, L. [Intercontrole, Rungis (France); Paradis, L. [CEA/CEREM, 91191, Gifs/Yvette cedex (France)

    1998-03-01

    Pressurized reactor vessels in France have been examined from the inside with ultrasonic focused transducers since the very first inspection (Gagnor and Levy (1993)). The developments carried out in collaboration with the French Atomic Energy Commission (CEA) to improve the characterization of flaws detected in the body of the vessels or in the nozzles, in the vicinity of the inner or the outer surfaces now have application throughout the CIVAMIS software. The processing modules of CIVAMIS, which are implemented on site since 1994 and used by INTERCONTROLE during the in service inspections of the French PWR vessels, allow full characterization of these specific flaws. The first module is devoted to the characterization of defects located near the outer surface of the vessel or the bottom head welds (OSD module). It includes the modeling software MEPHISTOMIS which predicts the echoes coming from the interaction between the ultrasonic beam and the defects. The second module of CIVAMIS (inner surface defect module called ISD), applied to the analysis of flaws expected near the inner surface of the vessels, has been used during performance demonstration exercises on qualification mock-ups, and also on-site in five expert appraisals since its qualification in 1995. The third module available on the system has beendeveloped and qualified for the analysis of flaws likely to appear near the inner surface of the no zzles. This module, named `undercladding crack defect` (UCD) module, provides the operators with a set of pre-defined processing configurations well adapted to the characteristics of the transducers. (orig.) 11 refs.

  3. Deformation behavior in reactor pressure vessel steels as a clue to understanding irradiation hardening.

    Energy Technology Data Exchange (ETDEWEB)

    DiMelfi, R. J.; Alexander, D. E.; Rehn, L. E.

    1999-10-25

    In this paper, we examine the post-yield true stress vs true strain behavior of irradiated pressure vessel steels and iron-based alloys to reveal differences in strain-hardening behavior associated with different irradiating particles (neutrons and electrons) and different alloy chernky. It is important to understand the effects on mechanical properties caused by displacement producing radiation of nuclear reactor pressure steels. Critical embrittling effects, e.g. increases in the ductile-to-brittle-transition-temperature, are associated with irradiation-induced increases in yield strength. In addition, fatigue-life and loading-rate effects on fracture can be related to the post-irradiation strain-hardening behavior of the steels. All of these properties affect the expected service life of nuclear reactor pressure vessels. We address the characteristics of two general strengthening effects that we believe are relevant to the differing defect cluster characters produced by neutrons and electrons in four different alloys: two pressure vessel steels, A212B and A350, and two binary alloys, Fe-0.28 wt%Cu and Fe-0.74 wt%Ni. Our results show that there are differences in the post-irradiation mechanical behavior for the two kinds of irradiation and that the differences are related both to differences in damage produced and alloy chemistry. We find that while electron and neutron irradiations (at T {le} 60 C) of pressure vessel steels and binary iron-based model alloys produce similar increases in yield strength for the same dose level, they do not result in the same post-yield hardening behavior. For neutron irradiation, the true stress flow curves of the irradiated material can be made to superimpose on that of the unirradiated material, when the former are shifted appropriately along the strain axis. This behavior suggests that neutron irradiation hardening has the same effect as strain hardening for all of the materials analyzed. For electron irradiated steels, the

  4. Swedish Work on Brittle-Fracture Problems in Nuclear Reactor Pressure Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Grounes, M.

    1966-03-15

    After a short review of the part of the Swedish nuclear energy program that is of interest in this context the Swedish reactor pressure vessels and the reasoning behind the choice of materials are surveyed. Problems and desirable aims for future reactors are discussed. Much work is now being done on new types of pressure vessel steels with high strength, low transition temperature and good corrosion resistance. These steels are of the martensitic austenitic type Bofors 2RMO (13 % Cr, 6 % Ni, 1. 5 % Mo) and of the ferritic martensitic austenitic type Avesta 248 SV (16 % Cr, 5 % Ni, 1 % Mo). An applied philosophy for estimating the brittle-fracture tendency of pressure vessels is described. As a criterion of this tendency we use the crack-propagation transition temperature, e. g. as measured by the Robertson isothermal crack-arrest test. An estimate of this transition temperature at the end of the reactor' s lifetime must take increases due to fabrication, welding, geometry, ageing and irradiation into account. The transition temperature vs. stress curve moves towards higher temperatures during the reactor' s lifetime. As long as this curve does not cross the reactor vessel stress vs. temperature curve the vessel is considered safe. The magnitude of the different factors influencing the final transition temperature are discussed and data for the Marviken reactor's pressure vessel are presented. At the end of the reactor's lifetime the estimated transition temperature is 115 deg C, which is below the maximum permissible value. A program for the study of strain ageing has been initiated owing to the uncertainty as to the extent of strain ageing at low strains. A study of a simple crack-arrest test, developed in Sweden, is in progress. An extensive irradiation-effects program on several steels is in progress. Results from tests on the Swedish carbon-manganese steels 2103/R3, SIS 142103 and SIS 142102, the low-alloy steels Degerfors DE-631A, Bofors NO

  5. The criteria of fracture in the case of the leak of pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Habil; Ziliukas, A.

    1997-04-01

    In order to forecast the break of the high pressure vessels and the network of pipes in a nuclear reactor, according to the concept of leak before break of pressure vessels, it is necessary to analyze the conditions of project, production, and mounting quality as well as of exploitation. It is also necessary to evaluate the process of break by the help of the fracture criteria. In the Ignalina Nuclear Power Plant of, in Lithuania, the most important objects of investigation are: the highest pressure pipes, made of Japanese steel 19MN5 and having an anticorrosive austenitic: coal inside, the pipes of distribution, which arc made of 08X1810T steel. The steel of the network of pipes has a quality of plasticity: therefore the only criteria of fragile is impossible to apply to. The process of break would be best described by the universal criteria of elastic - plastic fracture. For this purpose the author offers the criterion of the double parameter.

  6. Preliminary investigation of an ultrasound method for estimating pressure changes in deep-positioned vessels

    Science.gov (United States)

    Olesen, Jacob Bjerring; Villagomez-Hoyos, Carlos Armando; Traberg, Marie Sand; Chee, Adrian J. Y.; Yiu, Billy Y. S.; Ho, Chung Kit; Yu, Alfred C. H.; Jensen, Jørgen Arendt

    2016-04-01

    This paper presents a method for measuring pressure changes in deep-tissue vessels using vector velocity ultrasound data. The large penetration depth is ensured by acquiring data using a low frequency phased array transducer. Vascular pressure changes are then calculated from 2-D angle-independent vector velocity fields using a model based on the Navier-Stokes equations. Experimental scans are performed on a fabricated flow phantom having a constriction of 36% at a depth of 100 mm. Scans are carried out using a phased array transducer connected to the experimental scanner, SARUS. 2-D fields of angle-independent vector velocities are acquired using directional synthetic aperture vector flow imaging. The obtained results are evaluated by comparison to a 3-D numerical simulation model with equivalent geometry as the designed phantom. The study showed pressure drops across the constricted phantom varying from -40 Pa to 15 Pa with a standard deviation of 32%, and a bias of 25% found relative to the peak simulated pressure drop. This preliminary study shows that pressure can be estimated non-invasively to a depth that enables cardiac scans, and thereby, the possibility of detecting the pressure drops across the mitral valve.

  7. The biomechanics of erections: two- versus one-compartment pressurized vessel modeling of the penis.

    Science.gov (United States)

    Mohamed, Ahmed M; Erdman, Arthur G; Timm, Gerald W

    2010-12-01

    Previous biomechanical models of the penis simulated penile erections utilizing 2D geometry, simplified 3D geometry or made inaccurate assumptions altogether. These models designed the shaft of the penis as a one-compartment pressurized vessel fixed at one end when in reality it is a two-compartment pressurized vessel in which the compartments diverge as they enter the body and are fixed at two separate anatomic sites. This study utilizes the more anatomically correct two-compartment penile model to investigate erectile function. Simplified 2D and 3D models of the erect penis were developed using the finite element method with varying anatomical considerations for analyzing structural stresses, axial buckling, and lateral deformation. This study then validated the results by building and testing corresponding physical models. Finally, a more complex and anatomically accurate model of the penis was designed and analyzed. When subject to a lateral force of 0.5 N, the peak equivalent von Mises (EVM) stress in the two-compartment model increased by about 31.62%, while in the one-compartment model, the peak EVM stress increased by as high as 70.11%. The peak EVM stress was 149 kPa in the more complex and anatomically accurate penile model. When the perforated septum was removed, the peak EVM stress increased to 455 kPa. This study verified that there is significant difference between modeling the penis as a two- versus a one-compartment pressurized vessel. When subjected to external forces, a significant advantage was exhibited by two corporal based cavernosal bodies separated by a perforated septum as opposed to one corporal body. This is due to better structural integrity of the tunica albuginea when subjected to external forces.

  8. Evaluation of dynamic fracture toughness for Yong Gwang unit 5 reactor pressure vessel materials (Baseline Tests)

    Energy Technology Data Exchange (ETDEWEB)

    Chi Se Hwan; Kim, Joo Hag; Hong, Jun Hwa; Kwon, Sun Chil; Lee, Bong Sang [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    The dynamic fracture toughness (K{sub d}) of intermediate shell and its weld in SA 508 CI. 3 Yong Gwang 5 reactor pressure vessel was determined and evaluated. Precracked thirty six Charpy specimens were tested by using an instrumented impact tester. The purpose of present work is to evaluate and confirm the un-irradiated dynamic fracture toughness and to provide pre-irradiation baseline data for future evaluation on dynamic fracture toughness change during operation. 18 refs., 5 figs., 5 tabs. (Author)

  9. Towards economic design of a pressure vessel made of duplex stainless steel

    OpenAIRE

    Veljkovic, Milan, ed. lit.; Gozzi, Jonas

    2005-01-01

    The feasibility of manufacturing pressure vessels from duplex stainless steel (grade EN 1.4462, duplex 2205: 0.02%C, 22%Cr, 5.7%Ni, 3.1%Mo, 0.17%N) was evaluated by measurements of the mechanical properties of the parent metal and welds, and by finite element modelling. The tensile stress-strain properties of 4 mm sheet, and of the weld metal and the HAZ were measured in the transverse and rolling directions. Behaviour under biaxial stress was studied by pressurising circular sheet specimens ...

  10. Improvement of remote control system of automatic ultrasonic equipment for inspection of reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Cheong, Yong Moo; Jung, H. K.; Joo, Y. S.; Koo, K. M.; Hyung, H.; Sim, C. M.; Gong, U. S.; Kim, S. H.; Lee, J. P.; Rhoo, H. C.; Kim, M. S.; Ryoo, S. K.; Choi, C. H.; Oh, K. I

    1999-12-01

    One of the important issues related to the nuclear safety is in-service inspection of reactor pressure vessel (RPV). A remote controlled automatic ultrasonic method is applied to the inspection. At present the automatic ultrasonic inspection system owned by KAERI is interrupted due to degradation of parts. In order to resume field inspection new remote control system for the equipment was designed and installed to the existing equipment. New ultrasonic sensors and their modules for RPV inspection were designed and fabricated in accordance with the new requirements of the inspection codes. Ultrasonic sensors were verified for the use in the RPV inspection. (autho0008.

  11. Cutting Technology for Decommissioning of the Reactor Pressure Vessels in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kwan Seong; Kim, Geun Ho; Moon, Jei Kwon; Choi, Byung Seon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Lots of nuclear power plants have been decommissioned during the last 2 decades. An essential part of this work is the dismantling of the Reactor Pressure Vessel and its Internals. For this purpose a wide variety of different cutting technologies have been developed, adapted and applied. A detailed introduction to Plasma Arc cutting, Contact Arc Metal cutting and Abrasive Water Suspension Jet cutting is given, as it turned out that these cutting technologies are particularly suitable for these type of segmentation work. A comparison of these technologies including gaseous emissions, cutting power, manipulator requirements as well as selected design approaches are given. Process limits as well as actual limits of application are presented

  12. Review of the Palisades pressure vessel accumulated fluence estimate and of the least squares methodology employed

    Energy Technology Data Exchange (ETDEWEB)

    Griffin, P.J.

    1998-05-01

    This report provides a review of the Palisades submittal to the Nuclear Regulatory Commission requesting endorsement of their accumulated neutron fluence estimates based on a least squares adjustment methodology. This review highlights some minor issues in the applied methodology and provides some recommendations for future work. The overall conclusion is that the Palisades fluence estimation methodology provides a reasonable approach to a {open_quotes}best estimate{close_quotes} of the accumulated pressure vessel neutron fluence and is consistent with the state-of-the-art analysis as detailed in community consensus ASTM standards.

  13. Evaluation of the risk of pressure vessel failure due to errors in the manufacturing process

    Energy Technology Data Exchange (ETDEWEB)

    Marriott, D.L. (Illinois Univ., Urbana); Beyers, C.J.E.

    1983-01-01

    A pilot study based on the construction of a small pressure vessel is described together with a general safety assessment strategy. It is concluded that the application of safety assessment to a manufacturing process is feasible, and that useful information regarding the improvement of control of such processes can be so derived. The most important contribution is considered to be the development of a systematic strategy for the identification of potential material failure mechanisms in a given manufacturing process. An alternative probability measure is proposed for evaluating risk. This is a subjective measure of the uncertainty of the available information, and has implications for the possible quantification of quality assurance activities.

  14. Inspection of dissimilar metal welds in reactor pressure vessels in Spain

    Energy Technology Data Exchange (ETDEWEB)

    Gadea, J.R.; Regidor, J.J.; Pelaez, J.A.; Serrano, P. [Tecnatom, S.A., San Sebastian de los Reyes, Madrid (Spain)

    2011-07-01

    MRP-139 recommendations for inspection of dissimilar metal (DM) welds in PWR vessels were launched in the last years in the USA. Basically, it increases the frequency of the examinations in these type of welds, with major emphasis in the hot loops, adding one intermediate inspection at the ten years interval in outlet nozzles. The spanish nuclear power plants (NPP's) have begun the implementation of this type of inspections on the vessel nozzles DM welds. As this type of inspections could have an impact in the critical path duration of the outage, it is necessary the use of a mechanical equipment able to examine the nozzles DM welds in a short vessel occupation time (VOT) with high quality, qualified techniques and minimum requirements of the refuelling platform. Tecnatom undertook the design and development of a new more advanced equipment, named TENIS-DM, for implementing the reactor pressure vessel (RPV) nozzles examination. This equipment was designed in order to accomplish the stringent requirements and the updated examination techniques; it was used for the inspection of the DM welds of Asco 1 NPP inlet and outlet nozzles in March 2011. Examination techniques and procedures were qualified through the GRUVAL validation program, based on ENIC methodology. Mechanical scanner was equipped with a large number of examination probes, and TV cameras -for visual inspection and also for monitoring the ultrasonic inspections. A remote operated submarine was also used to give support to the operational personnel during the manipulation of the equipment and its movements from one nozzle to the others. During two months before the inspection, tests of the complete inspection system were made on a nozzle mock-up installed in a 4 meters deep well at Tecnatom's facilities; this scenario was also used during the training sessions of the inspection crew. The defined technical and practical objectives were achieved: use of qualified techniques and minimal impact on the

  15. The vascular Ca2+-sensing receptor regulates blood vessel tone and blood pressure.

    Science.gov (United States)

    Schepelmann, M; Yarova, P L; Lopez-Fernandez, I; Davies, T S; Brennan, S C; Edwards, P J; Aggarwal, A; Graça, J; Rietdorf, K; Matchkov, V; Fenton, R A; Chang, W; Krssak, M; Stewart, A; Broadley, K J; Ward, D T; Price, S A; Edwards, D H; Kemp, P J; Riccardi, D

    2016-02-01

    The extracellular calcium-sensing receptor CaSR is expressed in blood vessels where its role is not completely understood. In this study, we tested the hypothesis that the CaSR expressed in vascular smooth muscle cells (VSMC) is directly involved in regulation of blood pressure and blood vessel tone. Mice with targeted CaSR gene ablation from vascular smooth muscle cells (VSMC) were generated by breeding exon 7 LoxP-CaSR mice with animals in which Cre recombinase is driven by a SM22α promoter (SM22α-Cre). Wire myography performed on Cre-negative [wild-type (WT)] and Cre-positive (SM22α)CaSR(Δflox/Δflox) [knockout (KO)] mice showed an endothelium-independent reduction in aorta and mesenteric artery contractility of KO compared with WT mice in response to KCl and to phenylephrine. Increasing extracellular calcium ion (Ca(2+)) concentrations (1-5 mM) evoked contraction in WT but only relaxation in KO aortas. Accordingly, diastolic and mean arterial blood pressures of KO animals were significantly reduced compared with WT, as measured by both tail cuff and radiotelemetry. This hypotension was mostly pronounced during the animals' active phase and was not rescued by either nitric oxide-synthase inhibition with nitro-l-arginine methyl ester or by a high-salt-supplemented diet. KO animals also exhibited cardiac remodeling, bradycardia, and reduced spontaneous activity in isolated hearts and cardiomyocyte-like cells. Our findings demonstrate a role for CaSR in the cardiovascular system and suggest that physiologically relevant changes in extracellular Ca(2+) concentrations could contribute to setting blood vessel tone levels and heart rate by directly acting on the cardiovascular CaSR.

  16. Initial Probabilistic Evaluation of Reactor Pressure Vessel Fracture with Grizzly and Raven

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hoffman, William [Univ. of Idaho, Moscow, ID (United States); Sen, Sonat [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dickson, Terry [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bass, Richard [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. The first application of Grizzly has been to study fracture in embrittled reactor pressure vessels (RPVs). Grizzly can be used to model the thermal/mechanical response of an RPV under transient conditions that would be observed in a pressurized thermal shock (PTS) scenario. The global response of the vessel provides boundary conditions for local models of the material in the vicinity of a flaw. Fracture domain integrals are computed to obtain stress intensity factors, which can in turn be used to assess whether a fracture would initiate at a pre-existing flaw. These capabilities have been demonstrated previously. A typical RPV is likely to contain a large population of pre-existing flaws introduced during the manufacturing process. This flaw population is characterized stastistically through probability density functions of the flaw distributions. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation during a transient event. This report documents initial work to perform probabilistic analysis of RPV fracture during a PTS event using a combination of the RAVEN risk analysis code and Grizzly. This work is limited in scope, considering only a single flaw with deterministic geometry, but with uncertainty introduced in the parameters that influence fracture toughness. These results are benchmarked against equivalent models run in the FAVOR code. When fully developed, the RAVEN/Grizzly methodology for modeling probabilistic fracture in RPVs will provide a general capability that can be used to consider a wider variety of vessel and flaw conditions that are difficult to consider with current tools. In addition, this will provide access to advanced probabilistic techniques provided by RAVEN, including adaptive sampling and parallelism, which can dramatically

  17. Determination of Pressure Profile During Closed-Vessel Test Through CFD Simulation

    Institute of Scientific and Technical Information of China (English)

    Ahmed Bougamra∗; Huilin Lu

    2016-01-01

    Two⁃phase flow modeling of solid propellants has great potential for simulating and predicting the ballistic parameters in closed vessel tests as well as in guns. This paper presents a numerical model describing the combustion of a solid propellant in a closed chamber and takes into account what happens in such two⁃phase, unsteady, reactive⁃flow systems. The governing equations are derived in the form of coupled, non⁃linear axisymmetric partial differential equations. The governing equations with customized parameters are implemented into Ansys Fluent 14�5. The presented solutions predict the pressure profile inside the closed chamber. The results show that the present code adequately predicts the pressure⁃time history. The numerical results are in agreement with the experimental results. Some discussions are given regarding the effect of the grain shape and the sensitivity of these predictions to the initial pressure of the solid propellant bed. The study demonstrates the capability of using the present model implemented into Fluent, to simulate the combustion of solid propellants in a closed vessel and, eventually, the interior ballistic process in guns.

  18. Low cycle fatigue behavior of a ferritic reactor pressure vessel steel

    Science.gov (United States)

    Sarkar, Apu; Kumawat, Bhupendra K.; Chakravartty, J. K.

    2015-07-01

    The cyclic stress-strain response and the low cycle fatigue (LCF) behavior of 20MnMoNi55 pressure vessel steel were studied. Tensile strength and LCF properties were examined at room temperature (RT) using specimens cut from rolling direction of a rolled block. The fully reversed strain-controlled LCF tests were conducted at a constant total strain rate with different axial strain amplitude levels. The cyclic strain-stress relationships and the strain-life relationships were obtained through the test results, and related LCF parameters of the steel were calculated. The studied steel exhibits cyclic softening behavior. Furthermore, analysis of stabilized hysteresis loops showed that the steel exhibits non-Masing behavior. Complementary scanning electron microscopy examinations were also carried out on fracture surfaces to reveal dominant damage mechanisms during crack initiation, propagation and fracture. Multiple crack initiation sites were observed on the fracture surface. The investigated LCF behavior can provide reference for pressure vessel life assessment and fracture mechanisms analysis.

  19. A stochastic-bayesian model for the fracture probability of PWR pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Francisco, Alexandre S.; Duran, Jorge Alberto R., E-mail: afrancisco@metal.eeimvr.uff.br, E-mail: duran@metal.eeimvr.uff.br [Universidade Federal Fluminense (UFF), Volta Redonda, RJ (Brazil). Dept. de Engenharia Mecanica

    2013-07-01

    Fracture probability of pressure vessels containing cracks can be obtained by methodologies of easy understanding, which require a deterministic treatment, complemented by statistical methods. However, more accurate results are required, methodologies need to be better formulated. This paper presents a new methodology to address this problem. First, a more rigorous methodology is obtained by means of the relationship of probability distributions that model crack incidence and nondestructive inspection efficiency using the Bayes' theorem. The result is an updated crack incidence distribution. Further, the accuracy of the methodology is improved by using a stochastic model for the crack growth. The stochastic model incorporates the statistical variability of the crack growth process, combining the stochastic theory with experimental data. Stochastic differential equations are derived by the randomization of empirical equations. From the solution of this equation, a distribution function related to the crack growth is derived. The fracture probability using both probability distribution functions is in agreement with theory, and presents realistic value for pressure vessels. (author)

  20. Isothermal and thermal–mechanical fatigue of VVER-440 reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Fekete, Balazs, E-mail: fekete.mm.bme@gmail.com [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary); Department of Applied Mechanics, Budapest University of Technology and Economics, Muegyetem 5, Budapest H-1111 (Hungary); Trampus, Peter [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary)

    2015-09-15

    Highlights: • We aimed to determine the thermomechanical behaviour of VVER reactor steels. • Material tests were developed and performed on GLEEBLE 3800 physical simulator. • Coffin–Manson curves and parameters were derived. • High accuracy of the strain energy based evaluation was found. • The observed dislocation evolution correlates with the mechanical behaviour. - Abstract: The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of VVER-440 reactor pressure vessel under completely reserved total strain controlled low cycle fatigue tests were investigated. An advanced test facility was developed for GLEEBLE-3800 physical simulator which was able to perform thermomechanical fatigue experiments under in-service conditions of VVER nuclear reactors. The low cycle fatigue results were evaluated with the plastic strain based Coffin–Manson law, and plastic strain energy based model as well. It was shown that both methods are able to predict the fatigue life of reactor pressure vessel steels accurately. Interrupted fatigue tests were also carried out to investigate the kinetic of the fatigue evolution of the materials. On these samples microstructural evaluation by TEM was performed. The investigated low cycle fatigue behavior can provide reference for remaining life assessment and lifetime extension analysis.

  1. Isothermal and thermal-mechanical fatigue of VVER-440 reactor pressure vessel steels

    Science.gov (United States)

    Fekete, Balazs; Trampus, Peter

    2015-09-01

    The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of VVER-440 reactor pressure vessel under completely reserved total strain controlled low cycle fatigue tests were investigated. An advanced test facility was developed for GLEEBLE-3800 physical simulator which was able to perform thermomechanical fatigue experiments under in-service conditions of VVER nuclear reactors. The low cycle fatigue results were evaluated with the plastic strain based Coffin-Manson law, and plastic strain energy based model as well. It was shown that both methods are able to predict the fatigue life of reactor pressure vessel steels accurately. Interrupted fatigue tests were also carried out to investigate the kinetic of the fatigue evolution of the materials. On these samples microstructural evaluation by TEM was performed. The investigated low cycle fatigue behavior can provide reference for remaining life assessment and lifetime extension analysis.

  2. Low cycle fatigue behavior of a ferritic reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Sarkar, Apu, E-mail: asarkar@barc.gov.in; Kumawat, Bhupendra K.; Chakravartty, J.K.

    2015-07-15

    The cyclic stress–strain response and the low cycle fatigue (LCF) behavior of 20MnMoNi55 pressure vessel steel were studied. Tensile strength and LCF properties were examined at room temperature (RT) using specimens cut from rolling direction of a rolled block. The fully reversed strain-controlled LCF tests were conducted at a constant total strain rate with different axial strain amplitude levels. The cyclic strain–stress relationships and the strain–life relationships were obtained through the test results, and related LCF parameters of the steel were calculated. The studied steel exhibits cyclic softening behavior. Furthermore, analysis of stabilized hysteresis loops showed that the steel exhibits non-Masing behavior. Complementary scanning electron microscopy examinations were also carried out on fracture surfaces to reveal dominant damage mechanisms during crack initiation, propagation and fracture. Multiple crack initiation sites were observed on the fracture surface. The investigated LCF behavior can provide reference for pressure vessel life assessment and fracture mechanisms analysis.

  3. Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards, E706(0)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2002-01-01

    1.1 This master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel (PV) and support structure steels throughout a pressure vessel's service life (Fig. 1). Some of these are existing ASTM standards, some are ASTM standards that have been modified, and some are proposed ASTM standards. General requirements of content and consistency are discussed in Section 6 . More detailed writers' and users' information, justification, and specific requirements for the nine practices, ten guides, and three methods are provided in Sections 3-5. Referenced documents are discussed in Section 2. The summary-type information that is provided in Sections 3 and 4 is essential for establishing proper understanding and communications between the writers and users of this set of matrix standards. It was extracted from the referenced documents, Section 2 and references (1-106) for use by individual writers and users. 1...

  4. Current Status of Development of High Nickel Low Alloy Steels for Commercial Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Chul; Lee, B. S.; Park, S. G.; Lee, K. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-12-15

    SA508 Gr.3 Mn-Mo-Ni low alloy steels have been used for nuclear reactor pressure vessel steels up to now. Currently, the design goal of nuclear power plant is focusing at larger capacity and longer lifetime. Requirements of much bigger pressure vessels may cause critical problems in the manufacturing stage as well as for the welding stage. Application of higher strength steel may be required to overcome the technical problems. It is known that a higher strength and fracture toughness of low alloy steels such as SA508 Gr.4N low alloy steel could be achieved by increasing the Ni and Cr contents. Therefore, SA508 Gr.4N low alloy steel is very attractive as eligible RPV steel for the next generation PWR systems. In this report, we propose the possibility of SA508 Gr.4N low alloy steel for an application of next generation commercial RPV, based on the literature research result about development history of the RPV steels and SA508 specification. In addition, we have surveyed the research result of HSLA(High Strength Low Alloy steel), which has similar chemical compositions with SA508 Gr.4N, to understand the problems and the way of improvement of SA508 Gr.4N low alloy steel. And also, we have investigated eastern RPV steel(WWER-1000), which has higher Ni contents compared to western RPV steel.

  5. Simulation of the aspersion system of the core low pressure (LPCS) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de aspersion del nucleo a baja presion (LPCS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Membrillo G, O. E.; Chavez M, C., E-mail: garzo1012@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    The present work presents the modeling and simulation of the aspersion system to low pressure of reactor of the nuclear power plant of Laguna Verde using the nuclear code RELAP/SCDAP. The objective of the emergency systems inside a nuclear reactor is the cooling of the core, nor caring the performance of any other emergency system in the case of an accident design base for coolant loss. To obtain a simulation of the system is necessary to have a model based on their main components, pipes, pumps, valves, etc. This article describes the model for the simulation of the main line and the test line for the HPCS. At the moment we have the simulation of the reactor vessel and their systems associated to the nuclear power plant of Laguna Verde, this work will allow to associate the emergency system model LPCS to the vessel model. The simulation of the vessel and the emergency systems will allow knowing the behavior of the reactor in the stage of the coolant loos, giving the possibility to analyze diverse scenarios. The general model will provide an auxiliary tool for the training in classroom and at distance in the operation of nuclear power plants. (Author)

  6. A continuum damage analysis of hydrogen attack in a 2.25Cr–1Mo pressure vessel

    NARCIS (Netherlands)

    Burg, M.W.D. van der; Giessen, E. van der; Tvergaard, V.

    1998-01-01

    A micromechanically based continuum damage model is presented to analyze the stress, temperature and hydrogen pressure dependent material degradation process termed hydrogen attack, inside a pressure vessel. Hydrogen attack (HA) is the damage process of grain boundary facets due to a chemical reacti

  7. 46 CFR 53.01-3 - Adoption of section IV of the ASME Boiler and Pressure Vessel Code.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Adoption of section IV of the ASME Boiler and Pressure...) MARINE ENGINEERING HEATING BOILERS General Requirements § 53.01-3 Adoption of section IV of the ASME Boiler and Pressure Vessel Code. (a) Heating boilers shall be designed, constructed, inspected,...

  8. 46 CFR 52.01-2 - Adoption of section I of the ASME Boiler and Pressure Vessel Code.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Adoption of section I of the ASME Boiler and Pressure...) MARINE ENGINEERING POWER BOILERS General Requirements § 52.01-2 Adoption of section I of the ASME Boiler and Pressure Vessel Code. (a) Main power boilers and auxiliary boilers shall be designed,...

  9. BWR Steam Dryer Alternating Stress Assessment Procedures

    Energy Technology Data Exchange (ETDEWEB)

    Morante, R. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hambric, S. A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Ziada, S. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2016-12-01

    This report presents an overview of Boiling Water Reactor (BWR) steam dryer design; the fatigue cracking failures that occurred at the Quad Cities (QC) plants and their root causes; a history of BWR Extended Power Uprates (EPUs) in the USA; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluations (static and alternating stress).

  10. Effect of a Sulphate Transient on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Test 1)

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H. P

    2002-03-01

    Within the CASTOC-project (5th EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. Within WP 3 of this project, the Paul Scherrer Institut (PSI) investigates the effect of water chemistry transients on the EAC crack growth behaviour under periodical partial unloading (PPU) conditions. The present report is a summary of the first PSI test of WP 3 with a Na{sub 2}SO{sub 4} transient. In the first part of the report, the theoretical background on crack growth mechanisms, crack chemistry, mass transport and water chemistry transients as well as a brief literature survey on other water chemistry transient investigations is given. Furthermore, the experimental equipment and test procedure is presented, followed by a summary of the results of PSI test 1 of WP 3. Finally the results are discussed in detail and compared to literature data. In the first part of the experiment, an actively growing EAC crack was generated by PPU in oxygenated high-temperature, high-purity water (T = 288 {sup o}C, DO = 8 ppm, SO{sub 4}{sup 2-} < 0.6 ppb). Then a sulphate transient was applied. The duration ({approx} 300 h) and the amount of sulphate (SO{sub 4}{sup 2-} = 368 ppb) of the applied sulphate transient conservatively covered all sulphate transients, which might occur in BWR/normal water chemistry (NWC) practice. After the transient, outlet conductivity was lowered from ca. 1 {mu}S/cm to less than 0.15 {mu}S/cm within 2.6 h by a 'two-loop technique'. No accelerating effect of the sulphate transient on the EAC crack growth of both tested fracture mechanics specimens under highly oxidising BWR/NWC conditions was observed, making it impossible to deterrnine incubation or delay times. The EAC crack growth rates (CGR) before, during and after the

  11. Manufacturing Cost Analysis of Novel Steel/Concrete Composite Vessel for Stationary Storage of High-Pressure Hydrogen

    Energy Technology Data Exchange (ETDEWEB)

    Feng, Zhili [ORNL; Zhang, Wei [ORNL; Wang, Jy-An John [ORNL; Ren, Fei [ORNL

    2012-09-01

    A novel, low-cost, high-pressure, steel/concrete composite vessel (SCCV) technology for stationary storage of compressed gaseous hydrogen (CGH2) is currently under development at Oak Ridge National Laboratory (ORNL) sponsored by DOE s Fuel Cell Technologies (FCT) Program. The SCCV technology uses commodity materials including structural steels and concretes for achieving cost, durability and safety requirements. In particular, the hydrogen embrittlement of high-strength low-alloy steels, a major safety and durability issue for current industry-standard pressure vessel technology, is mitigated through the use of a unique layered steel shell structure. This report presents the cost analysis results of the novel SCCV technology. A high-fidelity cost analysis tool is developed, based on a detailed, bottom-up approach which takes into account the material and labor costs involved in each of the vessel manufacturing steps. A thorough cost study is performed to understand the SCCV cost as a function of the key vessel design parameters, including hydrogen pressure, vessel dimensions, and load-carrying ratio. The major conclusions include: The SCCV technology can meet the technical/cost targets set forth by DOE s FCT Program for FY2015 and FY2020 for all three pressure levels (i.e., 160, 430 and 860 bar) relevant to the hydrogen production and delivery infrastructure. Further vessel cost reduction can benefit from the development of advanced vessel fabrication technologies such as the highly automated friction stir welding (FSW). The ORNL-patented multi-layer, multi-pass FSW can not only reduce the amount of labor needed for assembling and welding the layered steel vessel, but also make it possible to use even higher strength steels for further cost reductions and improvement of vessel structural integrity. It is noted the cost analysis results demonstrate the significant cost advantage attainable by the SCCV technology for different pressure levels when compared to the

  12. Comparison of attenuation coefficients for VVER-440 and VVER-1000 pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Marek, M.; Rataj, J.; Vandlik, S. [Reactor Physics Dept., Research Centre Rez, Husinec 130, 25068 (Czech Republic)

    2011-07-01

    The paper summarizes the attenuation coefficient of the neutron fluence with E > 0.5 MeV through a reactor pressure vessel for vodo-vodyanoi energetichesky reactor (VVER) reactor types measured and/or calculated for mock-up experiments, as well as for operated nuclear power plant (NPP) units. The attenuation coefficient is possible to evaluate directly only by using the retro-dosimetry, based on a combination of the measured activities from the weld sample and concurrent ex-vessel measurement. The available neutron fluence attenuation coefficients (E > 0.5 MeV), calculated and measured at a mock-up experiment simulating the VVER-440-unit conditions, vary from 3.5 to 6.15. A similar situation is used for the calculations and mock-up experiment measurements for the VVER-1000 RPV, where the attenuation coefficient of the neutron fluence varies from 5.99 to 8.85. Because of the difference in calculations for the real units and the mock-up experiments, the necessity to design and perform calculation benchmarks both for VVER-440 and VVER-1000 would be meaningful if the calculation model is designed adequately to a given unit. (authors)

  13. Pressure vessels design methods using the codes, fracture mechanics and multiaxial fatigue

    Directory of Open Access Journals (Sweden)

    Fatima Majid

    2016-10-01

    Full Text Available This paper gives a highlight about pressure vessel (PV methods of design to initiate new engineers and new researchers to understand the basics and to have a summary about the knowhow of PV design. This understanding will contribute to enhance their knowledge in the selection of the appropriate method. There are several types of tanks distinguished by the operating pressure, temperature and the safety system to predict. The selection of one or the other of these tanks depends on environmental regulations, the geographic location and the used materials. The design theory of PVs is very detailed in various codes and standards API, such as ASME, CODAP ... as well as the standards of material selection such as EN 10025 or EN 10028. While designing a PV, we must design the fatigue of its material through the different methods and theories, we can find in the literature, and specific codes. In this work, a focus on the fatigue lifetime calculation through fracture mechanics theory and the different methods found in the ASME VIII DIV 2, the API 579-1 and EN 13445-3, Annex B, will be detailed by giving a comparison between these methods. In many articles in the literature the uniaxial fatigue has been very detailed. Meanwhile, the multiaxial effect has not been considered as it must be. In this paper we will lead a discussion about the biaxial fatigue due to cyclic pressure in thick-walled PV. Besides, an overview of multiaxial fatigue in PVs is detailed

  14. Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de despresurizacion automatica (ADS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez G, C.; Chavez M, C., E-mail: ces.raga@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

  15. Pressure Vessel Fluence Calculations for the Hungarian VVER-440 Units for the Power Uprate and the Llifetime Extension

    Directory of Open Access Journals (Sweden)

    Hordósy Gábor

    2016-01-01

    Full Text Available A major project was launched at Paks NPP, Hungary, to investigate the possibility of lifetime extension up to 60 years. At the same time, new fuel types with higher enrichment and containing pins with gadolinium have been introduced. Due to these plans, the radiation load of the pressure vessel was evaluated up to 60 years irradiation, taking into account the past and planned future cycles. The computational procedure, elaborated and validated earlier for the fast flux calculation in the pressure vessel was modified for the new fuel types. The neutron source at the core boundaries was taken from core design calculations and the neutron transport from the source to and through the pressure vessel was followed by Monte Carlo calculations. A number of calculations were performed to adequately follow the change of the neutron source. The paper details this procedure, the used Monte Carlo model, the influence of the different reloading schemes on the radiation load and the calculated results.

  16. Pressure Vessel Fluence Calculations for the Hungarian VVER-440 Units for the Power Uprate and the Llifetime Extension

    Science.gov (United States)

    Hordósy, Gábor; Hegyi, György; Keresztúri, András; Maráczy, Csaba; Temesvári, Emese; Zsolnay, Éva M.

    2016-02-01

    A major project was launched at Paks NPP, Hungary, to investigate the possibility of lifetime extension up to 60 years. At the same time, new fuel types with higher enrichment and containing pins with gadolinium have been introduced. Due to these plans, the radiation load of the pressure vessel was evaluated up to 60 years irradiation, taking into account the past and planned future cycles. The computational procedure, elaborated and validated earlier for the fast flux calculation in the pressure vessel was modified for the new fuel types. The neutron source at the core boundaries was taken from core design calculations and the neutron transport from the source to and through the pressure vessel was followed by Monte Carlo calculations. A number of calculations were performed to adequately follow the change of the neutron source. The paper details this procedure, the used Monte Carlo model, the influence of the different reloading schemes on the radiation load and the calculated results.

  17. Application of advanced master curve approaches on WWER-440 reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, Hans-Werner [Forschungszentrum Rossendorf e.V. (Germany)]. E-mail: h.w.viehrig@fz-rossendorf.de; Scibetta, Marc [SCK-CEN, Reactor Materials Research (Belgium); Wallin, Kim [VTT Industrial Systems, Materials and Structural Integrity (Finland)

    2006-08-15

    The master curve (MC) approach used to measure the transition temperature, T , was standarised in the ASTM Standard Test Method E 1921 in 1997. The basic MC approach for analysis of fracture test results is intended for macroscopically homogeneous steels with a body centred cubic (ferritic) structure only. In reality, due to the manufacturing process, the steels in question are seldom fully macroscopically homogeneous. The fracture toughness values measured on Charpy size SE(B) specimens of base metal from the Greifswald Unit 8 rector pressure vessel (RPV) show large scatter. The basic MC evaluation following ASTM E1921 supplies a MC with many fracture toughness values which lie below the 5% fracture probability line. It is therefore suspected that this material is macroscopically inhomogeneous. In this paper, two recent extensions of the MC for inhomogeneous materials are applied to these fracture toughness data.

  18. Atomic and dislocation dynamics simulations of plastic deformation in reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Monnet, Ghiath, E-mail: ghiathmonnet@yahoo.f [EDF-R and D, MMC, Avenue des Renardieres, 77818 Moret sur Loing (France); Domain, Christophe; Queyreau, Sylvain; Naamane, Sanae [EDF-R and D, MMC, Avenue des Renardieres, 77818 Moret sur Loing (France); Devincre, Benoit [LEM, CNRS-ONERA, 29 av. de la division Leclerc, 92130 Chatillon (France)

    2009-11-15

    The collective behavior of dislocations in reactor pressure vessel (RPV) steel involves dislocation properties on different phenomenological scales. In the multiscale approach, adopted in this work, we use atomic simulations to provide input data for larger scale simulations. We show in this paper how first-principles calculations can be used to describe the Peierls potential of screw dislocations, allowing for the validation of the empirical interatomic potential used in molecular dynamics simulations. The latter are used to compute the velocity of dislocations as a function of the applied stress and the temperature. The mobility laws obtained in this way are employed in dislocation dynamics simulations in order to predict properties of plastic flow, namely dislocation-dislocation interactions and dislocation interactions with carbides at low and high temperature.

  19. Qualification of non-destructive examination for belgian nuclear reactor pressure vessel inspection

    Energy Technology Data Exchange (ETDEWEB)

    Couplet, D. [TRACTEBEL, Brussels (Belgium); Francoise, T. [Intercontrole, 94 - Rungis (France)

    2001-07-01

    In Service Inspection (ISI) participates to the assessment of Nuclear Reactor Pressure Vessel Integrity. The performance of Non Destructive Examination (NDE) techniques must be demonstrated according to predefined objectives. The qualification process is essential to trust the reliability of the information provided by NDE. In Belgian Nuclear Power Plants, the qualification was conducted through a collaboration between the vendor and a technical group from the Electricity Utility. The important facts of this qualification will be presented: - the detailed definition of the inspection and qualifications objectives, based on a combination of the ASME code and the European Methodology for Qualification; - the systematic verification of the NDE performance and limitations, for each ISI objective, through an adequate combination of tests on blocks and technical justification; - the continuous improvement of the NDE procedure; - the feedback and the lessons learnt from site experience; - the necessary multi-disciplinary approach (NDE, degradation mechanisms, structural integrity)

  20. Irradiation embrittlement of reactor pressure vessel steel outside the astm specification A508 CL2

    Science.gov (United States)

    Pachur, D.; Krawczynski, S. J.; Derz, H.; Pott, G.

    1990-04-01

    Radiation embrittlement of reactor pressure vessel steels is of considerable significance for safety engineering. Steel manufacturers must therefore comply with specifications defined by national design codes. The extent to which a steel deviating from the specification is influenced by irradiation is being examined under the German Research Programme on the Integrity of Reactor Components. Charpy-V specimens were taken from a forged steel block longitudinally and vertically to the direction of main deformation and irradiated in the FRJ-1 research reactor at a temperature of 288 °C corresponding to the operating temperature of power reactors. The neutron fluences obtained ranged between 0.8 × 10 19 and 8 × 10 19n/ cm2. Instrumented pendulum impact tests have been evaluated and the load signals measured were analysed, fitting and calculating transition temperature curves and trend curves.

  1. On flux effects in a low alloy steel from a Swedish reactor pressure vessel

    Science.gov (United States)

    Boåsen, Magnus; Efsing, Pål; Ehrnstén, Ulla

    2017-02-01

    This study aims to investigate the presence of Unstable Matrix Defects in irradiated pressure vessel steel from weldments of the Swedish PWR Ringhals 4 (R4). Hardness tests have been performed on low flux (surveillance material) and high flux (Halden reactor) irradiated material samples in combination with heat treatments at temperatures of 330, 360 and 390 °C in order to reveal eventual recovery of any hardening features induced by irradiation. The experiments carried out in this study could not reveal any hardness recovery related to Unstable Matrix Defects at relevant temperatures. However, a difference in hardness recovery was found between the low and the high flux samples at heat treatments at higher temperatures than expected for the annihilation of Unstable Matrix Defects-the observed recovery is here attributed to differences of the solute clusters formed by the high and low flux irradiations.

  2. Atomic and dislocation dynamics simulations of plastic deformation in reactor pressure vessel steel

    Science.gov (United States)

    Monnet, Ghiath; Domain, Christophe; Queyreau, Sylvain; Naamane, Sanae; Devincre, Benoit

    2009-11-01

    The collective behavior of dislocations in reactor pressure vessel (RPV) steel involves dislocation properties on different phenomenological scales. In the multiscale approach, adopted in this work, we use atomic simulations to provide input data for larger scale simulations. We show in this paper how first-principles calculations can be used to describe the Peierls potential of screw dislocations, allowing for the validation of the empirical interatomic potential used in molecular dynamics simulations. The latter are used to compute the velocity of dislocations as a function of the applied stress and the temperature. The mobility laws obtained in this way are employed in dislocation dynamics simulations in order to predict properties of plastic flow, namely dislocation-dislocation interactions and dislocation interactions with carbides at low and high temperature.

  3. Reactor pressure vessel integrity research at the Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, W.R.; Pennell, W.E.; Pace, J.V.

    1995-12-31

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is the only key safety-related component of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the integrity inherent in the RPV. For this reason, the U.S. Nuclear Regulatory Commission has established the related research programs at ORNL described herein to provide for the development and confirmation of the methods used for: (1) establishing the irradiation exposure conditions within the RPV in the Embrittlement Data Base and Dosimetry Evaluation Program, (2) assessing the effects of irradiation on the RPV materials in the Heavy-Section Steel Irradiation Program, and (3) developing overall structural and fracture analyses of RPVs in the Heavy-Section Steel Technology Program.

  4. Microstructural investigations on Russian reactor pressure vessel steels by small-angle neutron scattering

    Science.gov (United States)

    Ulbricht, A.; Boehmert, J.; Strunz, P.; Dewhurst, C.; Mathon, M.-H.

    The effect of radiation embrittlement has a high safety significance for Russian VVER reactor pressure vessel steels. Heats of base and weld metals of the as-received state, irradiated state and post-irradiation annealed state were investigated using small-angle neutron scattering (SANS) to obtain insight about the microstructural features caused by fast neutron irradiation. The SANS intensities increase in the momentum transfer range between 0.8 and 3 nm-1 for all the material compositions in the irradiated state. The size distribution function of the irradiation-induced defect clusters has a pronounced maximum at 1 nm in radius. Their content varies between 0.1 and 0.7 vol.% dependent on material composition and increases with the neutron fluence. The comparison of nuclear and magnetic scattering indicates that the defects differ in their composition. Thermal annealing reduces the volume fraction of irradiation defect clusters.

  5. Positron annihilation study of Fe-ion irradiated reactor pressure vessel model alloys

    Science.gov (United States)

    Chen, L.; Li, Z. C.; Schut, H.; Sekimura, N.

    2016-01-01

    The degradation of reactor pressure vessel steels under irradiation, which results from the hardening and embrittlement caused by a high number density of nanometer scale damage, is of increasingly crucial concern for safe nuclear power plant operation and possible reactor lifetime prolongation. In this paper, the radiation damage in model alloys with increasing chemical complexity (Fe, Fe-Cu, Fe-Cu-Si, Fe-Cu-Ni and Fe-Cu-Ni-Mn) has been studied by Positron Annihilation Doppler Broadening spectroscopy after 1.5 MeV Fe-ion implantation at room temperature or high temperature (290 oC). It is found that the room temperature irradiation generally leads to the formation of vacancy-type defects in the Fe matrix. The high temperature irradiation exhibits an additional annealing effect for the radiation damage. Besides the Cu-rich clusters observed by the positron probe, the results show formation of vacancy-Mn complexes for implantation at low temperatures.

  6. Development of an ultrasonic imaging system for the inspection of nuclear reactor pressure vessels. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Becker, F.L.; Crow, V.L.; Davis, T.J.; Doctor, S.R.; Hildebrand, B.P.; Lemon, D.K.; Posakony, G.J.

    1979-10-01

    The development of an experimental model of an ultrasonic linear array system for the inspection of weldments in nuclear reactor pressure vessels is described. The imaging system is designed to operate in both pulse echo and holographic modes of operation. The system utilizes a sequentially pulsed, phase steered linear array to develop pulse echo images and a line focused illumination transducer in conjunction with a linear receiver array to develop holographic reconstructed images. The results recorded from the computer-based system demonstrate the capability of array technology. Excellent results from both the pulse echo and holographic modes of operation have been achieved. Pulse echo images of flaws in weldments are displayed in B-scan, C-scan, or isometric presentations. Reconstruction of the phase or holographic images are compared with pulse echo results and demonstrate the enhancement potential for the holographic procedure.

  7. Statistical analysis using the Bayesian nonparametric method for irradiation embrittlement of reactor pressure vessels

    Science.gov (United States)

    Takamizawa, Hisashi; Itoh, Hiroto; Nishiyama, Yutaka

    2016-10-01

    In order to understand neutron irradiation embrittlement in high fluence regions, statistical analysis using the Bayesian nonparametric (BNP) method was performed for the Japanese surveillance and material test reactor irradiation database. The BNP method is essentially expressed as an infinite summation of normal distributions, with input data being subdivided into clusters with identical statistical parameters, such as mean and standard deviation, for each cluster to estimate shifts in ductile-to-brittle transition temperature (DBTT). The clusters typically depend on chemical compositions, irradiation conditions, and the irradiation embrittlement. Specific variables contributing to the irradiation embrittlement include the content of Cu, Ni, P, Si, and Mn in the pressure vessel steels, neutron flux, neutron fluence, and irradiation temperatures. It was found that the measured shifts of DBTT correlated well with the calculated ones. Data associated with the same materials were subdivided into the same clusters even if neutron fluences were increased.

  8. Estimation of the coolant flow through a natural circulation BWR fuel channel applying and equivalent electrical model

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, DEPFI, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Campus Morelos en IMTA, Jiutepec, Morelos (Mexico); Espinosa P, G., E-mail: julfi_ig@yahoo.com.m [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2010-10-15

    This work presents the design and implementation of an advanced controller for a reduced order model of a BWR reactor core cooled by natural circulating water, which allows real time estimates of coolant flows through fuel assemblies about standard neutron flux strings. Nuclear power plants with boiling water reactors control individual fuel assembly coolant flows by forced circulation using external or internal water pumps and different core support plate orifices. These two elements reduce flow dependency on local channel pressure drops. In BWR reactors using only natural circulation coolant flows, these two elements are not available and therefore individual channel coolant flows are highly dependent in local conditions, such as power distributions and local pressure drops. Therefore it is expected that grater uncertainties in these variables be used during safety, fuel management and other analysis, which in turns may lead to increased operation penalties, such as tighter operating limits. The objective of this work is to asses by computer simulations means to reduce uncertainties in the measurement of fuel assembly coolant flows and eventually the associated penalties. During coolant phase transitions, pressure drops and local power may alter local natural circulation through fuel assemblies and flow estimates can be helped or not by control rod moves. This work presents the construction of an optimal controller for a core flow estimator based on a reduced order model of the coolant going though the reactor vessel components and nuclear core. This model is to be driven by plant signals from standard BWR instrumentation in order to estimate the coolant flows in selected fuel assemblies about a LPRM string. For this purpose an equivalent electrical model has been mathematically developed and numerically tested. The power-flow maps of typical BRW are used as steady state references for this equivalent model. Once these were fully reproduced for steady state

  9. Spin Forming Aluminum Crew Module (CM) Metallic Aft Pressure Vessel Bulkhead (APVBH) - Phase II

    Science.gov (United States)

    Hoffman, Eric K.; Domack, Marcia S.; Torres, Pablo D.; McGill, Preston B.; Tayon, Wesley A.; Bennett, Jay E.; Murphy, Joseph T.

    2015-01-01

    The principal focus of this project was to assist the Multi-Purpose Crew Vehicle (MPCV) Program in developing a spin forming fabrication process for manufacture of the Orion crew module (CM) aft pressure vessel bulkhead. The spin forming process will enable a single piece aluminum (Al) alloy 2219 aft bulkhead resulting in the elimination of the current multiple piece welded construction, simplify CM fabrication, and lead to an enhanced design. Phase I (NASA TM-2014-218163 (1)) of this assessment explored spin forming the single-piece CM forward pressure vessel bulkhead. The Orion MPCV Program and Lockheed Martin (LM) recently made two critical decisions relative to the NESC Phase I work scope: (1) LM selected the spin forming process to manufacture a single-piece aft bulkhead for the Orion CM, and (2) the aft bulkhead will be manufactured from Al 2219. Based on the Program's new emphasis related to the spin forming process, the NESC was asked to conduct a Phase II assessment to assist in the LM manufacture of the aft bulkhead and to conduct a feasibility study into spin forming the Orion CM cone. This activity was approved on June 19, 2013. Dr. Robert Piascik, NASA Technical Fellow for Materials at the Langley Research Center (LaRC), was selected to lead this assessment. The project plan was approved by the NASA Engineering and Safety Center (NESC) Review Board (NRB) on July 18, 2013. The primary stakeholders for this assessment were the NASA and LM MPCV Program offices. Additional benefactors are commercial launch providers developing CM concepts.

  10. Rısk analysis for pressure vessel with external corrosion using RBI method based on API 581

    Science.gov (United States)

    Naubnome, Viktor; Haryadi, Gunawan Dwi; Ismail, Rifky; Kim, Seon Jin

    2016-04-01

    Internal corrosion and external are the one major cause of accidents in liquid and natural gas in a pressure vessel. To lessen the vessel risk level, many companies have adopted and applied risk based inspection (RBI) methodology to risk reduction equipment, This study applied RBI methodology to optimize the inspection planing of the pressure vessel in power plant unit Jawa-Bali. In API 581, the risk situation for each type of equipment was classified into four levels: low risk level, medium-risk level, medium-high-risk level, and high level. This is expressed as a risk matrix. In this paper, semi-quantitative analysis method of risk-based inspection (RBI) was carried out for reducing the failure level of risk and optimized inspection plans, risk analysis of equipment failures resulting from corrosion need to be implemented. The result RBI analysis showed that pressure vessel has a medium high risk level and medium level. Failure mechanisms that occur in the pressure vessel is general thinning.

  11. Neural Network Prediction of Failure of Damaged Composite Pressure Vessels from Strain Field Data Acquired by a Computer Vision Method

    Science.gov (United States)

    Russell, Samuel S.; Lansing, Matthew D.

    1997-01-01

    This effort used a new and novel method of acquiring strains called Sub-pixel Digital Video Image Correlation (SDVIC) on impact damaged Kevlar/epoxy filament wound pressure vessels during a proof test. To predict the burst pressure, the hoop strain field distribution around the impact location from three vessels was used to train a neural network. The network was then tested on additional pressure vessels. Several variations on the network were tried. The best results were obtained using a single hidden layer. SDVIC is a fill-field non-contact computer vision technique which provides in-plane deformation and strain data over a load differential. This method was used to determine hoop and axial displacements, hoop and axial linear strains, the in-plane shear strains and rotations in the regions surrounding impact sites in filament wound pressure vessels (FWPV) during proof loading by internal pressurization. The relationship between these deformation measurement values and the remaining life of the pressure vessels, however, requires a complex theoretical model or numerical simulation. Both of these techniques are time consuming and complicated. Previous results using neural network methods had been successful in predicting the burst pressure for graphite/epoxy pressure vessels based upon acoustic emission (AE) measurements in similar tests. The neural network associates the character of the AE amplitude distribution, which depends upon the extent of impact damage, with the burst pressure. Similarly, higher amounts of impact damage are theorized to cause a higher amount of strain concentration in the damage effected zone at a given pressure and result in lower burst pressures. This relationship suggests that a neural network might be able to find an empirical relationship between the SDVIC strain field data and the burst pressure, analogous to the AE method, with greater speed and simplicity than theoretical or finite element modeling. The process of testing SDVIC

  12. Effects of irradiation at lower temperature on the microstructure of Cr-Mo-V-alloyed reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Grosse, M.; Boehmert, J.; Gilles, R. [Hahn-Meitner-Institut Berlin GmbH (Germany)

    1998-10-01

    The microstructural damage process due to neutron irradiation [1] proceeds in two stages: - formation of displacement cascades - evolution of the microstructure by defect reactions. Continuing our systematic investigation about the microstructural changes of Russian reactor pressure vessel steel due to neutron irradiation the microstructure of two laboratory heats of the VVER 440-type reactor pressure vessel steel after irradiation at 60 C was studied by small angle neutron scattering (SANS). 60 C-irradiation differently changes the irradiation-induced microstructure in comparison with irradiation at reactor operation temperature and can, thus, provide new insights into the mechanisms of the irradiation damage. (orig.)

  13. Common Defects of Pressure Vessel Welding and Prevention Measures%压力容器焊接常见缺陷的产生和防治措施

    Institute of Scientific and Technical Information of China (English)

    赵振芳

    2012-01-01

    The welding quality of pressure vessels is to ensure the safe operation of a pressure vessel key.The paper introduced several kinds of common pressure vessel welding defects and preventive measures.%压力容器焊接质量是保证压力容器安全运行的关键。文章介绍了压力容器焊接中几种常见缺陷并提出预防措施。

  14. Steel Containment Vessel Model Test: Results and Evaluation

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

    A high pressure test of the steel containment vessel (SCV) model was conducted on December 11-12, 1996 at Sandia National Laboratories, Albuquerque, NM, USA. The test model is a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of an improved Mark II boiling water reactor (BWR) containment. A concentric steel contact structure (CS), installed over the SCV model and separated at a nominally uniform distance from it, provided a simplified representation of a reactor shield building in the actual plant. The SCV model and contact structure were instrumented with strain gages and displacement transducers to record the deformation behavior of the SCV model during the high pressure test. This paper summarizes the conduct and the results of the high pressure test and discusses the posttest metallurgical evaluation results on specimens removed from the SCV model.

  15. Instability and "Sausage-String" Appearance in Blood Vessels during High Blood Pressure

    CERN Document Server

    Alstrøm, P; Colding-Jorgensen, M; Gustafsson, F; Holstein-Rathlou, N H; Alstrom, Preben; Eguiluz, Victor M.; Colding-Jorgensen, Morten; Gustafsson, Finn; Holstein-Rathlou, Niels-Henrik

    1999-01-01

    A new Rayleigh-type instability is proposed to explain the `sausage-string' pattern of alternating constrictions and dilatations formed in blood vessels under influence of a vasoconstricting agent. Our theory involves the nonlinear elasticity characteristics of the vessel wall, and provides predictions for the conditions under which the cylindrical form of a blood vessel becomes unstable.

  16. Association of body composition and blood pressure categories with retinal vessel diameters in primary school children.

    Science.gov (United States)

    Imhof, Katharina; Zahner, Lukas; Schmidt-Trucksäss, Arno; Hanssen, Henner

    2016-06-01

    Alterations in retinal vessel diameters have been shown to be predictive of cardiovascular risk in adults and children. The aim of our study was to examine the association of body composition and blood pressure (BP) categories with retinal vessel diameters in school children. We examined anthropometric parameters, BP and retinal arteriolar (CRAE) and venular (CRVE) diameters as well as the arteriolar-to-venular diameter ratio (AVR) in 391 children (age: 7.3, s.d. 0.4). Differences between the lowest and highest BP quartiles indicated that higher systolic and diastolic BP were associated with narrower CRAE (P<0.001 for both). Children in the highest weight quartile had narrower CRAE compared with the lowest quartile (P=0.05). In the regression analysis, systolic and diastolic BP were associated with arteriolar narrowing (-0.4 measuring units (mu) per mm Hg, 95% confidence interval: [-0.6; -0.3] and -0.6 mu per mm Hg [-0.7; -0.4], respectively; P<0.001 for both). An independent association was found for diastolic BP only. Compared with normotensives (NT; 74.4% of cohort), arteriolar narrowing was already seen in children categorized as pre-hypertensive (PHT) (11.5% of cohort), which was similar to HT children (14.1% of cohort) (NT: mean 207.2 [205.6; 208.7] mu; PHT: 201.7 [197.8; 205.7] mu; HT: 199.7 [196.2; 203.3] mu; P=0.01 for PHT vs. NT and P<0.001 for HT vs. NT in systolic BP). Our results suggest that systolic and diastolic BP are main determinants of retinal arteriolar diameters; and therefore, microvascular health in young children. Pre-hypertension seems to be associated with retinal microvascular alterations early in life.

  17. Development of automated welding process for field fabrication of thick walled pressure vessels. Fourth quarter, FY 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-19

    Progress is reported in research on the automated welding of heavy steel plate for the fabrication of pressure vessels. Information is included on: torch and shield adaptation; mechanical control of the welding process; welding parameters; joint design; filler wire optimizaton; nondestructive testing of welds; and weld repair. (LCL)

  18. J-integral elastic plastic fracture mechanics evaluation of the stability of cracks in nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Gomez, M. P.; McMeeking, R. M.; Parks, D. M.

    1980-06-01

    Contributions were made toward developing a new methodology to assess the stability of cracks in pressure vessels made from materials that exhibit a significant increase in toughness during the early increments of crack growth. It has a wide range of validity from linear elastic to fully plastic behavior.

  19. Simulation of the Lower Head Boiling Water Reactor Vessel in a Severe Accident

    Directory of Open Access Journals (Sweden)

    Alejandro Nuñez-Carrera

    2012-01-01

    Full Text Available The objective of this paper is the simulation and analysis of the BoilingWater Reactor (BWR lower head during a severe accident. The COUPLE computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the failure of the Reactor Pressure Vessel (RPV. The release of molten material to the primary containment and the possibility of steam explosion may produce the failure of the primary containment with high radiological consequences. Then, it is important to have a detailed model in order to predict the behavior of the reactor vessel lower head in a severe accident. In this paper, a hypothetical simulation of a Loss of Coolant Accident (LOCA with simultaneous loss of off-site power and without injection of cooling water is presented with the proposal to evaluate the temperature distribution and heatup of the lower part of the RPV. The SCDAPSIM/RELAP5 3.2 code was used to build the BWR model and conduct the numerical simulation.

  20. Standard practice for examination of Gas-Filled filament-wound composite pressure vessels using acoustic emission

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This practice provides guidelines for acoustic emission (AE) examination of filament-wound composite pressure vessels, for example, the type used for fuel tanks in vehicles which use natural gas fuel. 1.2 This practice requires pressurization to a level equal to or greater than what is encountered in normal use. The tanks' pressurization history must be known in order to use this practice. Pressurization medium may be gas or liquid. 1.3 This practice is limited to vessels designed for less than 690 bar [10,000 psi] maximum allowable working pressure and water volume less than 1 m3 or 1000 L [35.4 ft3]. 1.4 AE measurements are used to detect emission sources. Other nondestructive examination (NDE) methods may be used to gain additional insight into the emission source. Procedures for other NDE methods are beyond the scope of this practice. 1.5 This practice applies to examination of new and in-service filament-wound composite pressure vessels. 1.6 This practice applies to examinations conducted at amb...

  1. BWR Assembly Optimization for Minor Actinide Recycling

    Energy Technology Data Exchange (ETDEWEB)

    G. Ivan Maldonado; John M. Christenson; J.P. Renier; T.F. Marcille; J. Casal

    2010-03-22

    The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs).

  2. 44 BWR Waste Package Loading Curve Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    J.M. Scaglione

    2001-11-05

    The objective of this calculation is to evaluate the required minimum burnup as a function of average initial boiling water reactor (BWR) assembly enrichment that would permit loading of fuel into a potential 44 BWR waste package (WP). The potential WP design is illustrated in Attachment I. The scope of this calculation covers a range of initial enrichments from 1.5 through 5.0 weight percent U-235, and a burnup range of 0 through 50 GWd/mtU.

  3. Nondestructive characterization of embrittlement in reactor pressure vessel steels -- A feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    McHenry, H.I.; Alers, G.A. [National Inst. of Standards and Technology, Boulder, CO (United States). Materials Reliability Div.

    1998-03-01

    The Nuclear Regulatory Commission recently initiated a study by NIST to assess the feasibility of using physical-property measurements for evaluating radiation embrittlement in reactor pressure vessel (RPV) steels. Ultrasonic and magnetic measurements provide the most promising approaches for nondestructive characterization of RPV steels because elastic waves and magnetic fields can sense the microstructural changes that embrittle materials. The microstructural changes of particular interest are copper precipitation hardening, which is the likely cause of radiation embrittlement in RPV steels, and the loss of dislocation mobility that is an attribute of the ductile-to-brittle transition. Measurements were made on a 1% copper steel, ASTM grade A710, in the annealed, peak-aged and overaged conditions, and on an RPV steel, ASTM grade A533B. Nonlinear ultrasonic and micromagnetic techniques were the most promising measures of precipitation hardening. Ultrasonic velocity measurements and the magnetic properties associated with hysteresis-loop measurements were not particularly sensitive to either precipitation hardening or the ductile-to-brittle transition. Measurements of internal friction using trapped ultrasonic resonance modes detected energy losses due to the motion of pinned dislocations; however, the ultrasonic attenuation associated with these measurements was small compared to the attenuation caused by beam spreading that would occur in conventional ultrasonic testing of RPVs.

  4. Effect of tempering temperature on the microstructure and mechanical properties of a reactor pressure vessel steel

    Science.gov (United States)

    Li, C. W.; Han, L. Z.; Luo, X. M.; Liu, Q. D.; Gu, J. F.

    2016-08-01

    The microstructure and mechanical properties of reactor pressure vessel (RPV) steel were investigated after tempering at different temperatures ranging from 580 to 700 °C for 5 h. With increasing tempering temperature, the impact toughness, which is qualified by Charpy V-notch total absorbed energy, initially increases from 142 to 252 J, and then decreases to 47 J, with a maximum value at 650 °C, while the ultimate tensile strength varies in exactly the opposite direction. Comparing the microstructure and fracture surfaces of different specimens, the variations in toughness and strength with the tempering temperature were generally attributed to the softening of the bainitic ferrite, the agminated Fe3C carbides that resulted from decomposition of martensite/austenite (M/A) constituents, the precipitation of Mo2C carbides, and the newly formed M/A constituents at the grain boundaries. Finally, the correlation between the impact toughness and the volume fraction of the M/A constituents was established, and the fracture mechanisms for the different tempering conditions are explained.

  5. Nanostructure evolution of neutron-irradiated reactor pressure vessel steels: Revised Object kinetic Monte Carlo model

    Science.gov (United States)

    Chiapetto, M.; Messina, L.; Becquart, C. S.; Olsson, P.; Malerba, L.

    2017-02-01

    This work presents a revised set of parameters to be used in an Object kinetic Monte Carlo model to simulate the microstructure evolution under neutron irradiation of reactor pressure vessel steels at the operational temperature of light water reactors (∼300 °C). Within a "grey-alloy" approach, a more physical description than in a previous work is used to translate the effect of Mn and Ni solute atoms on the defect cluster diffusivity reduction. The slowing down of self-interstitial clusters, due to the interaction between solutes and crowdions in Fe is now parameterized using binding energies from the latest DFT calculations and the solute concentration in the matrix from atom-probe experiments. The mobility of vacancy clusters in the presence of Mn and Ni solute atoms was also modified on the basis of recent DFT results, thereby removing some previous approximations. The same set of parameters was seen to predict the correct microstructure evolution for two different types of alloys, under very different irradiation conditions: an Fe-C-MnNi model alloy, neutron irradiated at a relatively high flux, and a high-Mn, high-Ni RPV steel from the Swedish Ringhals reactor surveillance program. In both cases, the predicted self-interstitial loop density matches the experimental solute cluster density, further corroborating the surmise that the MnNi-rich nanofeatures form by solute enrichment of immobilized small interstitial loops, which are invisible to the electron microscope.

  6. Application of the Master Curve approach for the irradiation embrittlement evaluation of pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, H.W.; Boehmert, J. [Forschungszentrum Rossendorf e.V., Inst. fuer Sicherheitsforschung, Dresden (Germany)

    2003-09-01

    The master curve (MC) approach and the associated reference temperature, T{sub 0}, as defined in the test standard ASTM E1921, is rapidly moving from the research laboratory to application in integrity assessment of components and structures. T{sub 0} is the index temperature for the universal MC, which considers the toughness behaviour of a specific material. ''The Structural Integrity Assessment Procedures for European Industry'' (SINTAP) contain a MC extension for analysing the fracture behaviour of inhomogeneous ferritic steels. This paper presents the application of the MC approach to the T{sub 0} determination of different types of Russian WWER-type reactor pressure vessel (RPV) steels. In addition the SINTAP-MC approach was applied to determine an alternative reference temperature, T{sub R}. The influence of different microstructures and compositions within one type of RPV steel and the effect of irradiation with fast neutrons on T{sub 0} are experimentally evaluated. In general the MC based T{sub 0} is about 72 K below the Charpy V-notch transition temperature related to an impact energy of 48 J. The paper demonstrates the application of MC based T{sub 0} and T{sub R} as an alternative reference temperature for neutron embrittled RPV steels used in the RPV integrity assessment. (orig.)

  7. Fracture mechanics characterisation of the WWER-440 reactor pressure vessel beltline welding seam

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, Hans-Werner [Forschungszentrum Dresden-Rossendorf, P.O. Box 510119, 01314 Dresden (Germany)], E-mail: H.W.Viehrig@fzd.de; Schuhknecht, Jan [Forschungszentrum Dresden-Rossendorf, P.O. Box 510119, 01314 Dresden (Germany)

    2009-04-15

    The master curve (MC) approach as standardised in the ASTM Standard Test Method E1921 was applied to weld metal of the reactor pressure vessel (RPV) beltline welding seam of Greifswald unit 8 RPV. Charpy size SE(B) specimens from 13 locations equally spaced over the thickness of the welding seam were tested. The orientation of the specimens within the welding seam is TL and TS according to ASTM E399. The fracture toughness values measured on the SE(B) specimens with both orientations follow the course of the master curve. Nearly all values lie within the fracture toughness curves for 2% and 98% fracture probability. There is a strong variation of the reference temperature T{sub 0} through the thickness of the welding seam, which can be explained by microstructural differences. The scatter is more pronounced for the TS SE(B) specimens. It can be shown that specimens with TL and TS orientation in the welding seam have a differentiating and integrating behaviour, respectively.

  8. Nondestructive Methods and Special Test Instrumentation Supporting NASA Composite Overwrapped Pressure Vessel Assessments

    Science.gov (United States)

    Saulsberry, Regor; Greene, Nathanael; Cameron, Ken; Madaras, Eric; Grimes-Ledesma, Lorie; Thesken, John; Phoenix, Leigh; Murthy, Pappu; Revilock, Duane

    2007-01-01

    Many aging composite overwrapped pressure vessels (COPVs), being used by the National Aeronautics and Space Administration (NASA) are currently under evaluation to better quantify their reliability and clarify their likelihood of failure due to stress rupture and age-dependent issues. As a result, some test and analysis programs have been successfully accomplished and other related programs are still in progress at the NASA Johnson Space Center (JSC) White Sands Test Facility (WSTF) and other NASA centers, with assistance from the commercial sector. To support this effort, a group of Nondestructive Evaluation (NDE) experts was assembled to provide NDE competence for pretest evaluation of test articles and for application of NDE technology to real-time testing. Techniques were required to provide assurance that the test article had adequate structural integrity and manufacturing consistency to be considered acceptable for testing and these techniques were successfully applied. Destructive testing is also being accomplished to better understand the physical and chemical property changes associated with progression toward "stress rupture" (SR) failure, and it is being associated with NDE response, so it can potentially be used to help with life prediction. Destructive work also includes the evaluation of residual stresses during dissection of the overwrap, laboratory evaluation of specimens extracted from the overwrap to evaluate physical property changes, and quantitative microscopy to inform the theoretical micromechanics.

  9. Analysis for the Effect of Spatial Discretization Method on AP1000 Reactor Pressure Vessel Fluence Calculation

    Directory of Open Access Journals (Sweden)

    Junxiao Zheng

    2016-01-01

    Full Text Available Maintaining the structural integrity of the reactor pressure vessel (RPV is a critical concern related to the safe operation of nuclear power plants. To estimate the structural integrity over the designed lifetime and to support analyses for a potential plant life extension, an accurate calculation of the fast neutron fluence (E>1.0 MeV or E>0.1 MeV at the RPV is significant. The discrete ordinates method is one of the main methods to solve such problems. During the calculation process, many factors will affect the results. In this paper, the deviations introduced by different differencing schemes and mesh sizes on the AP1000 RPV fast neutron fluence have been studied, which are based on new discrete ordinates code ARES. The analysis shows that the differencing scheme (diamond difference with or without linear zero fix-up, theta weighted, directional theta weighted, and exponential directional weighted introduces a deviation within 4%. The coarse mesh (4 × 4 cm meshes in XY plane leads to approximately 23.7% calculation deviation compared to those of refined mesh (1 × 1 cm meshes in XY plane. Comprehensive study on the deviation introduced by differencing scheme and mesh size has great significance for reasoned evaluation of RPV fast neutron fluence calculation results.

  10. Effects of thermal annealing and reirradiation on toughness of reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Nanstad, R.K.; Iskander, S.K.; Sokolov, M.A. [Oak Ridge National Lab., TN (United States)] [and others

    1997-02-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response, or {open_quotes}recovery,{close_quotes} of several irradiated RPV steels; it also includes recent results from both ORNL and the Russian Research Center-Kurchatov Institute (RRC-KI) on a cooperative program of irradiation, annealing and reirradiation of both U.S. and Russian RPV steels. The cooperative program was conducted under the auspices of Working Group 3, U.S./Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS). The materials investigated are an RPV plate and various submerged-arc welds, with tensile, Charpy impact toughness, and fracture toughness results variously determined. Experimental results are compared with applicable prediction guidelines, while observed differences in annealing responses and reirradiation rates are discussed.

  11. Interactions between dislocations and irradiation-induced defects in light water reactor pressure vessel steels

    Science.gov (United States)

    Jumel, Stéphanie; Van Duysen, Jean-Claude; Ruste, Jacky; Domain, Christophe

    2005-11-01

    The REVE project (REactor for Virtual Experiments) is an international effort aimed at developing tools to simulate irradiation effects in light water reactors materials. In the framework of this project, a European team developed a first tool, called RPV-1 designed for reactor pressure vessel steels. This article is the third of a series dedicated to the presentation of the codes and models used to build RPV-1. It describes the simplified approach adopted to simulate the irradiation-induced hardening. This approach relies on a characterization of the interactions between a screw dislocation and irradiation-induced defects from molecular dynamics simulations. The pinning forces exerted by the defects on the dislocation were estimated from the obtained results and some hypotheses. In RPV-1, these forces are used as input parameters of a Foreman and Makin-type code, called DUPAIR, to simulate the irradiation-induced hardening at 20 °C. The relevance of the proposed approach was validated by the comparison with experimental results. However, this work has to be considered as an initial step to facilitate the development of a first tool to simulate irradiation effects. It can be improved by many ways (e.g. by use of dislocation dynamics code).

  12. A Comparison of Various Stress Rupture Life Models for Orbiter Composite Pressure Vessels and Confidence Intervals

    Science.gov (United States)

    Grimes-Ledesma, Lorie; Murthy, Pappu L. N.; Phoenix, S. Leigh; Glaser, Ronald

    2007-01-01

    In conjunction with a recent NASA Engineering and Safety Center (NESC) investigation of flight worthiness of Kevlar Overwrapped Composite Pressure Vessels (COPVs) on board the Orbiter, two stress rupture life prediction models were proposed independently by Phoenix and by Glaser. In this paper, the use of these models to determine the system reliability of 24 COPVs currently in service on board the Orbiter is discussed. The models are briefly described, compared to each other, and model parameters and parameter uncertainties are also reviewed to understand confidence in reliability estimation as well as the sensitivities of these parameters in influencing overall predicted reliability levels. Differences and similarities in the various models will be compared via stress rupture reliability curves (stress ratio vs. lifetime plots). Also outlined will be the differences in the underlying model premises, and predictive outcomes. Sources of error and sensitivities in the models will be examined and discussed based on sensitivity analysis and confidence interval determination. Confidence interval results and their implications will be discussed for the models by Phoenix and Glaser.

  13. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright; R. N. Wright

    2008-04-01

    The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have

  14. Bobbin-Tool Friction-Stir Welding of Thick-Walled Aluminum Alloy Pressure Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Dalder, E C; Pastrnak, J W; Engel, J; Forrest, R S; Kokko, E; Ternan, K M; Waldron, D

    2007-06-06

    It was desired to assemble thick-walled Al alloy 2219 pressure vessels by bobbin-tool friction-stir welding. To develop the welding-process, mechanical-property, and fitness-for-service information to support this effort, extensive friction-stir welding-parameter studies were conducted on 2.5 cm. and 3.8 cm. thick 2219 Al alloy plate. Starting conditions of the plate were the fully-heat-treated (-T62) and in the annealed (-O) conditions. The former condition was chosen with the intent of using the welds in either the 'as welded' condition or after a simple low-temperature aging treatment. Since preliminary stress-analyses showed that stresses in and near the welds would probably exceed the yield-strength of both 'as welded' and welded and aged weld-joints, a post-weld solution-treatment, quenching, and aging treatment was also examined. Once a suitable set of welding and post-weld heat-treatment parameters was established, the project divided into two parts. The first part concentrated on developing the necessary process information to be able to make defect-free friction-stir welds in 3.8 cm. thick Al alloy 2219 in the form of circumferential welds that would join two hemispherical forgings with a 102 cm. inside diameter. This necessitated going to a bobbin-tool welding-technique to simplify the tooling needed to react the large forces generated in friction-stir welding. The bobbin-tool technique was demonstrated on both flat-plates and plates that were bent to the curvature of the actual vessel. An additional issue was termination of the weld, i.e. closing out the hole left at the end of the weld by withdrawal of the friction-stir welding tool. This was accomplished by friction-plug welding a slightly-oversized Al alloy 2219 plug into the termination-hole, followed by machining the plug flush with both the inside and outside surfaces of the vessel. The second part of the project involved demonstrating that the welds were fit for the intended

  15. D0 Silicon Upgrade: Gas Helium Storage Tank Pressure Vessel Engineering Note

    Energy Technology Data Exchange (ETDEWEB)

    Rucinski, Russ; /Fermilab

    1996-11-11

    This is to certify that Beaird Industries, Inc. has done a white metal blast per SSPC-SP5 as required per specifications on the vessel internal. Following the blast, a black light inspection was performed by Beaird Quality Control personnel to assure that all debris, grease, etc. was removed and interior was clean prior to closing vessel for helium test.

  16. Validation of SCALE (SAS2H) Isotopic Predictions for BWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.

    1998-01-01

    Thirty spent fuel samples obtained from boiling-water-reactor (BWR) fuel pins have been modeled at Oak Ridge National Laboratory using the SAS2H sequence of the SCALE code system. The SAS2H sequence uses transport methods combined with the depletion and decay capabilities of the ORIGEN-S code to estimate the isotopic composition of fuel as a function of its burnup history. Results of these calculations are compared with chemical assay measurements of spent fuel inventories for each sample. Results show reasonable agreement between measured and predicted isotopic concentrations for important actinides; however, little data are available for most fission products considered to be important for spent fuel concerns (e.g., burnup credit, shielding, source-term calculations, etc.). This work is a follow-up to earlier works that studied the ability to predict spent fuel compositions in pressurized-water-reactor (PWR) fuel pins. Biases and uncertainties associated with BWR isotopic predictions are found to be larger than those of PWR calculations. Such behavior is expected, as the operation of a BWR is significantly more complex than that of a PWR plant, and in general the design of a BWR has a more heterogeneous configuration than that of a PWR. Nevertheless, this work shows that the simple models employed using SAS2H to represent such complexities result in agreement to within 5% (and often less than 1%) or less for most nuclides important for spent fuel applications. On the other hand, however, the set of fuel samples analyzed represent a small subset of the BWR fuel population, and results reported herein may not be representative of the full population of BWR spent fuel.

  17. Validation of SCALE (SAS2H) isotopic predictions for BWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.; DeHart, M.D.

    1998-09-01

    Thirty spent fuel samples obtained from boiling-water-reactor (BWR) fuel pins have been modeled at Oak Ridge National Laboratory using the SAS2H sequence of the SCALE code system. The SAS2H sequence uses transport methods combined with the depletion and decay capabilities of the ORIGEN-S code to estimate the isotopic composition of fuel as a function of its burnup history. Results of these calculations are compared with chemical assay measurements of spent fuel inventories for each sample. Results show reasonable agreement between measured and predicted isotopic concentrations for important actinides; however, little data are available for most fission products considered to be important for spent fuel concerns (e.g., burnup credit, shielding, source-term calculations, etc.). This work is a follow-up to earlier works that studied the ability to predict spent fuel compositions in pressurized-water-reactor (PWR) fuel pins. Biases and uncertainties associated with BWR isotopic predictions are found to be larger than those of PWR calculations. Such behavior is expected, as the operation of a BWR is significantly more complex than that of a PWR plant, and in general the design of a BWR has a more heterogeneous configuration than that of a PWR. Nevertheless, this work shows that the simple models employed using SAS2H to represent such complexities result in agreement to within 5% (and often less than 1%) or less for most nuclides important for spent fuel applications. On the other hand, however, the set of fuel samples analyzed represent a small subset of the BWR fuel population, and results reported herein may not be representative of the full population of BWR spent fuel.

  18. Safety-related considerations for reactor pressure vessels in consideration of hydrogen flaking

    Energy Technology Data Exchange (ETDEWEB)

    Dugan, S.; Herter, K.H.; Schuler, X.; Silcher, H. [Stuttgart Univ. (Germany). MPA

    2013-07-01

    During non-destructive inspection of the reactor pressure vessels in the Belgian nuclear power plants Doel 3 and Tihange 2, a large number of crack-like indications located in the base metal of the core shells were found. As part of the evaluation of these indications, which were identified as flake-like separations (hydrogen flakes), questions arise as to their cause, possible operational growth and the impact on the continued safe operation of the plant. In addition to the operational load cases, possible accidental and beyond design load cases are also of importance. Within the scope of the ''Research Project Component Safety'' (Forschungsvorhaben Komponentensicherheit - FKS) in the time frame mid-1970s to mid-1990s, numerous R and D activities on the material mechanics behavior and qualification of RPV materials were performed at MPA University of Stuttgart. The objectives of these investigations were focused on material mechanical issues related to the integrity of components and included standard material testing as well as component-like large scale specimen tests. Another major objective was the evaluation of non-destructive testing (NDT) methods with respect to their detection capabilities for such defects which developed during the manufacturing process. The investigations also included a study of the conditions favorable for formation or prevention of hydrogen flaking. In the context of this paper, the results from these R and D activities are presented in view of the current issues and in relation to the integrity concept for German RPVs. Ultrasonic testing (UT) techniques applied during manufacturing and during in-service inspections of German RPVs will also be discussed.

  19. Closed vessel combustion modelling by using pressure-time evolution function derived from two-zonal approach

    Directory of Open Access Journals (Sweden)

    Tomić Mladen A.

    2012-01-01

    Full Text Available In this paper a new method for burned mass fraction - pressure relation, x-p relation, for two-zone model combustion calculation is developed. The main application of the two-zone model is obtaining laminar burning velocity, SL, by using a pressure history from a closed vessel combustion experiment. The linear x-p relation by Lewis and Von Elbe is still widely used. For linear x-p relation, the end pressure is necessary as input data for the description of the combustion process. In this paper a new x-p relation is presented on the basis of mass and energy conservation during the combustion. In order to correctly represent pressure evolution, the model proposed in this paper needs several input parameters. They were obtained from different sources, like the PREMIX software (with GRIMECH 3.0 mechanism and GASEQ software, as well as thermodynamic tables. The error analysis is presented in regard to the input parameters. The proposed model is validated against the experiment by Dahoe and Goey, and compared with linear x-p relation from Lewis and Von Elbe. The proposed two zone model shows sufficient accuracy when describing the combustion process in a closed vessel without knowing the end pressure in advance, i.e. both peak pressure and combustion rates can be sufficiently correctly captured.

  20. Low background stainless steel for the pressure vessel in the PandaX-II dark matter experiment

    Science.gov (United States)

    Zhang, T.; Fu, C.; Ji, X.; Liu, J.; Liu, X.; Wang, X.; Yao, C.; Yuan, Xunhua

    2016-09-01

    We report on the custom produced low radiation background stainless steel and the welding rod for the PandaX experiment, one of the deep underground experiments to search for dark matter and neutrinoless double beta decay using xenon. The anthropogenic 60Co concentration in these samples is at the range of 1 mBq/kg or lower. We also discuss the radioactivity of nuclear-grade stainless steel from TISCO which has a similar background rate. The PandaX-II pressure vessel was thus fabricated using the stainless steel from CISRI and TISCO. Based on the analysis of the radioactivity data, we also made discussions on potential candidate for low background metal materials for future pressure vessel development.

  1. Views of TAGSI on the effects of gamma irradiation on the mechanical properties of irradiated ferritic steel reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Knott, J.F. [School of Engineering, Metallurgy and Materials, University of Birmingham, Edgbaston, Birmingham B15 2TT (United Kingdom); English, C.A. [Materials and Chemistry Consultancy, Nexia Solutions, 168 Harwell International Business Centre, Didcot, Oxon OX11 0QJ (United Kingdom); Weaver, D.R. [School of Physics, University of Birmingham, Edgbaston, Birmingham B15 2TT (United Kingdom); Lidbury, D.P.G. [Serco Assurance, Walton House, 404 The Quadrant, Birchwood Park, Warrington, Cheshire WA3 6AT (United Kingdom)

    2005-12-01

    The paper reviews and analyses the effects of gamma irradiation dose on the properties of ferritic steels used in reactor pressure vessels (RPVs). It explains factors that affect the embrittlement of a RPV steel induced by combinations of fast neutrons, thermal neutrons, and gamma irradiation. TAGSI were asked to consider the effects of gamma irradiation dose on the properties of steels used in reactor pressure vessels. TAGSI endorsed the use of the MCBEND code to calculate gamma fluxes and energetic gamma ray displacement cross-sections calculated using either Baumann or Alexander methods. TAGSI endorsed the calculation of the materials property changes due to an additional gamma dose using trend curves based on empirical correlation to neutron-induced damage (where k {sub {gamma}}{approx}1{+-}0.25)

  2. Low Background Stainless Steel for the Pressure Vessel in the PandaX-II Dark Matter Experiment

    CERN Document Server

    Zhang, Tao; Ji, Xiangdong; Liu, Jianglai; Liu, Xiang; Wang, Xuming; Yao, Chunfa; Yuan, Xunhua

    2016-01-01

    We report on the custom produced low radiation background stainless steel and the welding rod for the PandaX experiment, one of the deep underground experiments to search for dark matter and neutrinoless double beta decay using xenon. The anthropogenic 60 Co concentration in these samples is at the range of 1 mBq/kg or lower. We also discuss the radioactivity of nuclear-grade stainless steel from TISCO which has a similar background rate. The PandaX-II pressure vessel was thus fabricated using the stainless steel from CISRI and TISCO. Based on the analysis of the radioactivity data, we also made discussions on potential candidate for low background metal materials for future pressure vessel development.

  3. PSI contribution to the CASTOC round robin on EAC of low-alloy RPV steels under BWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2001-08-01

    Within the CASTOC-project (5th EU FW programme), the environmentally-assisted crack growth (EAC) behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state BWR power operation conditions by 6 European laboratories. The present report contains a summary of the PSI contribution to the Working Package 1 (WP1) of this project. WP1 is an interlaboratory round robin EAC test in simulated BWR/NWC environment under cyclic and static loading conditions. The round robin shall demonstrate the applicability of the used advanced test technique and establishes the technical basis for the decision of test conditions in the other working packages. In the first part of the report, the PSI testing facility/measurement instruments and the applied test and evaluation procedure are discussed in detail. In the second part, the exact test conditions and test results with detailed post-test fractographical evaluation in the SEM are presented. The test results are compared with other PSI results, literature data and nuclear codes. Stable and stationary test conditions within the specified range could be achieved in the PSI test during the whole conditioning and experimental phase. The cyclic crack growth rate results agree well with recent PSI results at a higher dissolved oxygen content of 8 ppm and are slightly below the 'high-sulphur line' of the PLEDGE-model. The crack growth rates are significantly above the ASME XI 'wet' curve. Compared to fatigue crack growth rates in air under otherwise identical test conditions, the effect of the high-temperature water environment resulted in an acceleration of crack growth by a factor of 150-250 under these low-cyclic loading conditions. The test results at constant load confirm the extremely low susceptibility to SCC crack growth under static load at 288 {sup o}C observed in tests at MPA, PSI and in a European Round Robin. They agree well with the RPV

  4. Mechanical properties of the as-forged and the forged-and-milled steels for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Bong Sang; Yoon, Ji Hyun; Kim, Joo Hak; Oh, Yong Jun; Hong, Jun Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-04-01

    The mechanical properties of the as-forged and the forged and milled SA508-Gr.3 reactor pressure vessel steels were evaluated. The full Charpy impact curves obtained for four different locations in test materials. The various data including yield strengths, tensile strengths, elongations were obtained from the tensile strengths, elongations were obtained from the tensile test results for two locations in test materials. The detailed test results were integrated and analysed in this report. 6 refs., 7 figs., 5 tabs. (Author)

  5. Exact and Numerical Elastic Analysis for the FGM Thick-Walled Cylindrical Pressure Vessels with Exponentially-Varying Properties

    Directory of Open Access Journals (Sweden)

    Nejad M. Zamani

    2016-09-01

    Full Text Available Assuming exponential-varying properties in the radial direction and based on the elasticity theory, an exact closed-form analytical solution is obtained to elastic analysis of FGM thick-walled cylindrical pressure vessels in the plane strain condition. Following this, radial distribution of radial displacement, radial stress, and circumferential stress are plotted for different values of material inhomogeneity constant. The displacements and stresses distributions are compared with the solutions of the finite element method (FEM.

  6. Preliminary investigation of an ultrasound method for estimating pressure changes in deep-positioned vessels

    DEFF Research Database (Denmark)

    Olesen, Jacob Bjerring; Villagómez Hoyos, Carlos Armando; Traberg, Marie Sand;

    2016-01-01

    This paper presents a method for measuring pressure changes in deep-tissue vessels using vector velocity ultrasound data. The large penetration depth is ensured by acquiring data using a low frequency phased array transducer. Vascular pressure changes are then calculated from 2-D angle-independen......This paper presents a method for measuring pressure changes in deep-tissue vessels using vector velocity ultrasound data. The large penetration depth is ensured by acquiring data using a low frequency phased array transducer. Vascular pressure changes are then calculated from 2-D angle......-independent vector velocity fields using a model based on the Navier-Stokes equations. Experimental scans are performed on a fabricated flow phantom having a constriction of 36% at a depth of 100 mm. Scans are carried out using a phased array transducer connected to the experimental scanner, SARUS. 2-D fields...... of angle-independent vector velocities are acquired using directional synthetic aperture vector flow imaging. The obtained results are evaluated by comparison to a 3-D numerical simulation model with equivalent geometry as the designed phantom. The study showed pressure drops across the constricted phantom...

  7. Blood pressure regulation V: in vivo mechanical properties of precapillary vessels as affected by long-term pressure loading and unloading.

    Science.gov (United States)

    Eiken, Ola; Mekjavic, Igor B; Kölegård, Roger

    2014-03-01

    Recent studies are reviewed, concerning the in vivo wall stiffness of arteries and arterioles in healthy humans, and how these properties adapt to iterative increments or sustained reductions in local intravascular pressure. A novel technique was used, by which arterial and arteriolar stiffness was determined as changes in arterial diameter and flow, respectively, during graded increments in distending pressure in the blood vessels of an arm or a leg. Pressure-induced increases in diameter and flow were smaller in the lower leg than in the arm, indicating greater stiffness in the arteries/arterioles of the leg. A 5-week period of intermittent intravascular pressure elevations in one arm reduced pressure distension and pressure-induced flow in the brachial artery by about 50%. Conversely, prolonged reduction of arterial/arteriolar pressure in the lower body by 5 weeks of sustained horizontal bedrest, induced threefold increases of the pressure-distension and pressure-flow responses in a tibial artery. Thus, the wall stiffness of arteries and arterioles are plastic properties that readily adapt to changes in the prevailing local intravascular pressure. The discussion concerns mechanisms underlying changes in local arterial/arteriolar stiffness as well as whether stiffness is altered by changes in myogenic tone and/or wall structure. As regards implications, regulation of local arterial/arteriolar stiffness may facilitate control of arterial pressure in erect posture and conditions of exaggerated intravascular pressure gradients. That increased intravascular pressure leads to increased arteriolar wall stiffness also supports the notion that local pressure loading may constitute a prime mover in the development of vascular changes in hypertension.

  8. Investigation of low-cycle fatigue behavior of austenitic stainless steel for cold-stretched pressure vessels

    Institute of Scientific and Technical Information of China (English)

    Cun-jian MIAO; Jin-yang ZHENG; Xiao-zhe GAO; Ze HUANG; A-bin GUO; Du-yi YE; Li MA

    2013-01-01

    Cold-stretched pressure vessels from austenitic stainless steels (ASS) are widely used for storage and transportation of liquefied gases,and have such advantages as thin wall and light weight.Fatigue is an important concern in these pressure vessels,which are subjected to alternative loads.Even though several codes and standards have guidelines on these pressure vessels,there are no relevant design methods on fatigue failure.To understand the fatigue properties of ASS 1.4301 (equivalents include UNS S30400 and AISI 304) in solution-annealed (SA) and cold-stretched conditions (9% strain level) and the response of fatigue properties to cold stretching (CS),low-cycle fatigue (LCF) tests were performed at room temperature,with total strain amplitudes ranging from ±0.4% to ±0.8%.Martensite transformations were measured during the tests.Comparisons on cyclic stress response,cyclic stress-strain behavior,and fatigue life were carried out between SA and CS materials.Results show that CS reduces the initial hardening stage,but prolongs the softening period in the cyclic stress response.Martensite transformation helps form a stable regime and subsequent secondary hardening.The stresses of monotonic and cyclic stress-strain curves are improved by CS,which leads to a lower plastic strain and a much higher elastic strain.The fatigue resistance of the CS material is better than that of the SA material,which is approximately 1 x l03 to 2×104 cycles.The S-N curve of the ASME standard for ASS is compared with the fatigue data and is justified to be suitable for the fatigue design of cold-stretched pressure vessels.However,considering the CS material has a better fatigue resistance,the S-N curve will be more conservative.The present study would be helpful in making full use of the advantages of CS to develop a new S-N curve for fatigue design of cold-stretched pressure vessels.

  9. Propagation of cracks by stress corrosion in conditions of BWR type reactor; Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua en ebullicion (BWR)

    Energy Technology Data Exchange (ETDEWEB)

    Merino C, F.J. [ININ, 52045 Estado de Mexico (Mexico); Fuentes C, P. [ITT, Metepec, Estado de Mexico (Mexico)]. E-mail: fjmc@nuclear.inin.mx

    2004-07-01

    In this work, the obtained results when applying the Hydrogen Chemistry to a test tube type Compact Tension (CT), built in austenitic stainless steel 304l, simulating the conditions to those that it operates a Boiling Water Reactor (BWR), temperature 288 C and pressure of 8 MPa are presented. With the application of this water chemistry, seeks to be proven the diminution of the crack propagation speed. (Author)

  10. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  11. 46 CFR 35.25-5 - Repairs of boilers and unfired pressure vessels and reports of repairs or accidents by chief...

    Science.gov (United States)

    2010-10-01

    ... reports of repairs or accidents by chief engineer-TB/ALL. 35.25-5 Section 35.25-5 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY TANK VESSELS OPERATIONS Engine Department § 35.25-5 Repairs of boilers and unfired pressure vessels and reports of repairs or accidents by chief engineer—TB/ALL. (a) Before...

  12. Fatigue test of carbon epoxy composite high pressure hydrogen storage vessel under hydrogen environment

    Institute of Scientific and Technical Information of China (English)

    Chuan-xiang ZHENG; Liang WANG; Rong LI; Zong-xin WEI; Wei-wei ZHOU

    2013-01-01

    A significant temperature raise within hydrogen vehicle cylinder during the fast filling process will be observed,while the strength and fatigue life of the cylinder will dramatically decrease at high temperature.In order to evaluate the strength and fatigue of composite hydrogen storage vessel,a 70-MPa fatigue test system using hydrogen medium was set up.Experimental study on the fatigue of composite hydrogen storage vessels under real hydrogen environment was performed.The experimental results show that the ultimate strength and fatigue life both decreased obviously compared with the values under hydraulic fatigue test.Furthermore,fatigue property,failure behavior,and safe hydrogen charging/discharging working mode of onboard hydrogen storage vessels were obtained through the fatigue tests.

  13. Signal analysis of acoustic and flow-induced vibrations of BWR main steam line

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Paredes, G., E-mail: gepe@xanum.uam.mx [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Prieto-Guerrero, A. [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Núñez-Carrera, A. [Comisión Nacional de Seguridad Nuclear y Salvaguardias, Doctor Barragán 779, Col. Narvarte, México, D.F. 03020 (Mexico); Vázquez-Rodríguez, A. [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Centeno-Pérez, J. [Instituto Politécnico Nacional, Escuela Superior de Física y Matemáticas Unidad Profesional “Adolfo López Mateos”, Av. IPN, s/n, México, D.F. 07738 (Mexico); Espinosa-Martínez, E.-G. [Departamento de Sistemas Energéticos, Universidad Nacional Autónoma de México, México, D.F. 04510 (Mexico); and others

    2016-05-15

    Highlights: • Acoustic and flow-induced vibrations of BWR are analyzed. • BWR performance after extended power uprate is considered. • Effect of acoustic side branches (ASB) is analyzed. • The ASB represents a reduction in the acoustic loads to the steam dryer. • Methodology developed for simultaneous analyzing the signals in the MSL. - Abstract: The aim of this work is the signal analysis of acoustic waves due to phenomenon known as singing in Safety Relief Valves (SRV) of the main steam lines (MSL) in a typical BWR5. The acoustic resonance in SRV standpipes and fluctuating pressure is propagated from SRV to the dryer through the MSL. The signals are analyzed with a novel method based on the Multivariate Empirical Mode Decomposition (M-EMD). The M-EMD algorithm has the potential to find common oscillatory modes (IMF) within multivariate data. Based on this fact, we implement the M-EMD technique to find the oscillatory mode in BWR considering the measurements obtained collected by the strain gauges located around the MSL. These IMF, analyzed simultaneously in time, allow obtaining an estimation of the effects of the multiple-SRV in the MSL. Two scenarios are analyzed: the first is the signal obtained before the installation of the acoustic dampers (ASB), and the second, the signal obtained after installation. The results show the effectiveness of the ASB to damp the strong resonances when the steam flow increases, which represents an important reduction in the acoustic loads to the steam dryer.

  14. Development and Testing of CTF to Support Modeling of BWR Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Salko, Robert K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Collins, Benjamin S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Godfrey, Andrew T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gosdin, Chris [Pennsylvania State Univ., University Park, PA (United States); Avramova, Maria [North Carolina State Univ., Raleigh, NC (United States)

    2016-01-29

    This milestone supports developing and assessing COBRA-TF (CTF) for the modeling of boiling water reactors (BWRs). This is achieved in three stages. First, a new preprocessor utility that is capable of handling BWR-specic design elements (e.g., channel boxes and large water rods) is developed. A previous milestone (L3:PHI.CTF.P12.01) led to the development of this preprocessor capability for single assembly models. This current milestone expands this utility so that it is applicable to multi-assembly BWR models that can be modeled in either serial or parallel. The second stage involves making necessary modications to CTF so that it can execute these new models. Specically, this means implementing an outer-iteration loop, specic to BWR models, that equalizes the pressure loss over all assemblies in the core (which are not connected due to the channel boxes) by adjusting inlet mass ow rate. A third stage involves assessing the standard convergence metrics that are used by CTF to determine when a simulation is steady-state. The nal stage has resulted in the implementation of new metrics in the code that give a better indication of how steady the solution is at convergence. This report summarizes these eorts and provides a demonstration of CTF's BWR-modeling capabilities. CASL-U-2016-1030-000

  15. Development of automated welding process for field fabrication of thick walled pressure vessels. Fourth quarter technical progress report for period ending September 30, 1979

    Energy Technology Data Exchange (ETDEWEB)

    None

    1979-01-01

    Progress in developing an automated welding process for the field fabrication of thick walled pressure vessels is reported. Plans for the demonstration facility, for nondestructive testing, and for the procurement of materials are discussed. (LCL)

  16. In vivo quantification of lymph viscosity and pressure in lymphatic vessels and draining lymph nodes of arthritic joints in mice.

    Science.gov (United States)

    Bouta, Echoe M; Wood, Ronald W; Brown, Edward B; Rahimi, Homaira; Ritchlin, Christopher T; Schwarz, Edward M

    2014-03-15

    Rheumatoid arthritis (RA) is a chronic inflammatory joint disease with episodic flares. In TNF-Tg mice, a model of inflammatory-erosive arthritis, the popliteal lymph node (PLN) enlarges during the pre-arthritic 'expanding' phase, and then 'collapses' with adjacent knee flare associated with the loss of the intrinsic lymphatic pulse. As the mechanisms responsible are unknown, we developed in vivo methods to quantify lymph viscosity and pressure in mice with wild-type (WT), expanding and collapsed PLN. While no differences in viscosity were detected via multiphoton fluorescence recovery after photobleaching (MP-FRAP) of injected FITC-BSA, a 32.6% decrease in lymph speed was observed in vessels afferent to collapsed PLN (P pressure (LNP) demonstrated a decrease in expanding PLN versus WT pressure (3.41 ± 0.43 vs. 6.86 ± 0.56 cmH2O; P pressure (LPP), measured indirectly by slowly releasing a pressurized cuff occluding indocyanine green (ICG), demonstrated an increase in vessels afferent to expanding PLN versus WT (18.76 ± 2.34 vs. 11.04 ± 1.47 cmH2O; P pressure, and provide evidence to support the hypothesis that lymphangiogenesis and lymphatic transport are compensatory mechanisms to prevent synovitis via increased drainage of inflamed joints. Furthermore, the decrease in lymphatic flow and loss of LPP during PLN collapse are consistent with decreased drainage from the joint during arthritic flare, and validate these biomarkers of RA progression and possibly other chronic inflammatory conditions.

  17. 30 CFR 57.13001 - General requirements for boilers and pressure vessels.

    Science.gov (United States)

    2010-07-01

    ... vessels. 57.13001 Section 57.13001 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL MINE SAFETY AND HEALTH SAFETY AND HEALTH STANDARDS-UNDERGROUND METAL AND... the standards and specifications of the American Society of Mechanical Engineers Boiler and...

  18. Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

    1989-01-01

    Recent data from the HFIR vessel surveillance program indicate a substantial radiation embrittlement rate effect at low irradiation temperatures (/approximately/120/degree/F) for A212-B, A350-LF3, A105-II, and corresponding welds. PWR vessel supports are fabricated of similar materials and are subjected to the same low temperatures and fast neutron fluxes (10/sup 8/ to 10/sup 9/ neutrons/cm/sup 2//center dot/s, E > 1.0 MeV) as those in the HFIR vessel. Thus, the embrittlement rate of these structures may be greater than previously anticipated. A study sponsored by the NRC is under way at ORNL to determine the impact of the rate effect on PWR vessel-support life expectancy. The scope includes the interpretation and application of the HFIR data, a survey of all light-water-reactor vessel support designs, and a structural and fracture-mechanics analysis of the supports for two specific PWR plants of particular interest with regard to a potential for support failure as a result of propagation of flaws. Calculations performed thus far indicate best-estimate critical flaw sizes, corresponding to 32 EFPY, of /approximately/0.2 in. for one plant and /approximately/0.4 in. for the other. These flaw sizes are small enough to be of concern. However, it appears that low-cycle fatigue is not a viable mechanism for creation of flaws of this size, and thus, presumably, such flaws would have to exist at the time of fabrication. 59 refs., 128 figs., 49 tabs.

  19. Process inherent ultimate safety/boiling-water reactor PIUS/BWR

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.

    1985-01-01

    This document is a series of viewgraphs on: design basis of PIUS/BWR, definition of PIUS/BWR, mechanisms of safe shutdown and afterheat cooling, advantages of PIUS/BWR, and research and development requirements. (DLC)

  20. Probabilistic Fracture Mechanics Analysis of Boling Water Reactor Vessel for Cool-Down and Low Temperature Over-Pressurization Transients

    Directory of Open Access Journals (Sweden)

    Jeong Soon Park

    2016-04-01

    Full Text Available The failure probabilities of the reactor pressure vessel (RPV for low temperature over-pressurization (LTOP and cool-down transients are calculated in this study. For the cool-down transient, a pressure–temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition (RTNDT. The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.

  1. Development of a Weibull model of cleavage fracture toughness for shallow flaws in reactor pressure vessel material

    Energy Technology Data Exchange (ETDEWEB)

    Bass, B.R.; Williams, P.T.; McAfee, W.J.; Pugh, C.E. [Oak Ridge National Lab., Heavy-Section Steel Technology Program, Oak Ridge, TN (United States)

    2001-07-01

    A primary objective of the United States Nuclear Regulatory Commission (USNRC) -sponsored Heavy-Section Steel Technology (HSST) Program is to develop and validate technology applicable to quantitative assessments of fracture prevention margins in nuclear reactor pressure vessels (RPVs) containing flaws and subjected to service-induced material toughness degradation. This paper describes an experimental/analytical program for the development of a Weibull statistical model of cleavage fracture toughness for applications to shallow surface-breaking and embedded flaws in RPV materials subjected to multi-axial loading conditions. The experimental part includes both material characterization testing and larger fracture toughness experiments conducted using a special-purpose cruciform beam specimen developed by Oak Ridge National Laboratory for applying biaxial loads to shallow cracks. Test materials (pressure vessel steels) included plate product forms (conforming to ASTM A533 Grade B Class 1 specifications) and shell segments procured from a pressurized-water reactor vessel intended for a nuclear power plant. Results from tests performed on cruciform specimens demonstrated that biaxial loading can have a pronounced effect on shallow-flaw fracture toughness in the lower-transition temperature region. A local approach methodology based on a three-parameter Weibull model was developed to correlate these experimentally-observed biaxial effects on fracture toughness. The Weibull model, combined with a new hydrostatic stress criterion in place of the more commonly used maximum principal stress in the kernel of the Weibull stress integral definition, is shown to provide a scaling mechanism between uniaxial and biaxial loading states for 2-dimensional flaws located in the A533-B plate material. The Weibull stress density was introduced as a matrice for identifying regions along a semi-elliptical flaw front that have a higher probability of cleavage initiation. Cumulative

  2. Prediction and Monitoring Systems of Creep-Fracture Behavior of 9Cr-1Mo Steels for Teactor Pressure Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Potirniche, Gabriel [Univ. of Idaho, Moscow, ID (United States); Barlow, Fred D. [Univ. of Idaho, Moscow, ID (United States); Charit, Indrajit [Univ. of Idaho, Moscow, ID (United States); Rink, Karl [Univ. of Idaho, Moscow, ID (United States)

    2013-11-26

    A recent workshop on next-generation nuclear plant (NGNP) topics underscored the need for research studies on the creep fracture behavior of two materials under consideration for reactor pressure vessel (RPV) applications: 9Cr-1Mo and SA-5XX steels. This research project will provide a fundamental understanding of creep fracture behavior of modified 9Cr-1Mo steel welds for through modeling and experimentation and will recommend a design for an RPV structural health monitoring system. Following are the specific objectives of this research project: Characterize metallurgical degradation in welded modified 9Cr-1Mo steel resulting from aging processes and creep service conditions; Perform creep tests and characterize the mechanisms of creep fracture process; Quantify how the microstructure degradation controls the creep strength of welded steel specimens; Perform finite element (FE) simulations using polycrystal plasticity to understand how grain texture affects the creep fracture properties of welds; Develop a microstructure-based creep fracture model to estimate RPVs service life; Manufacture small, prototypic, cylindrical pressure vessels, subject them to degradation by aging, and measure their leak rates; Simulate damage evolution in creep specimens by FE analyses; Develop a model that correlates gas leak rates from welded pressure vessels with the amount of microstructural damage; Perform large-scale FE simulations with a realistic microstructure to evaluate RPV performance at elevated temperatures and creep strength; Develop a fracture model for the structural integrity of RPVs subjected to creep loads; and Develop a plan for a non-destructive structural health monitoring technique and damage detection device for RPVs.

  3. Study on surface nanocrystallization and resisting H2S stresscorrosion properties of pressure vessel steel welding joints

    Institute of Scientific and Technical Information of China (English)

    2005-01-01

    Many efforts were spent on the homogenization of microstructure and property of welding joints. A new surface nanocrystallization technique named Supersonic Particles Bombarding(SSPB) can be used for this purpose. Two kinds of pressure vessel steel welding joints, 16MnR and 0Cr18Ni9Ti, were chosen to be treated by SSPB. Transmission electron microscopy was introduced to examine the surface microstructure. And their ability to resist H2 S stress corrosion was enhanced significantly after the SSPB treatment. The mechanism for the results were analyzed as well.

  4. Comparative assessment of cyclic J-R curve determination by different methods in a pressure vessel steel

    Science.gov (United States)

    Chowdhury, Tamshuk; Sivaprasad, S.; Bar, H. N.; Tarafder, S.; Bandyopadhyay, N. R.

    2016-04-01

    Cyclic J-R behaviour of a reactor pressure vessel steel using different methods available in literature has been examined to identify the best suitable method for cyclic fracture problems. Crack opening point was determined by moving average method. The η factor was experimentally determined for cyclic loading conditions and found to be similar to that of ASTM value. Analyses showed that adopting a procedure analogous to the ASTM standard for monotonic fracture is reasonable for cyclic fracture problems, and makes the comparison to monotonic fracture results straightforward.

  5. Nitrogen-Bearing Stainless Steels for Pressure Vessel%压力容器用含氮不锈钢

    Institute of Scientific and Technical Information of China (English)

    黄嘉琥

    2013-01-01

    压力容器用不锈钢几乎全为奥氏体不锈钢和奥氏体/铁素体双相不锈钢。在ASME -2011a, EN 13445:2009及GB 150-2011压力容器标准中所采用的奥氏体与双相不锈钢的所有牌号中,含氮钢牌号所占百分比分别为63.4%,76.4%和60%。在不锈钢材料标准中所有的双相不锈钢(指1971年后研制的牌号)、超级奥氏体不锈钢以及超级或特超级双相不锈钢中的氮含量基本均为0.1%~0.6%(称其为中氮型不锈钢)。EN 13445:2009规定可用于-273℃的奥氏体不锈钢10个牌号均为含氮钢。本文讨论了氮在不锈钢中的溶解度与含量、含氮不锈钢的类型、氮对不锈钢性能与组织的影响以及含氮不锈钢在压力容器中的应用。国外压力容器标准中N=0.1%~0.6%的中氮型奥氏体不锈钢已成为最重要的高性能不锈钢。这些钢在中国压力容器标准中尚未采用。而ASME-2011a中已用12个牌号,EN 13445:2009中已用14个牌号。中国仅在GB/T 20878-2007中已有23个牌号,GB/T 4237-2007(或GB/T 3280-2007)中已有13个牌号。本文建议在此基础上可进行更多的压力容器的应用工作。%Types of stainless steel for pressure vessel almost are all austenitic stainless steel and duplex austenitic/ferritic stainless steel.The percentages of designations of nitrogen-bearing steel grade individ-ually are 63 .4%,76 .4% and 60% among all designations of austenitic and duplex stainless steel types used in ASME-2011a,EN 13445:2009 and GB 150-2011 pressure vessel standards.Nitrogen contents of all duplex stainless steels(research after 1971),all super austenitic stainless steels and all super or hy-per duplex stainless steels in stainless steel standards basically are 0.1%~0.6%(call them middle ni-trogen-bearing stainless steel).10 designations of austenitic stainless steel can used for -273 ℃speci-fied in EN 13445:2009 are all nitrogen-bearing steel grade

  6. Comparison of microstructural features of radiation embrittlement of VVER-440 and VVER-1000 reactor pressure vessel steels

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Shtrombakh, Ya. I.; Erak, D. Yu.; Lavrenchuk, O. V.

    2002-02-01

    Comparative microstructural studies of both surveillance specimens and reactor pressure vessel (RPV) materials of VVER-440 and VVER-1000 light water reactor systems have been carried out, following irradiation to different fast neutron fluences and of the heat treatment for extended periods at the operating temperatures. It is shown that there are several microstructural features in the radiation embrittlement of VVER-1000 steels compared to VVER-440 RPV steels that can cause changes in the contributions of different radiation embrittlement mechanisms for VVER-1000 steel.

  7. Development of Long Shank Repair Tool for Defect of Pressure Vessel Bolt Hole in Pressure Vessel of Nuclear Power Plant%核电站压力容器螺孔长杆梳刀装置研制

    Institute of Scientific and Technical Information of China (English)

    黄新东; 黄辉; 洪龙; 李鑫

    2013-01-01

    In the reactor operation and operations with open lid,various defects may emerge on thread section of the main bolt hole pressure vessel.These defects must be dealt with before the closing of the lid.In view of the above conditions,this paper developed a pressure vessel screw rod cutter device,and expounds the long shank repair tool,a detailed description of the design scheme of the tool as well as the concrete structure.%在反应堆运行和开盖操作过程中,压力容器的主螺栓孔的螺纹段可能会产生各种缺陷,这些缺陷在再次扣盖前必须经过处理.针对上述工况,研制了一种压力容器螺孔长杆梳刀装置,本文阐述了该长杆梳刀装置的设计要求,详细描述该装置的设计方案以及具体结构形式.

  8. Blood pressure and sodium: Association with MRI markers in cerebral small vessel disease

    OpenAIRE

    Heye, Anna K.; Thrippleton, Michael J; Chappell, Francesca M; Valdés Hernández, Maria del C.; Armitage, Paul A.; Makin, Stephen D.; Muñoz Maniega, Susana; Sakka, Eleni; Flatman, Peter W.; Dennis, Martin S.; Wardlaw, Joanna M.

    2016-01-01

    Dietary salt intake and hypertension are associated with increased risk of cardiovascular disease including stroke. We aimed to explore the influence of these factors, together with plasma sodium concentration, in cerebral small vessel disease (SVD). In all, 264 patients with nondisabling cortical or lacunar stroke were recruited. Patients were questioned about their salt intake and plasma sodium concentration was measured; brain tissue volume and white-matter hyperintensity (WMH) load were m...

  9. Validation of 3D Code KATRIN For Fast Neutron Fluence Calculation of VVER-1000 Reactor Pressure Vessel by Ex-Vessel Measurements and Surveillance Specimens Results

    Directory of Open Access Journals (Sweden)

    Dzhalandinov A.

    2016-01-01

    Full Text Available Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.

  10. Validation of 3D Code KATRIN For Fast Neutron Fluence Calculation of VVER-1000 Reactor Pressure Vessel by Ex-Vessel Measurements and Surveillance Specimens Results

    Science.gov (United States)

    Dzhalandinov, A.; Tsofin, V.; Kochkin, V.; Panferov, P.; Timofeev, A.; Reshetnikov, A.; Makhotin, D.; Erak, D.; Voloschenko, A.

    2016-02-01

    Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV) for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.

  11. Sialyte(TM)-Based Composite Pressure Vessels for Extreme Environments Project

    Data.gov (United States)

    National Aeronautics and Space Administration — While traveling to Venus, electronics and instruments go through enormous pressure, temperature, and atmospheric environment changes. In the past, this has caused...

  12. Use of Closed Vessel as a Constant Pressure Apparatus for the Measurement of the Rate of Burning of Propellants

    Directory of Open Access Journals (Sweden)

    D. Vittal

    1980-04-01

    Full Text Available A method for the determination of burning rates of propellants whose from function is unknown is introduced. The method consists of burning in the closed vessel, a known charge weight of the test propellant alongwith a known pressure which remains nearly constant during the burning of the test propellant whose web size is the only quantity required for the evaluation of its rate of burning. The test propellants burns at near constant pressure conditions just as in the strand burner technique. This method can be applied to any unknown propellant of any shape whose web size can be measured and very large webs also can be used. In addition, the measurement of the records and the computation are very simple.

  13. 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.

  14. Modeling the Ductile Brittle Fracture Transition in Reactor Pressure Vessel Steels using a Cohesive Zone Model based approach

    Energy Technology Data Exchange (ETDEWEB)

    Pritam Chakraborty; S. Bulent Biner

    2013-10-01

    Fracture properties of Reactor Pressure Vessel (RPV) steels show large variations with changes in temperature and irradiation levels. Brittle behavior is observed at lower temperatures and/or higher irradiation levels whereas ductile mode of failure is predominant at higher temperatures and/or lower irradiation levels. In addition to such temperature and radiation dependent fracture behavior, significant scatter in fracture toughness has also been observed. As a consequence of such variability in fracture behavior, accurate estimates of fracture properties of RPV steels are of utmost importance for safe and reliable operation of reactor pressure vessels. A cohesive zone based approach is being pursued in the present study where an attempt is made to obtain a unified law capturing both stable crack growth (ductile fracture) and unstable failure (cleavage fracture). The parameters of the constitutive model are dependent on both temperature and failure probability. The effect of irradiation has not been considered in the present study. The use of such a cohesive zone based approach would allow the modeling of explicit crack growth at both stable and unstable regimes of fracture. Also it would provide the possibility to incorporate more physical lower length scale models to predict DBT. Such a multi-scale approach would significantly improve the predictive capabilities of the model, which is still largely empirical.

  15. Disclosure of the oscillations in kinetics of the reactor pressure vessel steel damage at fast neutron intensity decreasing

    Science.gov (United States)

    Krasikov, E.; Nikolaenko, V.

    2017-01-01

    Fast neutron intensity influence on reactor materials radiation damage is a critically important question in the problem of the correct use of the accelerated irradiation tests data for substantiation of the materials workability in real irradiation conditions that is low neutron intensity. Investigations of the fast neutron intensity (flux) influence on radiation damage and experimental data scattering reveal the existence of non-monotonous sections in kinetics of the reactor pressure vessels (RPV) steel damage. Discovery of the oscillations as indicator of the self-organization processes presence give reasons for new ways searching on reactor pressure vessel (RPV) steel radiation stability increasing and attempt of the self-restoring metal elaboration. Revealing of the wavelike process in the form of non monotonous parts of the kinetics of radiation embrittlement testifies that periodic transformation of the structure take place. This fact actualizes the problem of more precise definition of the RPV materials radiation embrittlement mechanisms and gives reasons for search of the ways to manage the radiation stability (nanostructuring and so on to stimulate the radiation defects annihilation), development of the means for creating of more stableness self recovering smart materials.

  16. Oxide evolution on Alloy X-750 in simulated BWR environment

    Science.gov (United States)

    Tuzi, Silvia; Göransson, Kenneth; Rahman, Seikh M. H.; Eriksson, Sten G.; Liu, Fang; Thuvander, Mattias; Stiller, Krystyna

    2016-12-01

    In order to simulate the environment experienced by spacer grids in a boiling water reactor (BWR), specimens of the Ni-based Alloy X-750 were exposed to a water jet in an autoclave at a temperature of 286 °C and a pressure of 80 bar. The oxide microstructure of specimens exposed for 2 h, 24 h, 168 h and 840 h has been investigated mainly using electron microscopy. The specimens suffer mass loss due to dissolution during exposure. At the same time a complex layered oxide develops. After the longest exposure the oxide consists of two outer spinel layers consisting of blocky crystals, one intermediate layer of nickel oxide interspersed with Ti-rich oxide needles, and an inner layer of oxidized base metal. The evolution of the oxide leading up to this structure is discussed and a model is presented.

  17. Benchmark calculation for radioactivity inventory using MAXS library based on JENDL-4.0 and JEFF-3.0/A for decommissioning BWR plants

    Directory of Open Access Journals (Sweden)

    Tanaka Ken-ichi

    2016-01-01

    Full Text Available We performed benchmark calculation for radioactivity activated in a Primary Containment Vessel (PCV of a Boiling Water Reactor (BWR by using MAXS library, which was developed by collapsing with neutron energy spectra in the PCV of the BWR. Radioactivities due to neutron irradiation were measured by using activation foil detector of Gold (Au and Nickel (Ni at thirty locations in the PCV. We performed activation calculations of the foils with SCALE5.1/ORIGEN-S code with irradiation conditions of each foil location as the benchmark calculation. We compared calculations and measurements to estimate an effectiveness of MAXS library.

  18. Research on the water hammer protection of the long distance water supply project with the combined action of the air vessel and over-pressure relief valve

    Science.gov (United States)

    Li, D. D.; Jiang, J.; Zhao, Z.; Yi, W. S.; Lan, G.

    2013-12-01

    We take a concrete pumping station as an example in this paper. Through the calculation of water hammer protection with a specific pumping station water supply project, and the analysis of the principle, mathematical models and boundary conditions of air vessel and over-pressure relief valve we show that the air vessel can protect the water conveyance system and reduce the transient pressure damage due to various causes. Over-pressure relief valve can effectively reduce the water hammer because the water column re-bridge suddenly stops the pump and prevents pipeline burst. The paper indicates that the combination set of air vessel and over-pressure relief valve can greatly reduce the quantity of the air valve and can eliminate the water hammer phenomenon in the pipeline system due to the vaporization and water column separation and re-bridge. The conclusion could provide a reference for the water hammer protection of long-distance water supply system.

  19. Radiological consequence assessments of degraded core accident scenarios derived from a generic Level 2 PSA of a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Homma, Toshimitsu; Ishikawa, Jun; Tomita, Kenichi; Muramatsu, Ken [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-12-01

    The radiological consequence assessments have been made of postulated core damage accidents with source terms derived from a generic Level 2 PSA of a BWR carried out by the Japan Atomic Energy Research Institute (JAERI). The source terms used were for the five core damage accident sequences with the drywell and wetwell failure cases, the release control case by venting of the containment and the accident termination case by the containment spray. The radiological consequences have been assessed for individual dose, collective dose, individual risk of early health effects and individual risk of late health effects by a probabilistic accident consequence assessment code, OSCAAR developed in JAERI. Following conclusions were obtained for the assumed source terms. In case of the over pressure failures of the primary containment vessel, the early fatalities can be mitigated through the implementation of early countermeasures, and the late cancer fatalities remains small. For the release control and accident termination cases, the individual and collective doses to the public can be reduced without any countermeasures due to the release reduction of the volatile radionuclides such as iodine and cesium. (author)

  20. Monitoring of the production quality of fibre-reinforced pressure vessels using acoustic emission testing; Ueberwachung der Fertigungsqualitaet von Faserverbund-Druckbehaeltern mittels Schallemissionspruefung

    Energy Technology Data Exchange (ETDEWEB)

    Duffner, Eric; Gregor, Christian; Bohse, Juergen [BAM Bundesanstalt fuer Materialforschung und -pruefung, Berlin (Germany)

    2011-07-01

    The investigation aimed at the validation of a test method for ensuring the production quality of reinforced-fibre pressure vessels in real fabrication conditions. The method is based on characteristics and permissible limiting values derived from acoustic emission curves during the first pressure test. The method had already been tested successfully on reinforced-fibre pressure vessels with metal liners and had been patented. With the current investigations, the possibility of detection fabrication defects in carbon fibre / glass fibre hybrid pressure vessels with polymer liners was evaluated. For this, fibre-reinforced pressure vessels were monitored by acoustic emission measurement during the first hydraulic pressure test; this test is commonly used for quality assurance of this type of pressure vessel, although without acoustic emission testing. Acoustic emission curves were registered for pressure vessels of a serial production, and the mean characteristics and their scatter were determined as reference values. These were compared with the acoustic emission curves of selectively induced fabrication defects. Fabrication defects are defects that may occur in serial production and are difficult or impossible to detect by conventional quality assurance methods. All investigated pressure vessel were then subject to stress until failure (leakage, bursting). This made it possible to verify the real influence of fabrication defects on the burst pressure and/or the fatigue characteristics of the pressure vessels and to assess the validity of acoustic emission testing. [German] Ziel der Untersuchung ist die Validierung einer Pruefmethodik zur Sicherung der Fertigungsqualitaet von Faserverbund - Druckbehaeltern unter realen Fertigungsbedingungen. Das Verfahren basiert auf Merkmalen und zulaessigen Grenzwerten, die aus Schallemissionsverlaeufen bei der Erstdruckpruefung abgeleitet werden [1]. Die Methodik konnte zuvor bereits erfolgreich an Faserverbund - Druckbehaeltern

  1. Potential impact of enhanced fracture-toughness data on fracture mechanics assessment of PWR vessel integrity for pressurized thermal shock

    Energy Technology Data Exchange (ETDEWEB)

    Dickson, T.L.; Theiss, T.J.

    1991-01-01

    The Heavy Section Steel Technology (HSST) Program is involved with the generation of enhanced fracture-initiation toughness and fracture-arrest toughness data of prototypic nuclear reactor vessel steels. These two sets of data are enhanced because they have distinguishing characteristics that could potentially impact PWR pressure vessel integrity assessments for the pressurized-thermal shock (PTS) loading condition which is a major plant-life extension issue to be confronted in the 1990's. A series of large-scale fracture-mechanics experiments have produced crack-arrest (K{sub Ia}) data with the distinguishing characteristic that the values are considerably above 220 MPA {center dot} {radical}m. The implicit limit of the ASME Code and the limit used in the Integrated Pressurized Thermal Shock (IPTS) studies. Currently, the HSST Program is planning experiments to verify and quantify for A533B steel the distinguishing characteristic of elevated the distinguishing characteristic of elevated initiation-fracture toughness for shallow flaws which has been observed for other steels. The results of the analyses indicated that application of the enhanced K{sub Ia} data does reduce the conditional probability of failure P(F{vert bar}E); however, it does not appear to have the potential to significantly impact the results of PTS analyses. The application of enhanced fracture-initiation-toughness data for shallow flaws also reduces P(F{vert bar}E), and does appear to have a potential for significantly affecting the results of PTS analyses. 19 refs., 11 figs., 1 tab.

  2. BWR mechanics and materials technology update

    Energy Technology Data Exchange (ETDEWEB)

    Kiss, E.

    1983-05-01

    This paper discusses technical results obtained from a variety of important programs underway at General Electric's Nuclear Engineering Division. The principal objective of these programs is to qualify and improve BWR product related technologies that fall broadly under the disciplines of Applied Mechanics and Materials Engineering. The paper identifies and deals with current technical issues that are of general importance to the LWR industry albeit the specific focus is directed to the development and qualification of analytical predictive methods and criteria, and improved materials for use in the design of the BWR. In this paper, specific results and accomplishments are summarized to provide a braod perspective of technology advances. Results are presented in sections which discuss: dynamic analysis and modeling; fatigue and fracture evaluation; materials engineering advances; and flow induced vibration.

  3. The modeling of core melting and in-vessel corium relocation in the APRIL code

    Energy Technology Data Exchange (ETDEWEB)

    Kim. S.W.; Podowski, M.Z.; Lahey, R.T. [Rensselaer Polytechnic Institute, Troy, NY (United States)] [and others

    1995-09-01

    This paper is concerned with the modeling of severe accident phenomena in boiling water reactors (BWR). New models of core melting and in-vessel corium debris relocation are presented, developed for implementation in the APRIL computer code. The results of model testing and validations are given, including comparisons against available experimental data and parametric/sensitivity studies. Also, the application of these models, as parts of the APRIL code, is presented to simulate accident progression in a typical BWR reactor.

  4. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright; R. N. Wright

    2010-07-01

    The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

  5. A high-throughput platform for low-volume high-temperature/pressure sealed vessel solvent extractions

    Energy Technology Data Exchange (ETDEWEB)

    Damm, Markus [Christian Doppler Laboratory for Microwave Chemistry (CDLMC) and Institute of Chemistry, Karl-Franzens-University Graz, Heinrichstrasse 28, A-8010 Graz (Austria); Kappe, C. Oliver, E-mail: oliver.kappe@uni-graz.at [Christian Doppler Laboratory for Microwave Chemistry (CDLMC) and Institute of Chemistry, Karl-Franzens-University Graz, Heinrichstrasse 28, A-8010 Graz (Austria)

    2011-11-30

    Highlights: Black-Right-Pointing-Pointer Parallel low-volume coffee extractions in sealed-vessel HPLC/GC vials. Black-Right-Pointing-Pointer Extractions are performed at high temperatures and pressures (200 Degree-Sign C/20 bar). Black-Right-Pointing-Pointer Rapid caffeine determination from the liquid phase. Black-Right-Pointing-Pointer Headspace analysis of volatiles using solid-phase microextraction (SPME). - Abstract: A high-throughput platform for performing parallel solvent extractions in sealed HPLC/GC vials inside a microwave reactor is described. The system consist of a strongly microwave-absorbing silicon carbide plate with 20 cylindrical wells of appropriate dimensions to be fitted with standard HPLC/GC autosampler vials serving as extraction vessels. Due to the possibility of heating up to four heating platforms simultaneously (80 vials), efficient parallel analytical-scale solvent extractions can be performed using volumes of 0.5-1.5 mL at a maximum temperature/pressure limit of 200 Degree-Sign C/20 bar. Since the extraction and subsequent analysis by either gas chromatography or liquid chromatography coupled with mass detection (GC-MS or LC-MS) is performed directly from the autosampler vial, errors caused by sample transfer can be minimized. The platform was evaluated for the extraction and quantification of caffeine from commercial coffee powders assessing different solvent types, extraction temperatures and times. For example, 141 {+-} 11 {mu}g caffeine (5 mg coffee powder) were extracted during a single extraction cycle using methanol as extraction solvent, whereas only 90 {+-} 11 were obtained performing the extraction in methylene chloride, applying the same reaction conditions (90 Degree-Sign C, 10 min). In multiple extraction experiments a total of {approx}150 {mu}g caffeine was extracted from 5 mg commercial coffee powder. In addition to the quantitative caffeine determination, a comparative qualitative analysis of the liquid phase coffee

  6. Twist seal for high-pressure vessels such as space shuttle rocket motors

    Science.gov (United States)

    von Pragenau, George L. (Inventor)

    1989-01-01

    Seals for sealing clevis and flange joints (14) of a solid rocket booster motor, and more particularly to a seal (30) which is twisted upon application of expansion forces to an edge seal (36). This twisting motion initially causes a leading edge seal (44) to be urged into sealing engagement with a surface (48) of an adjacent member (20) and thereafter, increasing fluid pressure on a pressurized side (64) of a seal (30) drives a broad sealing region (46) into sealing engagement with a surface (48).

  7. BWR Refill-Reflood Program, Task 4. 7 - model development: TRAC-BWR component models

    Energy Technology Data Exchange (ETDEWEB)

    Cheung, Y K; Parameswaran, V; Shaug, J C

    1983-09-01

    TRAC (Transient Reactor Analysis Code) is a computer code for best-estimate analysis for the thermal hydraulic conditions in a reactor system. The development and assessment of the BWR component models developed under the Refill/Reflood Program that are necessary to structure a BWR-version of TRAC are described in this report. These component models are the jet pump, steam separator, steam dryer, two-phase level tracking model, and upper-plenum mixing model. These models have been implemented into TRAC-B02. Also a single-channel option has been developed for individual fuel-channel analysis following a system-response calculation.

  8. Evaluation of ductile-brittle transition behavior with neutron irradiation in nuclear reactor pressure vessel steels using small punch test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M. C.; Lee, B. S. [KAERI, Taejon (Korea, Republic of); Oh, Y. J. [Hanbat National Univ., Taejon (Korea, Republic of)

    2003-10-01

    A Small Punch (SP) test was performed to evaluate the ductile-brittle transition temperature before and after neutron irradiation in Reactor Pressure Vessel (RPV) steels produced by different manufacturing (refining) processes. The results were compared to the standard transition temperature shifts from the Charpy test and Master Curve fracture toughness test in accordance with the ASTM standard E1921. The samples were taken from 1/4t location of the vessel thickness and machined into a 10x10x0.5mm dimension. Irradiation of the samples was carried out in the research reactor at KAERI (HANARO) at about 290 .deg. C of the different fluence levels respectively. SP tests were performed in the temperature range of RT to -196 .deg. C using a 2.4mm diameter ball. For the materials before and after irradiation, SP transition temperatures (T{sub sp}), which are determined at the middle of the upper and lower SP energies, showed a linear correlation with the Charpy index temperature, T{sub 41J}. T{sub sp} from the irradiated samples was increased as the fluence level increased and was well within the deviation range of the unirradiated data. The TSP had a correlation with the reference temperature (T{sub 0}) from the master curve method using a pre-cracked Charpy V-notched (PCVN) specimen.

  9. 高压液氢容器的研制%Development of High Pressure and Liquid Hydrogen Vessel

    Institute of Scientific and Technical Information of China (English)

    路兰卿; 于洋

    2013-01-01

      针对设计温度为-253℃、设计压力为17.7 MPa高压液氢贮存容器的设计、方案确定、加工制造等方面进行了详细介绍。容器承压结构为单层厚壁板卷,其绝热方式采用液氮夹套预冷屏和外堆积绝热的形式,目前该容器已成功完成了多次发动机的试验任务。%  This text introduces a high pressure liquid hydrogen vessel’s design, plan, manufacture etc which the design temperature is-253℃ and design pressure is 17.7MPa. The pressure bearing structure is one layer, insulation is LN2 precooling screen and the outer packing insulation. This vessel is successfully designed and applied in china.

  10. Safety Evaluation Report: Development of Improved Composite Pressure Vessels for Hydrogen Storage, Lincoln Composites, Lincoln, NE, May 25, 2010

    Energy Technology Data Exchange (ETDEWEB)

    Fort, III, William C.; Kallman, Richard A.; Maes, Miguel; Skolnik, Edward G.; Weiner, Steven C.

    2010-12-22

    Lincoln Composites operates a facility for designing, testing, and manufacturing composite pressure vessels. Lincoln Composites also has a U.S. Department of Energy (DOE)-funded project to develop composite tanks for high-pressure hydrogen storage. The initial stage of this project involves testing the permeation of high-pressure hydrogen through polymer liners. The company recently moved and is constructing a dedicated research/testing laboratory at their new location. In the meantime, permeation tests are being performed in a corner of a large manufacturing facility. The safety review team visited the Lincoln Composites site on May 25, 2010. The project team presented an overview of the company and project and took the safety review team on a tour of the facility. The safety review team saw the entire process of winding a carbon fiber/resin tank on a liner, installing the boss and valves, and curing and painting the tank. The review team also saw the new laboratory that is being built for the DOE project and the temporary arrangement for the hydrogen permeation tests.

  11. Methodological developments in the field of structural integrity analyses of large scale reactor pressure vessels in Hungary

    Directory of Open Access Journals (Sweden)

    Tamás Fekete

    2016-03-01

    Full Text Available Buildings, structures and systems of large scale and high value (e.g. conventional and nuclear power plants, etc. are designed for a certain, limited service lifetime. If the standards and guidelines of the time are taken into account during the design process, the resulting structures will operate safely in most cases. However, in the course of technical history there were examples of unusual, catastrophic failures of structures, even resulting in human casualties. Although the concept of Structural Integrity first appeared in industrial applications only two-three decades ago, its pertinence has been growing higher ever since. Four nuclear power generation units have been constructed in Hungary, more than 30 years ago. In every unit, VVER-440 V213 type light-water cooled, light-water moderated, pressurized water reactors are in operation. Since the mid-1980s, Pressurized Thermal Shock (PTS analyses of Reactor Pressure Vessels (RPV have been conducted in Hungary, where the concept of structural integrity was the basis of research and development. In the first part of the paper, a short historic overview is given, where the origins of the Structural Integrity concept are presented, and the beginnings of Structural Integrity in Hungary are summarized. In the second part, a new conceptual model of Structural Integrity is introduced. In the third part, a brief description of the VVER-440 V213 type RPV and its surrounding primary system is presented. In the fourth part, a conceptual model developed for PTS Structural Integrity Analyses is explained.

  12. Stress categorization in nozzle to pressure vessel connections finite elements models; Categorizacao de tensoes em modelos de elementos finitos de conexoes bocal-vaso de pressao

    Energy Technology Data Exchange (ETDEWEB)

    Albuquerque, Levi Barcelos de

    1999-07-01

    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

  13. Structural evaluation of the Shippingport Reactor Pressure Vessel and Neutron Shield Tank package for impact and puncture loads

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-1 and TSE-2

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.

    1976-09-01

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and two thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. The PWR calculations indicated that under some circumstances crack propagation could be expected and that experiments should be conducted for cracks that would have the potential for propagation at least halfway through the wall.

  15. Shape optimization on the nozzle of a spherical pressure vessel using the ranked bidirectional evolutionary structural optimization

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Shin; Ryu, Chung Hyun [Chungnam National Univ., Taejon (Korea, Republic of)

    2001-07-01

    To reduce stress concentration around the intersection between a spherical pressure vessel and a cylindrical nozzle under various load conditions using less material, the optimization for the distribution of reinforcement has researched. The Ranked Bidirectional Evolutionary Structural Optimization(R-BESO) method is developed recently, which adds elements based on a rank, and the performance indicator which can estimate a fully stressed model. The R-BESO method can obtain the optimum design using less iteration number than iteration number of the BESO. In this paper, the optimized intersection shape is sought using R-BESO method for a flush and a protruding nozzle. The considered load cases are a radial compression, torque and shear force.

  16. Structural model testing for prestressed concrete pressure vessels: a study of grouted vs nongrouted posttensioned prestressing tendon systems. [HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Naus, D.J.

    1979-04-01

    Nongrouted tendons are predominantly used in this country as the prestressing system for prestressed concrete pressure vessels (PCPVs) because they are more easily surveyed to detect reductions in prestressing level and distress such as results from corrosion. Grouted tendon systems, however, offer advantages which may make them cost-effective for PCPV applications. Literature was reviewed to (1) provide insight on the behavior of grouted tendon system, (2) establish performance histories for structures utilizing grouted tendons, (3) examine corrosion protection procedures for prestressing tendons, (4) identify arguments for and against using grouted tendons, and (5) aid in the development of the experimental investigation. The experimental investigation was divided into four phases: (1) grouted-nongrouted tendon behavior, (2) evaluation of selected new material systems, (3) bench-scale corrosion studies, and (4) preliminary evaluation of acoustic emission techniques for monitoring grouted tendons in PCPVs. The groutability of large tendon systems was also investigated.

  17. Influence of structural parameters on the tendency of VVER-1000 reactor pressure vessel steel to temper embrittlement

    Science.gov (United States)

    Gurovich, B.; Kuleshova, E.; Zabusov, O.; Fedotova, S.; Frolov, A.; Saltykov, M.; Maltsev, D.

    2013-04-01

    In this paper the influence of structural parameters on the tendency of steels to reversible temper embrittlement was studied for assessment of performance properties of reactor pressure vessel steels with extended service life. It is shown that the growth of prior austenite grain size leads to an increase of the critical embrittlement temperature in the initial state. An embrittlement heat treatment at the temperature of maximum manifestation of temper embrittlement (480 °C) shifts critical embrittlement temperature to higher values due to the increase of the phosphorus concentration on grain boundaries. There is a correlation between phosphorus concentration on boundaries of primary austenite grains and the share of brittle intergranular fracture (that, in turn, depends on impact test temperature) in the fracture surfaces of the tested Charpy specimens.

  18. The effect of complex exercise rehabilitation program on body composition, blood pressure, blood sugar, and vessel elasticity in elderly women with obesity

    Science.gov (United States)

    Lee, Eun-Ok; Lee, Kwon-Ho; Kozyreva, Olga

    2013-01-01

    The purpose of this study is to identify what kind of effects complex exercise rehabilitation program has on body composition of female, blood pressure, blood sugar, blood vessel elasticity and find more effective complex exercise program for elderly females. The subjects are selected 30 females applicants in exercise program in City of G and not restricted in mobility to perform the exercise without any particular disorders. Exercise program is a combination of aerobic and strength training with different ratio, for the first 6 months focused on strength training complex exercise, and for next 6 months focused on aerobic exercise. Except for strength training and aerobic exercise, durations for strength, rest, and wrapping-up are equal. The frequency of experiments is 90 min each, 2 times per a week. Body composition, blood pressure, and blood vessel elasticity are tested pre and post experiment to compare the effectiveness of both complex exercises. As results, in the complex exercise program focused on strength training, weight, percent body fat, fat mass, waist hip ratio, systolic blood pressure, and diastolic pressure increased. Blood vessel elasticity maintained its level or slightly decreased. In the complex exercise focused on aerobic exercise, weight, percent body fat, fat mass, waist hip ratio, systolic pressure, and diastolic pressure decreased. Blood vessel elasticity on left foot and right foot are slightly different. Therefore, aerobic exercise is more effective than strength training for old obese females. PMID:24409428

  19. PWR and BWR spent fuel assembly gamma spectra measurements

    Science.gov (United States)

    Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

    2016-10-01

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  20. Determination of stresses caused by fluctuation of acoustic load in the steam dryers of a BWR; Determinacion de esfuerzos originados por fluctuacion de carga acustica en los secadores de vapor de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Centeno P, J.; Quezada G, S.; Prieto G, A.; Vazquez R, A.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico); Nunez C, A., E-mail: javcuami26@hotmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2014-10-15

    The extended power up-rate (EPU) in a nuclear power plant cause various problems in BWR components also in the steam system. This due to increased steam flow generated in the reactor and is conveyed to the turbine by the four main steam lines (MSL). One of the most serious problems is the generation of acoustic pressure loads in the metal structure of the steam dryer which eventually leads to fatigue failure and even the appearance of cracks, and in turn it causes loose parts that are entrained by the steam and transported in the MSL. This problem is due to the fluctuation of load acoustics caused by the union of the safety or relief valves (SRV) with the MSL, spreading through these to reach the reactor pressure vessel (RPV) where the effect of resonance of the acoustic wave is amplified and impacts directly in the supporting structure of the steam dryer, skirt and the panels where the mixture liquid-steam is dried, by centrifugation effect and runoff of liquid water. Efforts in the steam dryer operating conditions of EPU for two cases will be analyzed in this work, the first is before the installation of Acoustic Side Branch (ASB), and in the second case we consider the installation of said ASB in the standpipes of SRV. The analysis was performed with numerical experiments on a platform for computational fluid dynamics with virtual geometries previously designed based on the actual components of the reactor and steam system. The model to study is delimited by the top of the RPV, the steam dryer and a section of each of the four MSL with ten standpipes of SRV. With the obtained data and considering the mechanical-structural properties of the steam dryer material, we can evaluate the mechanical resistance to impacts by acoustic pressure load and its possible deformation or cracking. (Author)

  1. A Review of Energy Release Processes from the Failure of Pneumatic Pressure Vessels

    Science.gov (United States)

    1988-08-01

    RT) is not a good approximation. There are several equations cf state that can be used for real gases (e.g., Van der Waal’s, Beattie - Bridgeman ...The gas pressure can be written in terms of an appropriate equation of state for either an ideal or real gas. Initial fragment velocity is...assumption3 reduce Equation (1) to: -w - AE - AU (2) The ideal gas law states that foz the expailsion of a gas: W - -C, AT (3) where: C, - constant

  2. CFD analysis of a regular sector of the ITER vacuum vessel. Part I: Flow distribution and pressure drop

    Energy Technology Data Exchange (ETDEWEB)

    Savoldi Richard, L., E-mail: laura.savoldi@polito.it [Dipartimento Energia, Politecnico di Torino, 10129 Torino (Italy); Bonifetto, R. [Dipartimento Energia, Politecnico di Torino, 10129 Torino (Italy); Zanino, R., E-mail: roberto.zanino@polito.it [Dipartimento Energia, Politecnico di Torino, 10129 Torino (Italy); Corpino, S.; Obiols-Rabasa, G. [Dipartimento di Ingegneria Meccanica e Aerospaziale, Politecnico di Torino, 10129 Torino (Italy); Izquierdo, J. [F4E, Barcelona (Spain); Le Barbier, R.; Utin, Y. [ITER IO, Cadarache (France)

    2013-12-15

    The 3D steady-state Computational Fluid Dynamics (CFD) analysis of the ITER vacuum vessel (VV) regular sector no. 5 is presented, starting from the CATIA models and using a suite of tools from the commercial software ANSYS FLUENT{sup ®}. The peculiarity of the problem is linked to the wide range of spatial scales involved in the analysis, from the millimeter-size gaps between in-wall shielding (IWS) plates to the more than 10 m height of the VV itself. After performing several simplifications in the geometrical details, a computational mesh with ∼50 million cells is generated and used to compute the steady-state pressure and flow fields from a Reynolds-Averaged Navier–Stokes model with SST k-ω turbulence closure. The coolant mass flow rate turns out to be distributed 10% through the inboard and the remaining 90% through the outboard. The toroidal and poloidal ribs present in the VV structure constitute significant barriers for the flow, giving rise to large recirculation regions. The pressure drop is mainly localized in the inlet and outlet piping.

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

    Energy Technology Data Exchange (ETDEWEB)

    Naus, D.J.

    1981-05-01

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

  4. Survey of welding processes for field fabrication of 2 1/4 Cr-1 Mo steel pressure vessels. [128 references

    Energy Technology Data Exchange (ETDEWEB)

    Grotke, G.E.

    1980-04-01

    Any evaluation of fabrication methods for massive pressure vessels must consider several welding processes with potential for heavy-section applications. These include submerged-arc and shielded metal-arc, narrow-joint modifications of inert-gas metal-arc and inert-gas tungsten-arc processes, electroslag, and electron beam. The advantage and disadvantages of each are discussed. Electroslag welding can be dropped from consideration for joining of 2 1/4 Cr-1 Mo steel because welds made with this method do not provide the required mechanical properties in the welded and stress relieved condition. The extension of electron-beam welding to sections as thick as 4 or 8 inches (100 or 200 mm) is too recent a development to permit full evaluation. The manual shielded metal-arc and submerged-arc welding processes have both been employed, often together, for field fabrication of large vessels. They have the historical advantage of successful application but present other disadvantages that make them otherwise less attractive. The manual shielded metal-arc process can be used for all-position welding. It is however, a slow and expensive technique for joining heavy sections, requires large amounts of skilled labor that is in critically short supply, and introduces a high incidence of weld repairs. Automatic submerged-arc welding has been employed in many critical applications and for welding in the flat position is free of most of the criticism that can be leveled at the shielded metal-arc process. Specialized techniques have been developed for horizontal and vertical position welding but, used in this manner, the applications are limited and the cost advantage of the process is lost.

  5. BIOASSAY VESSEL FAILURE ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    Vormelker, P

    2008-09-22

    Two high-pressure bioassay vessels failed at the Savannah River Site during a microwave heating process for biosample testing. Improper installation of the thermal shield in the first failure caused the vessel to burst during microwave heating. The second vessel failure is attributed to overpressurization during a test run. Vessel failure appeared to initiate in the mold parting line, the thinnest cross-section of the octagonal vessel. No material flaws were found in the vessel that would impair its structural performance. Content weight should be minimized to reduce operating temperature and pressure. Outer vessel life is dependent on actual temperature exposure. Since thermal aging of the vessels can be detrimental to their performance, it was recommended that the vessels be used for a limited number of cycles to be determined by additional testing.

  6. Main mechanisms of material properties degradation under reactor pressure vessel operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Karzov, Georgy; Timofeev, Boris [Central Research Inst. of Structural Materials ' prometey' , St. Petersburg (Russian Federation)

    1999-07-01

    In the process of NPP equipment operation materials are subjected to a prolonged influence of loads, associated with the variation of inner pressure and temperature under various conditions. Each equipment element damage is associated with some material fracture mechanism. For NPP equipment the mechanisms of irreversible damage accumulation are related with: irradiation embrittlement, thermal and strain aging, fatigue damages from mechanical and thermal loading, stress corrosion and fatigue corrosion, creep and thermal relaxation stresses, erosion and weak, thermal shock. The basic tasks of specialists working in the sphere of the provision of reliability and service life of nuclear power equipment are not only the determination of the main mechanisms of damages and reasons of their appearance, but also the study of methods which would permit to control these properties completely. By giving some examples of Russian NPP equipment with VVER-440 and VVER-1000 reactors the paper presents most typical degradation mechanisms of equipment material properties, including weldments, in the process of operation and methods to recover by using various technological means. (author)

  7. Development of automated welding process for field fabrication of thick walled pressure vessels. Fourth quarter technical progress report for period ending September 28, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    Progress is reported in research aimed at optimizing an automated welding process for the field fabrication of thick-walled pressure vessels and for evaluating the welded joints. Information is included on the welding equipment, mechanical control of the process, joint design, filler wire optimization, in-process nondestructive testing of welds, and repair techniques. (LCL)

  8. Intracranial pressure elevation reduces flow through collateral vessels and the penetrating arterioles they supply. A possible explanation for 'collateral failure' and infarct expansion after ischemic stroke.

    Science.gov (United States)

    Beard, Daniel J; McLeod, Damian D; Logan, Caitlin L; Murtha, Lucy A; Imtiaz, Mohammad S; van Helden, Dirk F; Spratt, Neil J

    2015-05-01

    Recent human imaging studies indicate that reduced blood flow through pial collateral vessels ('collateral failure') is associated with late infarct expansion despite stable arterial occlusion. The cause for 'collateral failure' is unknown. We recently showed that intracranial pressure (ICP) rises dramatically but transiently 24 hours after even minor experimental stroke. We hypothesized that ICP elevation would reduce collateral blood flow. First, we investigated the regulation of flow through collateral vessels and the penetrating arterioles arising from them during stroke reperfusion. Wistar rats were subjected to intraluminal middle cerebral artery (MCA) occlusion (MCAo). Individual pial collateral and associated penetrating arteriole blood flow was quantified using fluorescent microspheres. Baseline bidirectional flow changed to MCA-directed flow and increased by >450% immediately after MCAo. Collateral diameter changed minimally. Second, we determined the effect of ICP elevation on collateral and watershed penetrating arteriole flow. Intracranial pressure was artificially raised in stepwise increments during MCAo. The ICP increase was strongly correlated with collateral and penetrating arteriole flow reductions. Changes in collateral flow post-stroke appear to be primarily driven by the pressure drop across the collateral vessel, not vessel diameter. The ICP elevation reduces cerebral perfusion pressure and collateral flow, and is the possible explanation for 'collateral failure' in stroke-in-progression.

  9. Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR); Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua hirviente (BWR)

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes C, P

    2003-07-01

    This work presents the results of the assays carried out in the Laboratory of Hot Cells of the National Institute of Nuclear Research (ININ) to a type test tube Compact Tension (CT), built in steel austenitic stainless type 304L, simulating those conditions those that it operates a Boiling Water Reactor (BWR), at temperature 288 C and pressure of 8 MPa, to determine the speed to which the cracks spread in this material that is of the one that different components of a reactor are made, among those that it highlights the reactor core vessel. The application of the Hydrogen Chemistry of the Water is presented (HWC) that is one alternative to diminish the corrosion effect low stress in the component, this is gets controlling the quantity of oxygen and of hydrogen as well as the conductivity of the water. The rehearsal is made following the principles of the Mechanics of Elastic Lineal Fracture (LEFM) that considers a crack of defined size with little plastic deformation in the tip of this; the measurement of crack advance is continued with the technique of potential drop of direct current of alternating signal, this is contained inside the standard Astm E-647 (Method of Test Standard for the Measurement of Speed of Growth of Crack by fatigue) that is the one that indicates us as carrying out this test. The specifications that should complete the test tubes that are rehearsed as for their dimensions, it forms, finish and determination of mechanical properties (tenacity to the fracture mainly) they are contained inside the norm Astm E-399, the one which it is also based on the principles of the fracture mechanics. The obtained results were part of a database to be compared with those of other rehearsals under different conditions, Normal Chemistry of the Water (NWC) and it dilutes with high content of O{sub 2}; to determine the conditions that slow more the phenomena of stress corrosion cracking, as well as the effectiveness of the used chemistry and of the method of

  10. Simulation of a large break loss of coolant (LBLOCA), without actuation of the emergency injection systems (ECCS) for a BWR-5; Simulacion de un escenario de perdida de refrigerante grande (LBLOCA), sin actuacion de los sistemas de inyeccion de emergencia (ECCS) para un reactor BWR-5

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Mugica R, C. A.; Lopez M, R., E-mail: jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

    2015-09-15

    In this paper the analysis of scenario for the loss of coolant case was realized with break at the bottom of a recirculation loop of a BWR-5 with containment type Mark II and a thermal power of 2317 MWt considering that not have coolant injection. This in order to observe the speed of progression of the accident, the phenomenology of the scenario, the time to reach the limit pressure of containment venting and the amount of radionuclides released into the environment. This simulation was performed using the MELCOR code version 2.1. The scenario posits a break in one of the shear recirculation loops. The emergency core cooling system (ECCS) and the reactor core isolation cooling (Rcic) have not credit throughout the event, which allowed achieve greater severity on scenario. The venting of the primary containment was conducted via valve of 30 inches instead of the line of 24 inches of wet well, this in order to have a larger area of exhaust of fission products directly to the reactor building. The venting took place when the pressure in the primary containment reached the 4.5 kg/cm{sup 2} and remained open for the rest of the scenario to maximize the amount released of radionuclides to the atmosphere. The safety relief valves were considered functional they do not present mechanical failure or limit their ability to release pressure due to the large number of performances in safety mode. The results of the analysis covers about 48 hours, time at which the accident evolution was observed; behavior of level, pressure in the vessel and the fuel temperature profile was analyzed. For progression of the scenario outside the vessel, the pressure and temperature of the primary containment, level and temperature of the suppression pool, the hydrogen accumulation in the container and the radionuclides mass released into the atmosphere were analyzed. (Author)

  11. Simulation of the injection system of cooling water to low pressure (Lpci) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de inyeccion de agua de refrigeracion a baja presion (LPCI) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Delgado C, R. A.; Lopez S, E.; Chavez M, C., E-mail: renedelgado2015@hotmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    The present article describes the modeling and simulation of the Injection System of Cooling Water to Low Pressure (Lpci) for the nuclear power plant of Laguna Verde. Is very important to be able to predict the behavior of the nuclear plant in the case of an emergency stop, and while nearer to the reality are the results of a simulation, better is the safety protocol that can be devised. In the Engineering Faculty of the UNAM at the present is had logical models of the safety systems, but due to the nature of the same, these simulations do not provide of the quantity of enough information to be able to reproduce with more accuracy the behavior of the Lpci in the case of a severe accident. For this reason, the RELAP code was used for the flows modeling, components and structures of heat transfers in relation to the system Lpci. The modeling of the components is carried out with base on technical information of the nuclear plant and the results will be corroborated with information in reference documents as the Rasp (the Reactor analysis support package) and the Fsar (Final safety analysis report) for the nuclear power plant of Laguna Verde. (Author)

  12. Simulation of the aspersion system of the core at high pressure (HPCS) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de aspersion del nucleo alta presion (HPCS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas O, D.; Chavez M, C., E-mail: danmirnyi@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    A high-priority topic for the nuclear industry is the safety, consequently a nuclear power plant should have the emergency systems of cooling of the core (ECCS), designed exclusively to enter in operation in the event of an accident with coolant loss, including the design base accident. The objective of the aspersion system of the core at high pressure (HPCS) is to provide in an autonomous way the cooling to the core maintaining for if same the coolant inventory even when a small break is presented that does not allow the depressurization of the reactor and also avoiding excessive temperatures that affect the shielding of the fuel. The present work describes the development of the model and the simulation of the HPCS using the RELAP/SCDAP code. During the process simulation, for the setting in march of the system HPCS in an accident with coolant loss is necessary to implement the main components of the system taking into account what unites them, the main pump, the filled pump, the suction and injection valves, pipes and its water sources that can be condensed storage tanks and the suppression pool. The simulation of this system will complement the model with which counts the Analysis Laboratory in Nuclear Reactors Engineering of the UNAM regarding to the nuclear power plant of Laguna Verde which does not have a detailed simulation of the emergency cooling systems. (Author)

  13. Neutron radiation embrittlement studies in support of continued operation, and validation by sampling of Magnox reactor steel pressure vessels and components

    Energy Technology Data Exchange (ETDEWEB)

    Jones, R.B.; Bolton, C.J. [Magnox Electric plc, Berkeley Centre, Glos (United Kingdom)

    1997-02-01

    Magnox steel reactor pressure vessels differ significantly from US LWR vessels in terms of the type of steel used, as well as their operating environment (dose level, exposure temperature range, and neutron spectra). The large diameter ferritic steel vessels are constructed from C-Mn steel plates and forgings joined together with manual metal and submerged-arc welds which are stress-relieved. All Magnox vessels are now at least thirty years old and their continued operation is being vigorously pursued. Vessel surveillance and other programmes are summarized which support this objective. The current understanding of the roles of matrix irradiation damage, irradiation-enhanced copper impurity precipitation and intergranular embrittlement effects is described in so far as these influence the form of the embrittlement and hardening trend curves for each material. An update is given on the influence of high temperature exposure, and on the role of differing neutron spectra. Finally, the validation offered by the results of an initial vessel sampling exercise is summarized together with the objectives of a more extensive future sampling programme.

  14. Stress analysis in a non axisymmetric loaded reactor pressure vessel; Verificacao de tensoes em um vaso de pressao nuclear com carregamentos nao-axissimetricos

    Energy Technology Data Exchange (ETDEWEB)

    Albuquerque, Levi Barcelos; Assis, Gracia Menezes V. de [Coordenadoria para Projetos Especiais (COPESP), Sao Paulo, SP (Brazil); Miranda, Carlos Alexandre J.; Cruz, Julio Ricardo B.; Mattar Neto, Miguel [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

    1995-12-31

    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) 4 refs., 15 figs., 7 tabs.

  15. Evaluation of sildenafil pressurized metered dose inhalers as a vasodilator in umbilical blood vessels of chicken egg embryos.

    Science.gov (United States)

    Sawatdee, Somchai; Hiranphan, Phetai; Laphanayos, Kampanart; Srichana, Teerapol

    2014-01-01

    Sildenafil citrate is a selective phosphodiesterase-5 inhibitor used for the treatment for erectile dysfunction and pulmonary hypertension. The delivery of sildenafil directly to the lung could have several advantages over conventional treatments for pulmonary hypertension because of the local delivery, a more rapid onset of response, and reduced side effects. The major problem of sildenafil citrate is its limited solubility in water. Sildenafil citrate was complexed with cyclodextrins (CDs) to enhance its water solubility prior to development as an inhaled preparation. Four sildenafil citrate inhaled formulations were prepared with the aid of HP-β-CD (#1), α-CD (#2) and γ-CD (#3) and their effects were compared with the formulations without CDs (#4). The sildenafil citrate pressurized metered dose inhalers (pMDI) used ethanol as a solvent, PEG400 as a stabilizing agent, sorbitan monooleate as a surfactant and HFA-134a as a propellant. All formulations consisted of sildenafil citrate equivalent to a sildenafil content of 20μg/puff. These products were evaluated according to a standard guideline of inhalation products. Vasodilation testing was performed to investigate the efficacy of sildenafil pMDIs in relieving a vasoconstricted umbilical blood vessel of the chicken egg embryo. The sildenafil contents of the pMDI formulations #1-#3 were within the acceptance criteria (80-120%). The emitted doses (ED) were 102.3±11.5%, the fine particle fractions (FPF) were 60.5±5.6% and the mass median aerodynamic diameters (MMAD) were 2.3±0.3μm. The vasodilatory activity of those formulations reduced umbilical blood pressure by 67.1-73.7% after treatment by intravenous injection whereas only a 50.1-58.0% reduced blood pressure was obtained after direct spraying of the sildenafil pMDI containing CDs. With sildenafil formulations of a pMDI without CD the blood pressure was reduced by only 39.0% (P-valuevessels of chicken egg embryos after spraying sildenafil-CDs pMDIs was

  16. A BWR licensing experience in the USA

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J.; Ogura, C. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Arai, K. [Toshiba Corporation, Yokohama, Kanagawa (Japan); Thomas, S.; Mookhoek, B., E-mail: jim.powers@toshiba.com [Nuclear Innovation North America, Lake Jackson, Texas (United States)

    2015-09-15

    The US-Advanced Boiling Water Reactor (A BWR), certified by the United States Nuclear Regulatory Commission (US NRC), is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The STP3-4 project has finished the US NRC technical review of the Cola through the final meeting of the Advisory Committee on Reactor Safeguards (ACRS), and the Final Safety Evaluation Report (FSER) is scheduled to be issued by the US NRC in the middle of 2015. The next steps are to support the Mandatory Hearing process, and voting by the NRC commissioners on the motion to grant the Combined License, which is scheduled beginning of 2016 according to US NRC schedule as of March 30, 2015. This paper summarizes the history and progress of the US-A BWR licensing, including the experiences of the Licensee, Nina, and Toshiba as the Epc team worked through the Code of Federal Regulations Title 10 (10-Cfr) Part 52 process, and provides some perspectives on how the related licensing material would also be of value within a 10-Cfr Part 50, two-step process to minimize schedule and financial risks which could arise from ongoing technical developments and regulatory reviews. (Author)

  17. LBB application in Swedish BWR design

    Energy Technology Data Exchange (ETDEWEB)

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P. [ABB Atom, Vaesteras (Sweden)

    1997-04-01

    The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as these installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions.

  18. Evaluation of neutron irradiation embrittlement in the Korean reactor pressure vessel steels(I) (1st progress report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jun Hwa; Lee, Bong Sang; Park, Duck Gun; Byun, Tak Sang; Kim, Joo Hag; Oh, Yong Jun; Yoon, Ji Hyun; Chi, Sei Hwan; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-10-01

    The SA508-3 reactor pressure vessel materials degrade due to the application at high temperature, high pressure, and neutron irradiation. In the present study it is planned to examine the effects of neutron irradiation on the properties for assessing the integrity of domestic reactors. The key tests are the Charpy impact test, tensile test, static and dynamic fracture toughness test, J-R test. The additional tests for obtaining basic material properties, such as micro-hardness, microstructural properties, small punch energy etc., are also performed. The irradiation tests are being performed at HANARO of KAERI through the instrumented capsules designed by KAERI and the post-irradiation tests are being performed at IMEF(Irradiated Material Evaluation Facility) of material (UCN-4), Si+Al (YGN-5), UCN-4 weld metal, and UCN-4 HAZ. In the irradiation test the temperature should be controlled in the range of 290 {+-} 10 deg C and the test materials would be irradiated to 2 to 3 neutron fluence levels including the end-of-life fluence. The status of performing this project is that (1) the key data on mechanical properties, mainly related to the fracture toughness, of the unirradiated materials have been obtained, (2) the irradiation of the 1st instrumented capsule, a preliminary test capsule containing miniature specimens, has been completed and is being stored for testing in IMEF, and (3) the 2nd instrumented capsule is being manufactured and will be irradiated in the beginning or 1999. This report includes mainly the experimental methods and results. The status of the design and manufacturing of the instrumented capsules and specimens was also briefly described. (author). 13 refs., 15 figs., 10 tabs.

  19. Analysis of containment venting following a core damage at a BWR Mark I using THALES-2

    Energy Technology Data Exchange (ETDEWEB)

    Widodo, Surip [Nuclear Safety Technology Development Center, National Nuclear Energy Agency (BATAN), Tangerang (Indonesia); Ishikawa, Jun; Muramatsu, Ken [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Sakamoto, Toru [Toshiba Advanced System Co., Kawasaki, Kanagawa (Japan)

    2000-11-01

    Analysis of containment venting following a core damage at a boiling water reactor (BWR) Mark I using THALES-2 was performed. In this analysis, the effect of various parameters, namely, the areas of the vent path, containment venting pressure, and accident sequences on the containment thermodynamic response, and radionuclide transport and release in the containment venting at a BWR was examined. The code THALES-2B developed by the Japan Atomic Energy Research Institute (JAERI) was used in this analysis. The model plant in this analysis was the Browns Ferry plant. From this analysis was found that the 4-inch pipe of containment venting flow path is sufficient to maintain the containment pressure in the specified range if the containment was pressurized by the decay heat power. The entrainment by the pool swelling as well as by the flashing was not occurred during the containment venting. The source terms are not sensitive to the variation of containment venting flow path area. The containment venting pressure operation setting point has important rule in the containment venting. In the containment venting, the source terms are not sensitive to the accident sequence, except for Sr source term. In order to get better understanding on the containment venting strategy, the following analyses are necessary. Analyses of accident sequence which has a high power such as anticipated transient without scram are necessary, as well as analyses of accident sequence which pressurize the containment before the core damage. (author)

  20. ALICE HMPID Radiator Vessel

    CERN Multimedia

    2003-01-01

    View of the radiator vessels of the ALICE/HMPID mounted on the support frame. Each HMPID module is equipped with 3 indipendent radiator vessels made out of neoceram and fused silica (quartz) windows glued together. The spacers inside the vessel are needed to stand the hydrostatic pressure. http://alice-hmpid.web.cern.ch/alice-hmpid

  1. Neutron fluence at the reactor pressure vessel wall - a comparison of French and German procedures and strategies in PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Tricot, N. [Institut de Radioprotection et de Surete Nucleaire, IRSN/DES/SECCA, 92 - Fontenay aux Roses (France); Jendrich, U. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany)

    2003-01-01

    While the neutrons within the core may take part in the chain reaction, those neutrons emitted from the core are basically lost for the energy production. This 'neutron leakage' represents a loss of fuel efficiency and causes neutron embrittlement of the reactor pressure vessel (RPV) wall. The latter raises safety concerns, needs to be monitored closely and may necessitate mitigating measures. There are different strategies to deal with these two undesirable effects: The neutron emission may be reduced to some extent all around the core or just at the 'hot spots' of RPV embrittlement by tailored core loading patterns. A higher absorption rate of neutrons may also be achieved by a larger water gap between the core and the RPV. In this paper the inter-relations between the distribution of neutron flux, core geometry, core loading strategy, RPV embrittlement and its surveillance are discussed at first. Then the different strategies followed by the German and French operators are described. Finally the conclusions will highlight the communalities and differences between these strategies as different approaches to the same problem of safety as well as economy. (authors)

  2. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  3. Investigation of a Coolant Mixing Phenomena within the Reactor Pressure Vessel of a VVER-1000 Reactor with Different Simulation Tools

    Directory of Open Access Journals (Sweden)

    V. Sánchez

    2010-01-01

    Full Text Available The Institute of Neutron Physics and Reactor Technology (INR is involved in the qualification of coupled codes for reactor safety evaluations, aiming to improve their prediction capability and acceptability. In the frame of the VVER-1000 Coolant Transient Benchmark Phase 1, RELAP5/PARCS has been extensively assessed. Phase 2 of this benchmark was focused on both multidimensional thermal hydraulic phenomena and core physics. Plant data will be used to qualify the 3D models of TRACE and RELAP5/CFX, which were coupled for this purpose. The developed multidimensional models of the VVER-1000 reactor pressure vessel (RPV as well as the performed calculations will be described in detail. The predicted results are in good agreement with experimental data. It was demonstrated that the chosen 3D nodalization of the RPV is adequate for the description of the coolant mixing phenomena in a VVER-1000 reactor. Even though only a 3D coarse nodalization is used in TRACE, the integral results are comparable to those obtained by RELAP5/CFX.

  4. An object kinetic Monte Carlo model for the microstructure evolution of neutron-irradiated reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Messina, Luca; Olsson, Paer [KTH Royal Institute of Technology, Stockholm (Sweden); Chiapetto, Monica [SCK - CEN, Nuclear Materials Science Institute, Mol (Belgium); Unite Materiaux et Transformations (UMET), UMR 8207, Universite de Lille 1, ENSCL, Villeneuve d' Ascq (France); Becquart, Charlotte S. [Unite Materiaux et Transformations (UMET), UMR 8207, Universite de Lille 1, ENSCL, Villeneuve d' Ascq (France); Malerba, Lorenzo [SCK - CEN, Nuclear Materials Science Institute, Mol (Belgium)

    2016-11-15

    This work presents a full object kinetic Monte Carlo framework for the simulation of the microstructure evolution of reactor pressure vessel (RPV) steels. The model pursues a ''gray-alloy'' approach, where the effect of solute atoms is seen exclusively as a reduction of the mobility of defect clusters. The same set of parameters yields a satisfactory evolution for two different types of alloys, in very different irradiation conditions: an Fe-C-MnNi model alloy (high flux) and a high-Mn, high-Ni RPV steel (low flux). A satisfactory match with the experimental characterizations is obtained only if assuming a substantial immobilization of vacancy clusters due to solute atoms, which is here verified by means of independent atomistic kinetic Monte Carlo simulations. The microstructure evolution of the two alloys is strongly affected by the dose rate; a predominance of single defects and small defect clusters is observed at low dose rates, whereas larger defect clusters appear at high dose rates. In both cases, the predicted density of interstitial loops matches the experimental solute-cluster density, suggesting that the MnNi-rich nanofeatures might form as a consequence of solute enrichment on immobilized small interstitial loops, which are invisible to the electron microscope. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  5. Microstructural behavior of VVER-440 reactor pressure vessel steels under irradiation to neutron fluences beyond the design operation period

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Shtrombakh, Ya. I.; Nikolaev, Yu. A.; Pechenkin, V. A.

    2005-06-01

    Electron-microscopy and fractographic studies of the surveillance specimens from base and weld metals of VVER-440/213 reactor pressure vessel (RPV) in the original state and after irradiations to different fast neutron fluences from ˜5 × 10 23 n m -2 ( E > 0.5 MeV) up to over design values have been carried out. The maximum specimens irradiation time was 84 480 h. It is shown that there is an evolution in radiation-induced structural behavior with radiation dose increase, which causes a change in relative contribution of the mechanisms responsible for radiation embrittlement of RPV materials. Particularly, radiation coalescence of copper-enriched precipitates and extensive density increase of dislocation loops was observed. Increase in dislocation loop density was shown to provide the dominant contribution to radiation hardening at the late irradiation stages (after reaching double the design end-of-life neutron fluence of ˜4 × 10 24 n m -2). The fracture mechanism of the base metal at those stages was observed to change from transcrystalline to intercrystalline.

  6. Subaquatic, pressure vessels and LPG storage spheres internal inspection; Inspecao interna de esfera utilizando mergulho como acesso

    Energy Technology Data Exchange (ETDEWEB)

    Filgueira Filho, Rafael; Monteiro, Ayres [PETROBRAS, Rio de Janeiro, RJ (Brazil)

    2005-07-01

    Minimizing shut-down costs is a widespread target in the oil and gas industry. The use of new inspection techniques is one of the ways for that. This work presents a new procedure for internal inspections in pressure vessels by the non destructive testing - NDT, ACFM, using industrial diving techniques. As a pioneer experience, this method was applied in the inspection of the internal parts of the LPG sphere tank 5101 at PETROBRAS Transporte S.A. - TRANSPETRO, in Jequie's Terminal, in the state of Bahia, in december, 2003. This new method allows the reduction of indirect costs related to operational unavailability of the equipment, by the reduction of the shut-down time in approximately 50%, when compared to the demanded shut down time, when using scaffolds for accessing the internal parts. Despite of direct costs are still higher with the new methodology, this paper demonstrates the economical feasibility of this new method, based on the savings obtained with the fastest return of the equipment to operation. (author)

  7. The cryogenic bonding evaluation at the metallic-composite interface of a composite overwrapped pressure vessel with additional impact investigation

    Science.gov (United States)

    Clark, Eric A.

    A bonding evaluation that investigated the cryogenic tensile strength of several different adhesives/resins was performed. The test materials consisted of 606 aluminum test pieces adhered to a wet-wound graphite laminate in order to simulate the bond created at the liner-composite interface of an aluminum-lined composite overwrapped pressure vessel. It was found that for cryogenic applications, a flexible, low modulus resin system must be used. Additionally, the samples prepared with a thin layer of cured resin -- or prebond -- performed significantly better than those without. It was found that it is critical that the prebond surface must have sufficient surface roughness prior to the bonding application. Also, the aluminum test pieces that were prepared using a surface etchant slightly outperformed those that were prepared with a grit blast surface finish and performed significantly better than those that had been scored using sand paper to achieve the desired surface finish. An additional impact investigation studied the post impact tensile strength of composite rings in a cryogenic environment. The composite rings were filament wound with several combinations of graphite and aramid fibers and were prepared with different resin systems. The rings were subjected to varying levels of Charpy impact damage and then pulled to failure in tension. It was found that the addition of elastic aramid fibers with the carbon fibers mitigates the overall impact damage and drastically improves the post-impact strength of the structure in a cryogenic environment.

  8. Computational simulation of natural convection of a molten core in lower head of a PWR pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, Camila Braga; Romero, Gabriel Alves; Jian Su, E-mail: camila@lasme.coppe.ufrj.b, E-mail: gabrielromero@lasme.coppe.ufrj.b, E-mail: sujian@lasme.coppe.ufrj.b [Universidade Federal do Rio de Janeiro (COPPE/UFRJ), RJ (Brazil). Nuclear Engineering Program

    2010-07-01

    Computational simulation of natural convection in a molten core during a hypothetical severe accident in the lower head of a typical PWR pressure vessel was performed for two-dimensional semi-circular geometry with isothermal walls. Transient turbulent natural convection heat transfer of a fluid with uniformly distributed volumetric heat generation rate was simulated by using a commercial computational fluid dynamics software ANSYS CFX 12.0. The Boussinesq model was used for the buoyancy effect generated by the internal heat source in the flow field. The two-equation k-{omega} based SST (Shear Stress Transport) turbulence model was used to mould the turbulent stresses in the Reynolds-Average Navier-Stokes equations (RANS). Two Prandtl numbers, 6:13 and 7:0, were considered. Five Rayleigh numbers were simulated for each Prandtl number used (109, 1010, 1011, 1012, and 1013). The average Nusselt numbers on the bottom surface of the semicircular cavity were in excellent agreement with Mayinger et al. (1976) correlation and only at Ra = 109 the average Nusselt number on the top flat surface was in agreement with Mayinger et al. (1976) and Kulacki and Emara (1975) correlations. (author)

  9. News from the Library: A new key reference work for the engineer: ASME's Boiler and Pressure Vessel Code at the CERN Library

    CERN Multimedia

    CERN Library

    2011-01-01

    The Library is aiming at offering a range of constantly updated reference books, to cover all areas of CERN activity. A recent addition to our collections strengthens our offer in the Engineering field.   The CERN Library now holds a copy of the complete ASME Boiler and Pressure Vessel Code, 2010 edition. This code establishes rules of safety governing the design, fabrication, and inspection of boilers and pressure vessels, and nuclear power plant components during construction. This document is considered worldwide as a reference for mechanical design and is therefore important for the CERN community. The Code published by ASME (American Society of Mechanical Engineers) is kept current by the Boiler and Pressure Committee, a volunteer group of more than 950 engineers worldwide. The Committee meets regularly to consider requests for interpretations, revision, and to develop new rules. The CERN Library receives updates and includes them in the volumes until the next edition, which is expected to ...

  10. Study of transient turbine shot without bypass in a BWR; Estudio del transitorio disparo de turbina sin bypass en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: amhed_jvq@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

    2015-09-15

    The study and analysis of operational transients are important for predicting the behavior of a system to short-terms events and the impact that would cause this transition. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could result in an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis of results of a turbine shot transient, which is not taken into operation the bypass is presented. The study is realized for a BWR of 2027 MWt, to an intermediate cycle life and using the computer code Simulate-3K a depressurization stage of the vessel is created which shows the response of other security systems and gives a coherent prediction to the event presented type. (Author)

  11. The application of ultrasonic testing technology of pressure vessels.%压力容器超声检测技术的应用

    Institute of Scientific and Technical Information of China (English)

    2013-01-01

    Pressure vessels are widely used in chemical,polymer physics and other fields,because all needs to run in under extreme conditions,such as high pressure high temperature,the pressure vessel manufacture,use process requirements are higher,nondestructive testing technology is to guarantee the safe operation of the production and one of the important methods of product quality and reliable. Pressure vessel in the manufacturing process,rely mainly on ultrasonic testing this paper mainly introduces characteristics and applicable scope of ultrasonic testing technology,in order to better work convenient.%  压力容器被广泛地应用于化工、高分子物理等领域,因均需在高压高温等极端条件下运行,对压力容器的制造、使用过程要求都较高,无损检测技术是保证生产安全运行和产品质量可靠的重要方法之一。压力容器在制造过程中,主要依靠超声进行检测本文主要介绍超声检测技术的特点及适用范围,以更好的方便开展工作。

  12. Experimental study on the pressure and pulse wave propagation in viscoelastic vessel tubes-effects of liquid viscosity and tube stiffness.

    Science.gov (United States)

    Ikenaga, Yuki; Nishi, Shohei; Komagata, Yuka; Saito, Masashi; Lagrée, Pierre-Yves; Asada, Takaaki; Matsukawa, Mami

    2013-11-01

    A pulse wave is the displacement wave which arises because of ejection of blood from the heart and reflection at vascular bed and distal point. The investigation of pressure waves leads to understanding the propagation characteristics of a pulse wave. To investigate the pulse wave behavior, an experimental study was performed using an artificial polymer tube and viscous liquid. A polyurethane tube and glycerin solution were used to simulate a blood vessel and blood, respectively. In the case of the 40 wt% glycerin solution, which corresponds to the viscosity of ordinary blood, the attenuation coefficient of a pressure wave in the tube decreased from 4.3 to 1.6 dB/m because of the tube stiffness (Young's modulus: 60 to 200 kPa). When the viscosity of liquid increased from approximately 4 to 10 mPa·s (the range of human blood viscosity) in the stiff tube, the attenuation coefficient of the pressure wave changed from 1.6 to 3.2 dB/m. The hardening of the blood vessel caused by aging and the increase of blood viscosity caused by illness possibly have opposite effects on the intravascular pressure wave. The effect of the viscosity of a liquid on the amplitude of a pressure wave was then considered using a phantom simulating human blood vessels. As a result, in the typical range of blood viscosity, the amplitude ratio of the waves obtained by the experiments with water and glycerin solution became 1:0.83. In comparison with clinical data, this value is much smaller than that seen from blood vessel hardening. Thus, it can be concluded that the blood viscosity seldom affects the attenuation of a pulse wave.

  13. Structural mechanics program: progress in 1981. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tagart, S.W. Jr.; Marston, T.U.; Nickell, R.E.; Norris, D.M.

    1982-10-01

    The goal of the EPRI Structural Mechanics Program is to improve nuclear plant reliability and availability. The program is directed toward characterization of materials, evaluation and analysis of flaws, and application and technology transfer. There are fourteen topics involving more than forty separate contracts. The largest efforts are: (1) the continuation of projects aimed at developing a valid radiation embrittlement data base for evaluating the fracture toughness of irradiated pressure vessel steels; (2) the development of weld repair procedures for reactor pressure vessels as alternatives to the half bead repair method; (3) the development of simplified design methodology for the prediction of crack initiation, stable crack growth, and instability of ductile material in the presence of flaws; and (4) evaluation of the thermal anneal remedy for reactor pressure vessel irradiation damage. The significant progress made in 1981 in this program is reviewed and the interrelationships of the projects are discussed.

  14. The Effective Convectivity Model for Simulation and Analysis of Melt Pool Heat Transfer in a Light Water Reactor Pressure Vessel Lower Head

    Energy Technology Data Exchange (ETDEWEB)

    Tran, Chi Thanh

    2009-09-15

    Reactor (BWR) lower plenum. It is found that during the accident progression, the CRGT cooling p lays a very important role in reducing the thermal loads on the reactor vessel wall. Results of the ECM and PECM simulations suggest a high potential of the CRGT cooling to be an effective measure for severe accident management in BWRs

  15. Fracture mechanics characterisation of the WWER-440 reactor pressure vessel beltline welding seam of Greifswald unit 8

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, Hans-Werner; Schuhknecht, Jan [Forschungszentrum Dresden-Rossendorf (Germany)

    2008-07-01

    WWER-440 second generation (V-213) reactor pressure vessels (RPV) were produced by IZHORA in Russia and by SKODA in the former Czechoslovakia. The surveillance Charpy-V and fracture mechanics SE(B) specimens of both producers have different orientations. The main difference is the crack extension direction which is through the RPV thickness and circumferential for ISHORA and SKODA RPV, respectively. In particular for the investigation of weld metal from multilayer submerged welding seams the crack extension direction is of importance. Depending on the crack extension direction in the specimen there are different welding beads or a uniform structure along the crack front. The specimen orientation becomes more important when the fracture toughness of the weld metal is directly determined on surveillance specimens according to the Master Curve (MC) approach as standardised in the ASTM Standard Test Method E1921. This approach was applied on weld metal of the RPV beltline welding seam of Greifswald Unit 8 RPV. Charpy size SE(B) specimens from 13 locations equally spaced over the thickness of the welding seam were tested. The specimens are in TL and TS orientation. The fracture toughness values measured on the SE(B) specimens with both orientations follow the course of the MC. Nearly all values lie within the fracture toughness curves for 5% and 95% fracture probability. There is a strong variation of the reference temperature T{sub 0} though the thickness of the welding seam, which can be explained with structural differences. The scatter is more pronounced for the TS SE(B) specimens. It can be shown that specimens with TS and TL orientation in the welding seam have a differentiating and integrating behaviour, respectively. The statistical assumptions behind the MC approach are valid for both specimen orientations even if the structure is not uniform along the crack front. By comparison crack extension, JR, curves measured on SE(B) specimens with TL and TS orientation

  16. The effect of non-metallic inclusions on the fracture toughness master curve in high copper reactor pressure vessel welds

    Science.gov (United States)

    Oh, Yong-Jun; Lee, Bong-Sang; Hong, Jun-Hwa

    2002-03-01

    The fracture toughness of two high copper reactor pressure vessel welds having low upper shelf energy was evaluated in accordance with the master curve method of ASTM E1921. The resultant data were correlated to the metallurgical factors involved in the brittle fracture initiation to provide a metallurgical-based understanding of the master curve. The tests were performed using pre-cracked Charpy V-notched specimens and the master curve was made with an average of T0 values determined at different temperatures. In all specimens, the cleavage fracture initiated at non-metallic inclusion ranging from 0.7 to 3.5 μm in diameter showing a scatter with the specimens and testing temperatures. Temperature dependency of the triggering particle size was not found. The fracture toughness ( KJC) was inversely proportional to the square root of the triggering inclusion diameter ( di) at respective temperatures. From this relationship, we determined median KJC values which correspond to the average value of triggering inclusion diameter of all tested specimens and defined them as a modified median KJC ( K'JC(med) ). The obtained K'JC(med) values showed quite smaller deviation from the master curve at different temperatures than the experimental median KJC values. This suggests that the master curve is on the premise of a constant dimension of key microstructural factor in a material regardless of the testing temperature. But the inclusion size at trigger point played an important role in the absolute position of the master curve with temperature and the consequent T0 value.

  17. Microstructural parameters governing cleavage fracture behaviors in the ductile-brittle transition region in reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Won-Jon; Lee, Bong-Sang; Oh, Yong-Jun; Huh, Moo-Young; Hong, Jun-Hwa

    2004-08-15

    The fracture behaviors in the ductile-brittle transition region of reactor pressure vessel (RPV) steels with similar chemical compositions but different manufacturing processes were examined in view of cleavage fracture stress at crack-tip. The steels typically had a variation in grain size and carbide size distribution through the different manufacturing processes. Fracture toughness was evaluated by using a statistical method in accordance to the ASTM standard E1921. From the fractography of the tested specimens, it was found that fracture toughness of the steels increased with increasing distance from the crack-tip to the cleavage initiating location, namely cleavage initiation distance (CID, X{sub f}) and its statistical mean value (K{sub JC(med)}) was proportional to the cleavage fracture stress ({sigma}{sub f}) determined from finite-element (FE) calculation at cleavage initiating location. On the other hand, {sigma}{sub f} could also be calculated by applying the size of microstructural parameters, such as carbide, grain and bainite packet, into the Griffith's theory for brittle fracture. Among the parameters, the {sigma}{sub f} obtained from the mean diameter of the carbides above 1% of the total population was in good agreement with the {sigma}{sub f} value from the FE calculation for the five different steels. The results suggest that the fracture toughness of bainitic RPV steels in the transition region is mostly influenced by only some 1% of total carbides and the critical step for cleavage fracture of the RPV steels should be the propagation of this carbide size crack to the adjacent ferrite matrix.

  18. The effect of non-metallic inclusions on the fracture toughness master curve in high copper reactor pressure vessel welds

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Yong-Jun E-mail: yjoh@kaeri.re.kr; Lee, Bong-Sang; Hong, Jun-Hwa

    2002-03-01

    The fracture toughness of two high copper reactor pressure vessel welds having low upper shelf energy was evaluated in accordance with the master curve method of ASTM E1921. The resultant data were correlated to the metallurgical factors involved in the brittle fracture initiation to provide a metallurgical-based understanding of the master curve. The tests were performed using pre-cracked Charpy V-notched specimens and the master curve was made with an average of T{sub 0} values determined at different temperatures. In all specimens, the cleavage fracture initiated at non-metallic inclusion ranging from 0.7 to 3.5 {mu}m in diameter showing a scatter with the specimens and testing temperatures. Temperature dependency of the triggering particle size was not found. The fracture toughness (K{sub J{sub C}}) was inversely proportional to the square root of the triggering inclusion diameter (d{sub i}) at respective temperatures. From this relationship, we determined median K{sub J{sub C}} values which correspond to the average value of triggering inclusion diameter of all tested specimens and defined them as a modified median K{sub J{sub C}} (K{sup '}{sub J{sub C}}{sub (med)}). The obtained K{sup '}{sub J{sub C}}{sub (med)} values showed quite smaller deviation from the master curve at different temperatures than the experimental median K{sub J{sub C}} values. This suggests that the master curve is on the premise of a constant dimension of key microstructural factor in a material regardless of the testing temperature. But the inclusion size at trigger point played an important role in the absolute position of the master curve with temperature and the consequent T{sub 0} value.

  19. Assessment of segregation kinetics in water-moderated reactors pressure vessel steels under long-term operation

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Lavrukhina, Z. V.; Saltykov, M. A.; Fedotova, S. V.; Khodan, A. N.

    2016-08-01

    In reactor pressure vessel (RPV) bcc-lattice steels temper embrittlement is developed under the influence of both operating temperature of ∼300 °C and neutron irradiation. Segregation processes in the grain boundaries (GB) begin to play a special role in the assessment of the safe operation of the RPV in case of its lifetime extension up to 60 years or more. The most reliable information on the RPV material condition can be obtained by investigating the surveillance specimens (SS) that are exposed to operational factors simultaneously with the RPV itself. In this paper the GB composition in the specimens with different thermal exposure time at the RPV operating temperature as well as irradiated by fast neutrons (E ≥ 0.5 MeV) to different fluences (20-71)·1022 m-2 was studied by means of Auger electron spectroscopy (AES) including both impurity and main alloying elements content. The data obtained allowed to trace the trend of the operating temperature and radiation-stimulated diffusion influence on the overall segregants level in GB. The revealed differences in the concentration levels of GB segregants in different steels, are due to the different chemical composition of the steels and also due to different grain boundary segregation levels in initial (unexposed) state. The data were used to estimate the RPV steels working capacity for 60 years. The estimation was carried out using both the well-known Langmuir-McLean model and the one specially developed for RPV steels, which takes into account the structure and phase composition of VVER-1000 RPV steels, as well as the long-term influence of operational factors.

  20. Nanostructure evolution under irradiation of Fe(C)MnNi model alloys for reactor pressure vessel steels

    Science.gov (United States)

    Chiapetto, M.; Becquart, C. S.; Domain, C.; Malerba, L.

    2015-06-01

    Radiation-induced embrittlement of bainitic steels is one of the most important lifetime limiting factors of existing nuclear light water reactor pressure vessels. The primary mechanism of embrittlement is the obstruction of dislocation motion produced by nanometric defect structures that develop in the bulk of the material due to irradiation. The development of models that describe, based on physical mechanisms, the nanostructural changes in these types of materials due to neutron irradiation are expected to help to better understand which features are mainly responsible for embrittlement. The chemical elements that are thought to influence most the response under irradiation of low-Cu RPV steels, especially at high fluence, are Ni and Mn, hence there is an interest in modelling the nanostructure evolution in irradiated FeMnNi alloys. As a first step in this direction, we developed sets of parameters for object kinetic Monte Carlo (OKMC) simulations that allow this to be done, under simplifying assumptions, using a "grey alloy" approach that extends the already existing OKMC model for neutron irradiated Fe-C binary alloys [1]. Our model proved to be able to describe the trend in the buildup of irradiation defect populations at the operational temperature of LWR (∼300 °C), in terms of both density and size distribution of the defect cluster populations, in FeMnNi model alloys as compared to Fe-C. In particular, the reduction of the mobility of point-defect clusters as a consequence of the presence of solutes proves to be key to explain the experimentally observed disappearance of detectable point-defect clusters with increasing solute content.

  1. Short-Term Blood Pressure Variability Relates to the Presence of Subclinical Brain Small Vessel Disease in Primary Hypertension.

    Science.gov (United States)

    Filomena, Josefina; Riba-Llena, Iolanda; Vinyoles, Ernest; Tovar, José L; Mundet, Xavier; Castañé, Xavier; Vilar, Andrea; López-Rueda, Antonio; Jiménez-Baladó, Joan; Cartanyà, Anna; Montaner, Joan; Delgado, Pilar

    2015-09-01

    Blood pressure (BP) variability is associated with stroke risk, but less is known about subclinical cerebral small vessel disease (CSVD). We aimed to determine whether CSVD relates to short-term BP variability independently of BP levels and also, whether they improve CSVD discrimination beyond clinical variables and office BP levels. This was a cohort study on asymptomatic hypertensives who underwent brain magnetic resonance imaging and 24-hour ambulatory BP monitoring. Office and average 24-hour, daytime and nighttime BP levels, and several metrics of BP variability (SD, weighted SD, coefficient of variation, and average real variability [ARV]) were calculated. Definition of CSVD was based on the presence of lacunar infarcts and white matter hyperintensity grades. Multivariate analysis and integrated discrimination improvement were performed to assess whether BP variability and levels were independently associated with CSVD and improved its discrimination. Four hundred eighty-seven individuals participated (median age, 64; 47% women). CSVD was identified in 18.9%, related to age, male sex, diabetes mellitus, use of treatment, ambulatory BP monitoring-defined BP levels, and ARV of systolic BP at any period. The highest prevalence (33.7%) was found in subjects with both 24-hour BP levels and ARV elevated. BP levels at any period and ARV (24 hours and nocturnal) emerged as independent predictors of CSVD, and discrimination was incrementally improved although not to a clinically significant extent (integrated discrimination improvement, 5.31%, 5.17% to 5.4%). Ambulatory BP monitoring-defined BP levels and ARV of systolic BP relate to subclinical CSVD in hypertensive individuals.

  2. Behaviour of the reactivity for BWR fuel cells; Comportamiento de la reactividad para celdas de combustible BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, J. A.; Alonso, G.; Delfin, A.; Vargas, S. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: galonso@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico)

    2011-11-15

    In this work the behaviour of the reactivity of a fuel assembly type BWR was studied, the objective is to obtain some expressions that consider the average enrichment of U-235 and the gadolinium concentration like a function of the fuel cells burnt. Also, the applicability of the lineal reactivity model was analyzed for fuel cells type BWR. The analysis was carried out with the CASMO-4 code. (Author)

  3. Cyclic Crack Growth Testing of an A.O. Smith Multilayer Pressure Vessel with Modal Acoustic Emission Monitoring and Data Assessment

    Science.gov (United States)

    Ziola, Steven M.

    2014-01-01

    Digital Wave Corp. (DWC) was retained by Jacobs ATOM at NASA Ames Research Center to perform cyclic pressure crack growth sensitivity testing on a multilayer pressure vessel instrumented with DWC's Modal Acoustic Emission (MAE) system, with captured wave analysis to be performed using DWCs WaveExplorerTM software, which has been used at Ames since 2001. The objectives were to document the ability to detect and characterize a known growing crack in such a vessel using only MAE, to establish the sensitivity of the equipment vs. crack size and / or relevance in a realistic field environment, and to obtain fracture toughness materials properties in follow up testing to enable accurate crack growth analysis. This report contains the results of the testing.

  4. TRACE Model for Simulation of Anticipated Transients Without Scram in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Cheng L. Y.; Baek J.; Cuadra,A.; Aronson, A.; Diamond, D.; Yarsky, P.

    2013-11-10

    A TRACE model has been developed for using theTRACE/PARCS computational package [1, 2] to simulate anticipated transients without scram (ATWS) events in a boiling water reactor (BWR). The model represents a BWR/5 housed in a Mark II containment. The reactor and the balance of plant systems are modeled in sufficient detail to enable the evaluation of plant responses and theeffectiveness of automatic and operator actions tomitigate this beyond design basis accident.The TRACE model implements features thatfacilitate the simulation of ATWS events initiated by turbine trip and closure of the main steam isolation valves (MSIV). It also incorporates control logic to initiate actions to mitigate the ATWS events, such as water levelcontrol, emergency depressurization, and injection of boron via the standby liquid control system (SLCS). Two different approaches have been used to model boron mixing in the lower plenum of the reactor vessel: modulate coolant flow in the lower plenum by a flow valve, and use control logic to modular.

  5. In-vessel coolability and steam explosion in Nordic BWRs

    Energy Technology Data Exchange (ETDEWEB)

    Ma, W.; Hansson, R.; Li, L.; Kudinov, P.; Cadinu, F.; Tran, C-.T. (Royal Institute of Technology (KTH), Stockholm (Sweden))

    2010-05-15

    The INCOSE project is to reduce the uncertainty in quantification of steam explosion risk and in-vessel coolability in Nordic BWR plants with the cavity flooding as a severe accident management (SAM) measure. During 2009 substantial advances and new insights into physical mechanisms were gained for studies of: (i) in-vessel corium coolability - development of the methodologies to assess the efficiency of the control rod guide tube (CRGT) cooling as a potential SAM measure; (ii) debris bed coolability - characterization of the effective particle diameter of multi-size particles and qualification of friction law for two-phase flow in the beds packed with multi-size particles; and (iii) steam explosion - investigation of the effect of binary oxides mixture's properties on steam explosion. An approach for coupling of ECM/PECM models with RELAP5 was developed to enhance predictive fidelity for melt pool heat transfer. MELCOR was employed to examine the CRGT cooling efficiency by considering an entire accident scenario, and the simulation results show that the nominal flowrate (approx10kg/s) of CRGT cooling is sufficient to maintain the integrity of the vessel in a BWR of 3900 MWth, if the water injection is activated no later than 1 hour after scram. The POMECO-FL experimental data suggest that for a particulate bed packed with multi-size particles, the effective particle diameter can be represented by the area mean diameter of the particles, while at high velocity (Re>7) the effective particle diameter is closer to the length mean diameter. The pressure drop of two-phase flow through the particulate bed can be predicted by Reed's model. The steam explosion experiments performed at high melt superheat (>200oC) using oxidic mixture of WO3-CaO didn't detect an apparent difference in steam explosion energetics and preconditioning between the eutectic and noneutectic melts. This points out that the next step of MISTEE experiment will be conducted at lower

  6. 核电压力容器焊接过程中的生产管理%Production Control to Welding of Nuclear Reactor Pressure Vessels

    Institute of Scientific and Technical Information of China (English)

    宋桂艳

    2014-01-01

    The article describes the welding method for nuclear reactor pressure vessels and the key points of plan control and welding production control and process control in welding operation.%概述核电压力容器焊接的特点,指出焊接生产过程中计划管理、焊接生产过程管理和过程控制的重点内容。

  7. Analytically predicted versus measured response of a free-standing steel containment vessel subjected to safety-relief valve discharge loads

    Energy Technology Data Exchange (ETDEWEB)

    Good, H.; Lewis, M.; Fitch, J.; Mattson, R.

    1987-06-01

    Following the actuation of safety-relief valves in BWR nuclear power plants, first water then air and steam are cleared from the discharge lines through quencher devices into a suppression pool. This clearing results in water spike, air bubble, and condensation pressure loads applied to structures in the pool, and the surrounding containment vessel. The Leibstadt Nuclear Power Plant has the only fre-standing steel Mark III containment vessel in the world. All other steel Mark III containment vessels have concrete backing in the suppression pool region, which dampens clearing load responses. As such, it is of interest to note how this steel vessel responds to discharge pressures, and compare these responses to analytically predicted results. The purpose of this paper is to compare the analytical results used to design the steel containment vessel with the responses measured during in-plant testing. The analytical methods considered the effects of fluid-structure interaction. The test program included initial and consecutive actuations of a single valve, and initial actuation of multiple (four) valves. The conclusion of the comparison is that, in general there are large conservatisms in the analytical predictions versus measured responses.

  8. 高温、高压、临氢在用压力容器 的氢腐蚀检验%HYDROGEN ATTACK INSPECTION ON HIGH PRESSURE VESSEL IN SERVICE UNDER HIGH TEMPERATURE,HIGH PRESSURE AND HYDROGEN ENVIROMENT

    Institute of Scientific and Technical Information of China (English)

    乔学福

    2001-01-01

    The material property of pressure vessel in high temperature,high pressure and hydrogen service can be clanged chaused by hydrogen attack.In combination with the main material inspection of 123-c heat exchanger in synthetic ammonia unit put into production for more than 20 years,the key paints and mehtod of inspecton on pressure vessel in service under high temperature,high pressure and hydrogen serice are mentioned,the mechanisum of hydrogen attack is analyzed.The determination of safety grade for hydrogen attacked pressure vessel in serice is given by the author with personal viewpoint.%高温、高压、临氢压力容器,由于氢的腐蚀其材质的性能会发生变化。作者结合对投用20多年的合成氨装置123—C换热器主体材质的检验,阐述了高温、高压、临氢在用压力容器的检验要点和方法,分析了氢腐蚀产生的机理,对已发生了氢腐蚀的在用压力容器的安全等级判定提出了个人的看法。

  9. Design of a redundant meteorological station for a BWR reactor; Diseno de una estacion meteorologica redundante para un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, R.; Celis del Angel, L.; Bucio, F.; Rivero, T.; Palacios, J. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: ramses@nuclear.inin.mx

    2008-07-01

    In this work the design of a meteorological station for a reactor type BWR is proposed. Two independent channels of data acquisition that allow him to have a bigger readiness is exposed. It is incorporate sensors without mobile parts to measure speed, wind direction and pluvial precipitation. It also counts, with sensors of global solar radiation, net radiation, barometric pressure, relative humidity and ambient temperature; with them they are possible to be calculated, moreover, other variables as temperature differential, dew point and atmospheric stability. The sensors are placed on a tower to different heights and send their information (each second) to a local registration system, the one which in turn, it remits the data to the monitoring office so that a computer is linked with the system, display and management the information in real time and automatic way. The redundant structure allows that in the event of maintenance the data acquisition is not interrupted, even if the information is transferred to another place. In all the station sections it is used protocols of standard communication to allow that a great quantity of devices can be connected without major problem. The above-mentioned would allow to the operators in the control room to have reliable information during the whole time of the reactor operation. (Author)

  10. Fracture mechanics characterisation of the WWER-440 reactor pressure vessel beltline welding seam of Greifswald unit 8

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, Hans-Werner; Schuhknecht, Jan [Forschungszentrum Dresden-Rossendorf, Dresden (Germany)

    2009-07-01

    Russian type WWER reactors are operated in Russia and many other European countries like Finland, Czech Republic, Slovak Republic, Hungary, Bulgaria and Ukraine. Surveillance specimen programmes for the inspection of the aging of the reactor pressure vessel (RPV) materials were implemented for the second generation of WWER-440/V-213 reactors. The test results and the RPV integrity assessment has been evaluated according to national codes based on the Russian code PNAE G-7-002-86 ''Strength Calculation Norms for Nuclear Power Plant Equipment and Piping'' [1]. This is an indirect and correlative approach of determining the fracture toughness of the RPV steels in the initial and irradiated condition. The Master Curve (MC) approach as adopted in the test procedure ASTM E1921 [2] for assessing the fracture toughness of sampled irradiated materials has been gaining acceptance throughout the world [3]. The MC approach is more naturally suited to probabilistic analyses because it defines both a mean transition toughness value and a distribution around that value. It contains the assumptions of macroscopically homogenous material with uniform distribution of crack initiating defects along the crack front. In contrast to present indirect and correlative approach the specimen orientation and especially the crack extension direction in multilayer weld metal becomes more important for the direct measurement of the fracture toughness with Charpy size SE(B) specimens. The orientation of the Charpy- and SE(B) specimens is different for RPVs manufactured in Russia and by the SKODA company in the former Czechoslovakia [4,5]. Particularly with regard to weld metal it can be expected that the parameters of fracture toughness measured with Charpy-V or SE(B) specimens are strongly influenced by the specimen orientation. It raises the question whether the MC approach is also applicable when the structure varies along the crack front which is happen in TL oriented SE

  11. 1-Dimensional simulation of thermal annealing in a commercial nuclear power plant reactor pressure vessel wall section

    Energy Technology Data Exchange (ETDEWEB)

    Nakos, J.T.; Rosinski, S.T.; Acton, R.U.

    1994-11-01

    The objective of this work was to provide experimental heat transfer boundary condition and reactor pressure vessel (RPV) section thermal response data that can be used to benchmark computer codes that simulate thermal annealing of RPVS. This specific protect was designed to provide the Electric Power Research Institute (EPRI) with experimental data that could be used to support the development of a thermal annealing model. A secondary benefit is to provide additional experimental data (e.g., thermal response of concrete reactor cavity wall) that could be of use in an annealing demonstration project. The setup comprised a heater assembly, a 1.2 in {times} 1.2 m {times} 17.1 cm thick [4 ft {times} 4 ft {times} 6.75 in] section of an RPV (A533B ferritic steel with stainless steel cladding), a mockup of the {open_quotes}mirror{close_quotes} insulation between the RPV and the concrete reactor cavity wall, and a 25.4 cm [10 in] thick concrete wall, 2.1 in {times} 2.1 in [10 ft {times} 10 ft] square. Experiments were performed at temperature heat-up/cooldown rates of 7, 14, and 28{degrees}C/hr [12.5, 25, and 50{degrees}F/hr] as measured on the heated face. A peak temperature of 454{degrees}C [850{degrees}F] was maintained on the heated face until the concrete wall temperature reached equilibrium. Results are most representative of those RPV locations where the heat transfer would be 1-dimensional. Temperature was measured at multiple locations on the heated and unheated faces of the RPV section and the concrete wall. Incident heat flux was measured on the heated face, and absorbed heat flux estimates were generated from temperature measurements and an inverse heat conduction code. Through-wall temperature differences, concrete wall temperature response, heat flux absorbed into the RPV surface and incident on the surface are presented. All of these data are useful to modelers developing codes to simulate RPV annealing.

  12. Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (CAPSULE 2) of Ulchin Nuclear Power Plant Unit 3

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

    2006-12-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejon after the capsule was transported from Ulchin site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Ulchin unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.674E 18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.920 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.913E 18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 9.249E 18 and 1.834E 19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

  13. Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (Capsule 2) of Ulchin Nuclear Power Plant Unit 4

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

    2007-04-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejon after the capsule was transported from Ulchin site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Ulchin Unit 4 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.306E+18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.918 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.615E+18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 8.478E+18 and 1.673E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin Unit 4 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

  14. 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...

  15. BWR ASSEMBLY SOURCE TERMS FOR WASTE PACKAGE DESIGN

    Energy Technology Data Exchange (ETDEWEB)

    T.L. Lotz

    1997-02-15

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide boiling water reactor (BWR) assembly radiation source term data for use during Waste Package (WP) design. The BWR assembly radiation source terms are to be used for evaluation of radiolysis effects at the WP surface, and for personnel shielding requirements during assembly or WP handling operations. The objectives of this evaluation are to generate BWR assembly radiation source terms that bound selected groupings of BWR assemblies, with regard to assembly average burnup and cooling time, which comprise the anticipated MGDS BWR commercial spent nuclear fuel (SNF) waste stream. The source term data is to be provided in a form which can easily be utilized in subsequent shielding/radiation dose calculations. Since these calculations may also be used for Total System Performance Assessment (TSPA), with appropriate justification provided by TSPA, or radionuclide release rate analysis, the grams of each element and additional cooling times out to 25 years will also be calculated and the data included in the output files.

  16. Transient and stability analysis of a BWR core with thorium-uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779 Col. Narvarte, 03020 Mexico, DF (Mexico); Espinosa-Paredes, Gilberto [Division de Ciencias Basicas e Ingenieria, Universidad Autonoma Metropolitana, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico, DF (Mexico)], E-mail: gepe@xanum.uam.mx; Francois, Juan-Luis [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, 62550 Jiutepec Mor. (Mexico)

    2008-08-15

    The kinetic response of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material, in an integrated blanket-seed assembly, is presented in this work. Additionally an in-house code was developed to evaluate this core under steady state and transient conditions including a stability analysis. The code has two modules: (a) the time domain module for transient analysis and (b) the frequency domain module for stability analysis. The thermal-hydraulic process is modeled by a set of five equations, considering no homogeneous flow with drift-flux approximation and non-equilibrium thermodynamic. The neutronic process is calculated with a point kinetics model. Typical BWR reactivity effects are considered: void fraction, fuel temperature, moderator temperature and control rod density. Collapsed parameters were included in the code to represent the core using an average fuel channel. For the stability analysis, in the frequency domain, the transfer function is determined by applying Laplace-transforming to the calculated pressure drop perturbations in each of the considered regions where a constant total pressure drop was considered. The transfer function was used to study the system response in the frequency domain when an inlet flow perturbation is applied. The results show that the neutronic behavior of the core with thorium uranium fuel is similar to a UO{sub 2} core, even during transient conditions. The stability and transient analysis show that the thorium-uranium fuel can be operated safely in current BWRs.

  17. Program to develop acoustic emission-flaw relationship for inservice monitoring of nuclear pressure vessels. Annual report, July 1, 1976 - October 1, 1977

    Energy Technology Data Exchange (ETDEWEB)

    Hutton, P.H.; Kurtz, R.J.; Schwenk, E.B.; Pavloff, C.

    1978-06-01

    Laboratory mechanical tests were conducted to evaluate AE during uniaxial tensile, fracture and fatigue crack growth in A533B pressure vessel steel. The A533B steel included two heats of Class 1, one heat of Class 2 and a weldment made for the Heavy Section Steel Technology (HSST) Program. Specimen types included uniaxial tensile specimens, size 2 compact tension specimens for fatigue crack growth and fracture tests, and a single-edge notch specimen also for fatigue crack growth through material that was uniformly strained 3% prior to fatigue testing. In addition, AE monitoring was conducted on the HSST V-7B 6-inch thick pressure vessel test. AE data were partitioned into four ranges of signal amplitude and rise time. All the AE data were analyzed, with respect to mechanical behavior of A533B steel. Linear elastic fracture mechanics analysis methods were used to relate AE parameters to fracture and fatigue crack growth parameters. AE data from the V-7B vessel test were correlated with stress intensity factor and crack opening displacement. AE data from the fatigue crack growth tests were investigated using models based on fatigue crack growth rate, fatigue crack area and theoretical crack tip plastic zone size.

  18. Pressure and temperature profile data collected by the NOAA vessel Bay Hydrographer during survey operations along the NE US coast, 03 February 2005 to 21 November 2005 (NODC Accession 0002670)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Pressure and temperature profile data were collected using CTD casts from the NOAA Survey Vessel BAY HYDROGRAPHER. Data were collected in the Chesapeake Bay from...

  19. 60th Anniversary of electricity production from light water reactors: Historical review of the contribution of materials science to the safety of the pressure vessel

    Science.gov (United States)

    van Duysen, J. C.; Meric de Bellefon, G.

    2017-02-01

    The first light water nuclear reactor dedicated to electricity production was commissioned in Shippingport, Pennsylvania in the United States in 1957. Sixty years after the event, it is clear that this type of reactor will be a major source of electricity and one of the key solutions to limit climate change in the 21st century. This article pays homage to the teams that contributed to this achievement by their involvement in research and development and their determination to push back the frontiers of knowledge. Via a few examples of scientific or technological milestones, it describes the evolution of ideas, models, and techniques during the last 60 years, and gives the current state-of-the-art in areas related to the safety of the reactor pressure vessel. Among other topics, it focuses on vessel manufacturing, steel fracture mechanics analysis, and understanding of irradiation-induced damage.

  20. Analysis of dosimetry from the H.B. Robinson unit 2 pressure vessel benchmark using RAPTOR-M3G and ALPAN

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, G.A. [Westinghouse Electric Company, LLC, 1000 Westinghouse Dr., Cranberry Township, PA 16066 (United States)

    2011-07-01

    Document available in abstract form only, full text of document follows: The dosimetry from the H. B. Robinson Unit 2 Pressure Vessel Benchmark is analyzed with a suite of Westinghouse-developed codes and data libraries. The radiation transport from the reactor core to the surveillance capsule and ex-vessel locations is performed by RAPTOR-M3G, a parallel deterministic radiation transport code that calculates high-resolution neutron flux information in three dimensions. The cross-section library used in this analysis is the ALPAN library, an Evaluated Nuclear Data File (ENDF)/B-VII.0-based library designed for reactor dosimetry and fluence analysis applications. Dosimetry is evaluated with the industry-standard SNLRML reactor dosimetry cross-section data library. (authors)

  1. Confinement Vessel Dynamic Analysis

    Energy Technology Data Exchange (ETDEWEB)

    R. Robert Stevens; Stephen P. Rojas

    1999-08-01

    A series of hydrodynamic and structural analyses of a spherical confinement vessel has been performed. The analyses used a hydrodynamic code to estimate the dynamic blast pressures at the vessel's internal surfaces caused by the detonation of a mass of high explosive, then used those blast pressures as applied loads in an explicit finite element model to simulate the vessel's structural response. Numerous load cases were considered. Particular attention was paid to the bolted port connections and the O-ring pressure seals. The analysis methods and results are discussed, and comparisons to experimental results are made.

  2. An overview of the BWR ECCS strainer blockage issues

    Energy Technology Data Exchange (ETDEWEB)

    Serkiz, A.W.; Marshall, M.L. Jr.; Elliott, R. [Nuclear Regulatory Commission, Washington, DC (United States)

    1996-03-01

    This Paper provides a brief overview of actions taken in the mid 1980s to resolve Unresolved Safety Issue (USI) A-43, {open_quotes}Containment Emergency Sump Performance,{close_quotes} and their relationship to the BWR strainer blockage issue; the importance of insights gained from the Barseback-2 (a Swedish BWR) incident in 1992 and from ECCS strainer testing and inspections at the Perry nuclear power plant in 1992 and 1993; an analysis of an US BWR/4 with a Mark I containment; an international community sharing of knowledge relevant to ECCS strainer blockage, additional experimental programs; and identification of actions needed to resolve the strainer blockage issue and the status of such efforts.

  3. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - models and correlations

    Energy Technology Data Exchange (ETDEWEB)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code`s capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.

  4. Estimate of coolant flow in assemblies of a natural circulation BWR applying and equivalent electric model; Estimacion del flujo de refrigerante en los ensambles de un BWR de circulacion natural aplicando un modelo electrico equivalente

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, DEPFI, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico); Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)], e-mail: julfi_jg@yahoo.com.mx

    2009-10-15

    The present work exposes the design and implementation of an advanced controller that it allows to estimate the coolant flow in fuel assemblies of a natural circulation BWR in real time. the complete development of this study is part of a doctoral project in course. In this work the construction of optimal controller is shown that allows to estimate the coolant flows in reactor and its operation applied to an equivalent electric model to natural circulation ESBWR. The controller design that allows the completely automatic starter of natural circulation reactor, required of a variables estimator not meter directly of nuclear power plant and use of local distributions estimates of coolant flow, (this controller type at the moment is utilized in the A BWR and several BWR in operation in Japan). The construction of estimator controller is mathematically based in the theory referring to Kalman filter, whose algorithm provides an advanced control of system. To prove the estimator operation was developed a simplified model that reproduces the basic dynamic of coolant flowing in the ESBWR, a practice way and very interesting of representing this phenomenon is by means the use of an equivalent electric model, which was developed starting from analogies that there is among the relation that keep the pressure differences with the mass flow and differences of electric potential with electric current. A detailed analysis of equivalence among models will be presented in a later article. (Author)

  5. Dust explosions in spherical vessels: prediction of the pressure evolution and determination of the burning velocity and flame thickness

    NARCIS (Netherlands)

    Dahoe, A.E.; Zevenbergen, J.F.; Verheijen, P.J.T.; Lemkowitz, S.M.; Scarlett, B.

    1996-01-01

    A well known limitation of the ’cube-root-law’ is that it becomes invalid when the flame thickness is significant with respect to the vessel radius. In the literature flame thicknesses in dust-air mixtures ranging from 15 to 80 centimeters have been reported [1], which exceed the radii of the 20-1it

  6. Effect of Initial Moisture Content on the in-Vessel Composting Under Air Pressure of Organic Fraction of MunicipalSolid Waste in Morocco

    Directory of Open Access Journals (Sweden)

    Abdelhadi Makan

    2013-01-01

    Full Text Available This study aimed to evaluate the effect of initial moisture content on the in-vessel composting under air pressure of organic fraction of municipal solid waste in Morocco in terms of internal temperature, produced gases quantity, organic matter conversion rate, and the quality of the final composts.For this purpose, in-vessel bioreactor was designed and used to evaluate both appropriate initial air pressure and appropriate initial moisture content for the composting process. Moreover, 5 experiments were carried out within initial moisture content of 55%, 65%, 70%, 75% and 85%. The initial air pressure and the initial moisture content of the mixture showed a significant effect on the aerobic composting. The experimental results demonstrated that for composting organic waste, relatively high moisture contents are better at achieving higher temperatures and retaining them for longer times.This study suggested that an initial moisture content of around 75%, under 0.6 bar, can be considered as being suitable for efficient composting of organic fraction of municipal solid waste. These last conditions, allowed maximum value of temperature and final composting product with good physicochemical properties as well as higher organic matter degradation and higher gas production. Moreover, final compost obtained showed good maturity levels and can be used for agricultural applications.

  7. The mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels: The case of Fe-Cu model alloys

    Science.gov (United States)

    Subbotin, A. V.; Panyukov, S. V.

    2016-08-01

    Mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels is proposed and developed in case of Fe-Cu model alloys. The suggested solute-drag mechanism is analogous to the well-known zone-refining process. We show that the obtained results are in good agreement with available experimental data on the parameters of clusters enriched with the alloying elements. Our model explains why the formation of solute-enriched clusters does not happen in austenitic stainless steels with fcc lattice structure. It also allows to quantify the method of evaluation of neutron irradiation dose for the process of RPV steels hardening.

  8. The mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels: the case of Fe-Cu model alloys

    CERN Document Server

    Subbotina, A V

    2016-01-01

    Mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels is proposed and developed in case of Fe-Cu model alloys. We show that the obtained results are in a good agreement with available experimental data on the parameters of clusters enriched with the alloying elements. The suggested solute-drag mechanism is analogous to the well-known zone-refining process. Our model explains why the formation of solute-enriched clusters does not happen in austenitic stainless steels with fcc lattice structure. It also allows to quantify the method of evaluation of neutron irradiation dose for the process of RPV steels hardening.

  9. Evaluation of defects induced by neutron radiation in reactor pressure vessels steels; Evaluacion de los defectos inducidos por la radiacion neutronica en los aceros de vasijas

    Energy Technology Data Exchange (ETDEWEB)

    Lopez Jimenez, J.

    1978-07-01

    We have developed a method for calculating the production of neutron induced defects (depleted zone and crowdions) in ferritic pressure vessel steels for different neutron spectra. They have been analysed both the recoil primary atoms produced by elastic and inelastic collisions with fast neutrons and the ones produced by gamma-ray emission by thermal neutron absorption. Theoretical modelling of increasing in the ductile-brittle transition temperature of ferritic steels has been correlated with experimental data at irradiation temperature up to 400 degree centigree (Author) 15 refs.

  10. Crack growth behaviour of low alloy steels for pressure boundary components under transient light water reactor operating conditions (CASTOC)

    Energy Technology Data Exchange (ETDEWEB)

    Foehl, J.; Weissenberg, T. [Materialpruefungsanstalt, Univ. Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas (CIEMAT) (Spain); Ernestova, M.; Zamboch, M. [Nuclear Research Inst. (NRI) (Czech Republic); Seifert, H.P.; Ritter, S. [Paul Scherrer Inst. (PSI) (Switzerland); Roth, A.; Devrient, B. [Framatome ANP GmbH (F ANP) (Germany); Ehrnsten, U. [Technical Research Centre of Finland (VTT) (Finland)

    2004-07-01

    water at stress intensity factors above the limit for linear elastic fracture mechanics. There is evidence that the prediction curves of the ASME Boiler and Pressure Vessel Code Section XI, Appendix A are not conservative for some relevant cases with regard to crack growth rates under cyclic load even in oxygenated high purity BWR water. The CASTOC results have provided an important contribution to the understanding of crack growth behavior on the one hand as a function of time and on the other hand as a consequence of the number and height of loading events. This is an important key for the evaluation of transient events, which may occur in a plant during service. (orig.)

  11. Revisiting the reactor pressure vessel for long-time operation; Revisitando la vasija a presion del reactor para largos tiempos de operacion

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

    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 LONGIFFE 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)

  12. Safety/relief valve quencher loads: evaluation for BWR Mark II and III containments

    Energy Technology Data Exchange (ETDEWEB)

    Su, T.M.

    1982-10-01

    Boiling water reactor (BWR) plants are equipped with safety/relief valves (SRVs) to protect the reactor from overpressurization. Plant operational transients, such as turbine trips, will actuate the SRV. Once the SRV opens, the air column within the partially submerged discharge line is compressed by the high-pressure steam released from the reactor. The compressed air discharged into the suppression pool produces high-pressure bubbles. Oscillatory expansion and contraction of these bubbles create hydrodynamic loads on the containment structures, piping, and equipment inside containment. This report presents the results of the staff's evaluation of SRV loads. The evaluation, however, is limited to the quencher devices used in Mark II and III containments. With respect to Mark I containments, the SRV acceptance criteria are presented in NUREG-0661 issued July 1980. The staff acceptance criteria for SRV loads for Mark II and III containments are presented in this report.

  13. Mechanical analysis and reasonable design for Ti-Al alloy liner wound with carbon fiber resin composite high pressure vessel

    Institute of Scientific and Technical Information of China (English)

    Chuan-xiang ZHENG; Fan YANG; Ai-shi ZHU

    2009-01-01

    To consider the internal pressure loaded by both the cylindrical Ti-AI alloy liner and the carbon fiber resin composite (CFRC) wound layers, two models are built. The first one is a cylinder loaded with the internal pressure in the hoop direction only. In this model, the total hoop direction load is distributed over all layers under the internal pressure. The second one is a cylinder loaded with the internal pressure in the axial direction only. In this model, the total axial load is distributed over all cylinders under the internal pressure. Taking the boundary conditions of the continuous displacement between layers into account, a group of equations are built. From these equations, we get the solutions of stresses in both hoop direction and axial direction loaded by every layer under internal pressures. After the stresses are obtained, a reasonable design can be done. An example is given in the final section of this study.

  14. The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Yonggwang nuclear power plant unit 2

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-02-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.762E+18, 1.5391E+19, 3.5119E+19, and 4.2610E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.899 for the 1st through 4th testing and the calculational uncertainty, 12.3% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.357E+19 n/cm{sup 2} based on the end of 11th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.525E+19, 4.337E+19, 5.148E+19 and 5.960E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 42 tabs. (Author)

  15. Final report for the 5th surveillance test of the reactor pressure vessel material (capsule Y) of Yonggwang Nuclear Power Plant unit 2

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sam Lai; Kim, ByoungChul; Chang, Kee Ok (and others)

    2006-02-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejeon after the capsule was transported from Yonggwang site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X, W and Y are 5.777E+18, 1.5371E+19, 3.7634E+19, 4.3045E+19, and 4.8662E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.953 for the 1st through 5th testing and the calculational uncertainty,7.2% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.659E+19n/cm{sup 2} based on the end of 13th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 48, 56 and 64EFPY would reach 3.625E+19, 5.293E+19, 6.127E+19 and 6.960E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the Pressurized Thermal Shock(PTS) during the operation until design life.

  16. The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Younggwang nuclear power plant unit1

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-08-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Yonggwang site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 1 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.555E+18, 1.662E+19, 3.358E+19, and 4.521E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.859 for the 1st through 4th testing and the calculational uncertainty, 11.80% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.551E+19 n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.929E+19, 4.880E+19, 5.831E+19 and 6.782E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 1 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 4 refs., 41 figs., 35 tabs. (Author)

  17. BWR stability using a reducing dynamical model; Estabilidad de un BWR con un modelo dinamico reducido

    Energy Technology Data Exchange (ETDEWEB)

    Ballestrin Bolea, J. M.; Blazquez Martinez, J. B.

    1990-07-01

    BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical structure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations is non-linear. Simple parametric calculation of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author)

  18. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs.

  19. The Advanced BWR Nuclear Plant: Safe, economic nuclear energy

    Energy Technology Data Exchange (ETDEWEB)

    Redding, J.R. [GE Nuclear Energy, San Jose, CA (United States)

    1994-12-31

    The safety and economics of Advanced BWR Nuclear Power Plants are outlined. The topics discussed include: ABWR Programs: status in US and Japan; ABWR competitiveness: safety and economics; SBWR status; combining ABWR and SBWR: the passive ABWR; and Korean/GE partnership.

  20. Comparison of MELCOR modeling techniques and effects of vessel water injection on a low-pressure, short-term, station blackout at the Grand Gulf Nuclear Station

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, J.J.

    1995-06-01

    A fully qualified, best-estimate MELCOR deck has been prepared for the Grand Gulf Nuclear Station and has been run using MELCOR 1.8.3 (1.8 PN) for a low-pressure, short-term, station blackout severe accident. The same severe accident sequence has been run with the same MELCOR version for the same plant using the deck prepared during the NUREG-1150 study. A third run was also completed with the best-estimate deck but without the Lower Plenum Debris Bed (BH) Package to model the lower plenum. The results from the three runs have been compared, and substantial differences have been found. The timing of important events is shorter, and the calculated source terms are in most cases larger for the NUREG-1150 deck results. However, some of the source terms calculated by the NUREG-1150 deck are not conservative when compared to the best-estimate deck results. These results identified some deficiencies in the NUREG-1150 model of the Grand Gulf Nuclear Station. Injection recovery sequences have also been simulated by injecting water into the vessel after core relocation started. This marks the first use of the new BH Package of MELCOR to investigate the effects of water addition to a lower plenum debris bed. The calculated results indicate that vessel failure can be prevented by injecting water at a sufficiently early stage. No pressure spikes in the vessel were predicted during the water injection. The MELCOR code has proven to be a useful tool for severe accident management strategies.

  1. Investigation of the Impact of ENDF/B-VI Cross Sections on the H.B. Robinson-2 Pressure-Vessel Flux Prediction

    Energy Technology Data Exchange (ETDEWEB)

    Remec, I

    1999-06-01

    This report discusses the impact of the change from the SAILOR cross-section library, based on the ENDF/B-IV data, to the BUGLE-96 cross-section library, based on the ENDF/B-VI data, on the neutron flux prediction in the H. B. Robinson-2 pressure vessel, in the surveillance capsule, and in the cavity. The fast flux (E > 1 MeV) from the transport calculations with the BUGLE-96 library is {approximately}6% higher in the surveillance capsule and at the PV inner wall, and {approximately}25% higher in the reactor cavity than the flux from the transport calculations with the SAILOR library. These changes result from the combined effect of the changes in the cross sections, which cause significant increases in the calculated fluxes, and much smaller decreases in the fast fluxes due to the changes in the fission spectra. The increase in the calculated fast flux due to the changes in the cross sections only is {approximately}9% in the capsule and at the pressure vessel (PV) wall, and {approximately}30% in the cavity. The changes in the fission spectra lead to decreases in the order of {approximately}3-4% in calculated fast fluxes.

  2. Microstructure and embrittlement of VVER 440 reactor pressure vessel steels; Microstructure et fragilisation des aciers de cuve des reacteurs nucleaires VVER 440

    Energy Technology Data Exchange (ETDEWEB)

    Hennion, A

    1999-03-15

    27 VVER 440 pressurised water reactors operate in former Soviet Union and in Eastern Europe. The pressure vessel, is made of Cr-Mo-V steel. It contains a circumferential arc weld in front of the nuclear core. This weld undergoes a high neutron flux and contains large amounts of copper and phosphorus, elements well known for their embrittlement potency under irradiation. The embrittlement kinetic of the steel is accelerated, reducing the lifetime of the reactor. In order to get informations on the microstructure and mechanical properties of these steels, base metals, HAZ, and weld metals have been characterized. The high amount of phosphorus in weld metals promotes the reverse temper embrittlement that occurs during post-weld heat treatment. The radiation damage structure has been identified by small angle neutron scattering, atomic probe, and transmission electron microscopy. Nanometer-sized clusters of solute atoms, rich in copper with almost the same characteristics as in western pressure vessels steels, and an evolution of the size distribution of vanadium carbides, which are present on dislocation structure, are observed. These defects disappear during post-irradiation tempering. As in western steels, the embrittlement is due to both hardening and reduction of interphase cohesion. The radiation damage specificity of VVER steels arises from their high amount of phosphorus and from their significant density of fine vanadium carbides. (author)

  3. Analysis of a German BWR core with TRACE/PARCS using different cross section sets

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, C., E-mail: Christoph.Hartmann@kit.edu [Karlsruhe Inst. of Tech. (KIT), Inst. for Neutron Physics and Reactor Technology (INR), Eggenstein-Leopoldshafen (Germany); Westinghouse Electric Germany GmbH, Mannheim (Germany); Sanchez, V.H. [Karlsruhe Inst. of Tech. (KIT), Inst. for Neutron Physics and Reactor Technology (INR), Eggenstein-Leopoldshafen (Germany); Tietsch, W. [Westinghouse Electric Germany GmbH, Mannheim (Germany)

    2011-07-01

    'Full text:' Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools as well as appropriate cross sections (XS) libraries depending on the individual thermal hydraulic state parameters. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running and user-friendly lattice codes such as Casco and Helios. At research institutions and universities, alternative tools to the commercial ones with full access to the source code as well as moderate cost are urgently needed. The Scale system is being developed and improved for lattice physics calculations of real core loading of Light Water Reactors (LWR). It represents a promising alternative to the commercial lattice codes. At Karlsruhe Institute of Technology (Kit) a computational route based on Scale/Triton/Newt for BWR core loading is under development. The generated XS-data sets have to be transformed in PMAXS-format for use in the reactor dynamic code PARCS. This task is performed by the module GenPMAXS being developed and tested at the Michigan University. To verify the computational route, a BWR fuel assembly depletion problem was calculated by PARCS and compared to the CASMO results. Since the SCALE/TRITON XS-file does actually not contain all required neutronic data, FORTRAN routines have been developed to incorporate the missing data e.g. the yields of Iodine, Xenon and Promethium into the XS-data sets in the PMAXS-format. The comparison of the results obtained with PARCS (using the corrected PMAXS file) and CASMO for the depletion problem exhibited a good agreement. Consequently, this approach was followed for the generation of a complete XS-set for a real BWR core to be used in subsequent transient analysis. Then 3D neutronic and thermal hydraulic core model were elaborated for a TRACE/PARCS analysis. The thermal hydraulic model is based on the 3D VESSEL

  4. Effect of a Chloride Transient on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Test 2)

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.-P

    2002-11-01

    Within the CASTOC-project (5{sup t}h EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. The present report is a summary of the second test of working package (WP) 3 with a NaCl transient, performed at Paul Scherrer Institut (PSI). In the first part of the experiment, an actively growing EAC crack with a crack growth rate (CGR) in the range of the 'low-sulphur SCC line' of the GE-model was generated by periodical partial unloading (PPU) in oxygenated high-temperature, high-purity water (T = 288 {sup o}C, DO = 8 ppm). Then a chloride transient of 49 ppb Cl{sup -} was applied for {approx}40 h. After this transient, the load level was reduced and the loading conditions were changed to pure cyclic loading. Thereupon a second transient with a chloride concentration of 49 ppb was applied. In both RPV steels, the first chloride transient of 49 ppb Cl{sup -} resulted in an acceleration of the EAC crack growth by more than one order of magnitude and in fast, stationary SCC crack growth during the constant load phase of the PPU cycles at K{sub I} values < 60 MPa.m{sup 1/2}. 3 h after adding chloride to the high-purity water, the EAC CGR started to increase in the high-sulphur RPV steel during the constant load phase of a PPU cycle and after 20 h a stationary EAC CGR value in the range of the 'high-sulphur SCC curve' of the GE-model was reached. After 5 h in high-purity water, the crack growth began to slow down after a partial unloading cycle and 15 h later it reached again a stationary CGR value in the range of the 'low-sulphur SCC curve' of the GE-model. The second chloride transient did not result in an acceleration of the crack growth in both investigated specimens. This was explained by crack closure effects

  5. Ex-Vessel corium coolability and steam explosion energetics in nordic light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Ma, W.M.; Karbojian, A.; Kudinov, P.; Tran, C.T.; Hansson, C.R. [Royal Institute of Technology (KTH), (Sweden)

    2008-03-15

    This report presents advances and insights from the KTH's study on corium pool heat transfer in the BWR lower head; debris bed formation; steam explosion energetics; thermal hydraulics and coolability in bottom-fed and heterogeneous debris beds. Specifically, for analysis of heat transfer in a BWR lower plenum an advanced threedimensional simulation tool was developed and validated, using a so-called effective convectivity approach and Fluent code platform. An assessment of corium retention and coolability in the reactor pressure vessel (RPV) lower plenum by means of water supplied through the Control Rod Guide Tube (CRGT) cooling system was performed. Simulant material melt experiments were performed in an intermediate temperature range (1300-1600K) on DEFOR test facility to study formation of debris beds in high and low subcooled water pools characteristic of in-vessel and ex-vessel conditions. Results of the DEFOR-E scoping experiments and related analyses strongly suggest that porous beds formed in ex-vessel from a fragmented high-temperature debris is far from homogeneous. Calculation results of bed thermal hydraulics and dryout heat flux with a two-dimensional thermal-hydraulic code give the first basis to evaluate the extent by which macro and micro inhomogeneity can enhance the bed coolability. The development and validation of a model for two-phase natural circulation through a heated porous medium and its application to the coolability analysis of bottom-fed beds enables quantification of the significant effect of dryout heat flux enhancement (by a factor of 80-160%) due to bottom coolant injection. For a qualitative and quantitative understanding of steam explosion, the SHARP system and its image processing methodology were used to characterize the dynamics of a hot liquid (melt) drop fragmentation and the volatile liquid (coolant) vaporization. The experimental results provide a basis to suggest that the melt drop preconditioning is instrumental to

  6. Numerical Simulation of the Pressure Distribution in the Reactor Vessel Downcomer Region Fluctuated by the Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Dong Hwa; Jung, Byung Ryul; Jang, Ho Cheol; Yune, Seok Jeong; Kim, Eun Kee [KEPCO EnC, Daejeon (Korea, Republic of)

    2015-10-15

    In this study the numerical simulation of the pressure distribution in the downcomer region resulting from the pressure pulsation by the Reactor Coolant Pump (RCP) is performed using the Finite Difference Method (FDM). Simulation is carried out for the cylindrical shaped 2-dimensional model equivalent to the outer surface of the Core Support Barrel (CSB) of APR1400 and a 1/2 model is adopted based on the bilateral symmetry by the inlet nozzle. The fluid temperature is 555 .deg. F and the forcing frequencies are 120Hz, 240Hz, 360Hz and 480Hz. Simulation results of the axial pressure distributions are provided as the Root Mean Square (RMS) values at the five locations of 0°, 45°, 90°, 135° and 180° in the circumferential direction from the inlet nozzle location. In the study, the numerical simulation of pressure distributions in the downcomer region induced by the RCP was performed using FDM and the results were reviewed. The interference of the waves returned from both boundaries in the axial direction and the source of the sinusoidal wave is shown on the inlet nozzle interface pressure point. It seems that the maximum pressures result from the superposition of the waves reflected from the seating surface and the waves newly arrived from the inlet nozzle interface pressure location.

  7. Critical discharge of initially subcooled water through slits. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Amos, C N; Schrock, V E

    1983-09-01

    This report describes an experimental investigation into the critical flow of initially subcooled water through rectangular slits. The study of such flows is relevant to the prediction of leak flow rates from cracks in piping, or pressure vessels, which contain sufficient enthalpy that vaporization will occur if they are allowed to expand to the ambient pressure. Two new analytical models, which allow for the generation of a metastable liquid phase, are developed. Experimental results are compared with the predictions of both these new models and with a Fanno Homogeneous Equilibrium Model.

  8. 10 MW高温气冷堆反应堆压力容器的出厂水压试验%Hydraulic Pressure Test of Pressure Vessel of 10 MW High Temperature Gas-cooled Reactor

    Institute of Scientific and Technical Information of China (English)

    刘俊杰; 张征明; 何树延; 王金海

    2001-01-01

    The hydraulic pressure test of 10MW Hight Temperature Gas-cooled Reactorc(HTR-10) pressure vessel was successfully performed according to the requirement of the section NB-6200, ASME Ⅲ code. The test requirement, the test results and the test evaluations are described in detail. The test tension was effectively and rationally done through an hydraulic tensionor, which was developed at institue of nuclear energy technology of Tsinghua University. The strain and deformation of the HTR-10 pressure vessel were also measured.%根据ASME规范第Ⅲ卷NB-6200节的规定,对10MW高温气冷堆压力容器的水压试验要求、试验过程,试验结果及评价进行了叙述。用清华大学核能技术设计研究院研制的液压张拉机对主螺栓实施了合理及有效的张拉,对压力容器进行了应变和变形测量,取得了反应堆压力容器水压试验的圆满成功。

  9. Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL; Ade, Brian J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Mertyurek, Ugur [ORNL; Radulescu, Georgeta [ORNL

    2015-01-01

    credit at peak reactivity requires a different set of experiments than for pressurized-water reactor burnup credit analysis because of differences in actinide compositions, presence of residual gadolinium absorber, and lower fission product concentrations. A survey of available critical experiments is presented along with a sample criticality code validation and determination of undercoverage penalties for some nuclides. The validation of depleted fuel compositions at peak reactivity presents many challenges which largely result from a lack of radiochemical assay data applicable to BWR fuel in this burnup range. In addition, none of the existing low burnup measurement data include residual gadolinium measurements. An example bias and uncertainty associated with validation of actinide-only fuel compositions is presented.

  10. Effect of Loading Transients on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Tests 3 and 4)

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.-P

    2003-04-01

    Within the CASTOC-project (5{sup t}h EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. The present report is a summary of the third and fourth test of working package (WP) 3 with loading transients, performed at Paul Scherrer Institut (PSI). Two different low-alloy steels (20 MnMoNi 5 5, 0.015 wt.% S and 22 NiMoCr 3 7, 0.007 wt. %S) were investigated in oxygenated high-temperature, high-purity water (T = 240 {sup o}C, DO = 400 ppb) in a daisy chain at two different load ratios (R = 0.8 and 0.2). In the first part of the experiments, asymmetrical saw tooth loading with different rise times {delta}t{sub R} of the load and different loading frequencies were applied. Then the loading conditions were changed to an asymmetrical trapezoid waveform loading (periodical partial unloading, PPU) and the hold time {delta}t{sub H} at maximum load was varied. In the final phase of WP 3 PSI tests 3 and 4 the SCC behaviour was investigated under constant load. With decreasing loading frequency the corrosion fatigue (CF) crack advance per cycle {delta}a/{delta}N{sub EAC} of material A increased. Sustained EAC crack growth could be maintained down to low frequencies of 10{sup -5} Hz. The time-based crack growth rate (CGR) da/dt{sub EAC} decreased with decreasing frequency. In material B no effect of the loading frequency could be resolved. Up to a hold time of 1 h at maximum constant load the CGR da/dt{sub EAC} seemed to be independent of the hold time. Above hold times of 1 h the CGR decreased and dropped down to CGR values in the range or below the BWR VIP 60 SCC disposition lines. This behaviour was observed in both investigated materials. The cycle-based CGR {delta}a/{delta}N{sub EAC} remained approximately constant with increasing hold time. The

  11. 反应堆退役压力容器放射性活度估算方法%Method for Estimation of Activity in Decommissioned Nuclear Reactor Pressure Vessel

    Institute of Scientific and Technical Information of China (English)

    郭武仁; 林晓玲; 郑宁宁

    2011-01-01

    The theoretical calculation and experimental measurement methods for estimation of activity in the decommissioned nuclear reactor pressure vessel were introduced. The physical estimation model was described,and Monte Carlo compute code and 0RIGEN2 code were recommended to be employed for the calculation of the neutron flux and activity in the reactor pressure vessel. Two methods commonly used for determining the activity in the reactor pressure vessel were introduced in detail,I.e.,sampling from the reactor pressure vessel and from the irradiation tube. The neutron flux profile was established to predict the activity in the reactor pressure vessel.%介绍了反应堆退役压力容器放射性活度估算的理论计算和实验测定方法.描述了物理估算模型,推荐采用蒙特卡罗程序和ORIGEN2程序分别计算中子通量密度和放射性活度.对确定压力容器的放射性活度时经常使用的两种方法(压力容器直接取样分析和对辐照监督管取样分析)做了详细介绍.建立了推算压力容器的放射性活度中子通量密度比例曲线.

  12. Three dimensional parameterization modeling method for pressure vessel head%压力容器封头参数化三维建模方法

    Institute of Scientific and Technical Information of China (English)

    张义顺; 梁盈; 刘海波

    2011-01-01

    为了满足压力容器设计行业的需要,提高压力容器封头、简体、法兰及弯头等组件的设计效率,采用内嵌在AutoCAD内部的VBA编程工具建立UCS,编写相关函数,增加插入点功能并融合空间坐标系转换技术,利用布尔运算、旋转等方法实现了封头三维建模.利用该参数化建模方法,设计人员只需输入封头的相关数据,便可在任意视角下、任意捕捉点处得到符合要求的三维封头模型.此外,结合AutoCAD三维造型技术,配合VBA编程,可对压力容器其他组件如法兰、弯头等进行参数化建模,这对于开发设计相关行业AutoCAD三维插件具有一定的指导意义.%To meet the need of pressure vessel design industry and raise the design efficiency of such components as pressure vessel head, cylinder, flange and elbow, UCS was established with VBA programming tool embedded in AutoCAD software and the relative functions were compiled. Through adding the insertion point function and combing the transform technology of spatial coordinate system, three-dimensional modeling for the head was realized with such methods as Boolean operation and rotation. With the proposed parameterization modeling method, 3D head model to meet the requirement can be obtained at arbitrary visual angle and target point only through entering the relevant data of the head by designer. Besides, in combination with AutoCAD 3D modeling technology and VBA programming, the parameterization modeling for other pressure vessel components such as flange and elbow can be performed. The present research can provide a certain reference for exploiting and designing the AutoCAD 3D plug-in in relevant industries.

  13. BWR refill-reflood program: core spray distribution experimental task plan

    Energy Technology Data Exchange (ETDEWEB)

    Eckert, T.

    1981-02-01

    An experimental task plan for the BWR/4 core spray task of the Refill-Reflood Test Program is presented. The test program will provide core spray distribution data for a 30 degree sector of the BWR/4 and 5-218 design. This design uses different nozzle types and different sparger elevations than the BWR/6-218 design which was tested previously. Test parameter ranges are specified; individual tests are defined; and measurement and data utilization plans are defined.

  14. Development of the radiation models of a BWR type reactor and it facility in the SUN-RAH; Desarrollo de modelos de radiacion de un reactor tipo BWR y su instalacion en el SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Barron A, I. [Facultad de Ingenieria, UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: isbarron@yahoo.com.mx

    2005-07-01

    This work about generation models, transport in processes and radioactive contamination of areas of a BWR central, is an amplification to the project developed in the UNAM to have a support tool in subjects or electric generation courses. It is planned about the implementation of models of radiation generation in a BWR type reactor for complement the functions developed in the University Simulator of Nucleo electric- Boiling water reactor (SUN-RAH) which it has been implemented in Simulink of MatLab and it has a model for the dynamics of one nucleo electric central that presents the main characteristics of the reactor vessel, the recirculation system, steam lines, turbines, generator, condensers and feeding water, defined by the main processes that intervene in the generation of energy of these plants. By this way the radiation monitoring systems for area and process, operate simultaneously with the processes of energy generation, with that is possible to observe the changes that present with respect to the operation conditions of the plant, and likewise to appreciate the radiation transport process through the components of the reactor, steam lines and turbines, for different operation conditions and possible faults that they could be presented during the reactor operation. (Author)

  15. In-Situ Nondestructive Evaluation of Kevlar(Registered Trademark)and Carbon Fiber Reinforced Composite Micromechanics for Improved Composite Overwrapped Pressure Vessel Health Monitoring

    Science.gov (United States)

    Waller, Jess; Saulsberry, Regor

    2012-01-01

    NASA has been faced with recertification and life extension issues for epoxy-impregnated Kevlar 49 (K/Ep) and carbon (C/Ep) composite overwrapped pressure vessels (COPVs) used in various systems on the Space Shuttle and International Space Station, respectively. Each COPV has varying criticality, damage and repair histories, time at pressure, and pressure cycles. COPVs are of particular concern due to the insidious and catastrophic burst-before-leak failure mode caused by stress rupture (SR) of the composite overwrap. SR life has been defined [1] as the minimum time during which the composite maintains structural integrity considering the combined effects of stress level(s), time at stress level(s), and associated environment. SR has none of the features of predictability associated with metal pressure vessels, such as crack geometry, growth rate and size, or other features that lend themselves to nondestructive evaluation (NDE). In essence, the variability or surprise factor associated with SR cannot be eliminated. C/Ep COPVs are also susceptible to impact damage that can lead to reduced burst pressure even when the amount of damage to the COPV is below the visual detection threshold [2], thus necessitating implementation of a mechanical damage control plan [1]. Last, COPVs can also fail prematurely due to material or design noncompliance. In each case (SR, impact or noncompliance), out-of-family behavior is expected leading to a higher probability of failure at a given stress, hence, greater uncertainty in performance. For these reasons, NASA has been actively engaged in research to develop NDE methods that can be used during post-manufacture qualification, in-service inspection, and in-situ structural health monitoring. Acoustic emission (AE) is one of the more promising NDE techniques for detecting and monitoring, in real-time, the strain energy release and corresponding stress-wave propagation produced by actively growing flaws and defects in composite

  16. Study of atomic clusters in neutron irradiated reactor pressure vessel surveillance samples by extended X-ray absorption fine structure spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Cammelli, S. [LWV, NES, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Fachbereich C - Physik, Bergische Universitaet Wuppertal, Gauss-Str. 20, 42097 Wuppertal (Germany)], E-mail: Sebastiano.cammelli@psi.ch; Degueldre, C.; Kuri, G.; Bertsch, J. [LWV, NES, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Luetzenkirchen-Hecht, D.; Frahm, R. [Fachbereich C - Physik, Bergische Universitaet Wuppertal, Gauss-Str. 20, 42097 Wuppertal (Germany)

    2009-03-31

    Copper and nickel impurities in nuclear reactor pressure vessel (RPV) steel can form nano-clusters, which have a strong impact on the ductile-brittle transition temperature of the material. Thus, for control purposes and simulation of long irradiation times, surveillance samples are submitted to enhanced neutron irradiation. In this work, surveillance samples from a Swiss nuclear power plant were investigated by extended X-ray absorption fine structure spectroscopy (EXAFS). The density of Cu and Ni atoms determined in the first and second shells around the absorber is affected by the irradiation and temperature. The comparison of the EXAFS data at Cu and Ni K-edges shows that these elements reside in arrangements similar to bcc Fe. However, the EXAFS analysis reveals local irradiation damage in the form of vacancy fractions, which can be determined with a precision of {approx}5%. There are indications that the formation of Cu and Ni clusters differs significantly.

  17. The investigation and analysis about the defects detected in the piping and the pressure vessel of the Kori Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Park, Eun Soo; Park, Ik Geon; Lee, Jeong Soon; Kim, Hyeon Ju; Eom, Byeong Gook; Kim, Hyeon Mook; Cho, Dong Soo [Seoul National Univ. of Technology, Seoul (Korea, Republic of)

    1997-07-15

    The objective of this report os to analysis and investigate the defects detected in the pipe and the pressure vessel of the Kori Nuclear Power Plants during PSI/ISI. In this report, an intelligent database program on windows 95, computer operating system has been built for the defects in the Kori nuclear power plant during PSI/ISI. An intelligent data bases program has been constructed for the effective management of NDE(Nondestructive Evaluation) data carried out the Kori nuclear power plant. Data bases program can be applied to statistical analysis and investigation of the defect data detected during PSI/ISI under fully compatible with windows 95. It is also possible to investigate the NDE data inspected repetitively and the contents of treatment about them.

  18. Fracture toughness evaluation of 20MnMoNi55 pressure vessel steel in the ductile to brittle transition regime: Experiment & numerical simulations

    Science.gov (United States)

    Gopalan, Avinash; Samal, M. K.; Chakravartty, J. K.

    2015-10-01

    In this work, fracture behaviour of 20MnMoNi55 reactor pressure vessel (RPV) steel in the ductile to brittle transition regime (DBTT) is characterised. Compact tension (CT) and single edged notched bend (SENB) specimens of two different sizes were tested in the DBTT regime. Reference temperature 'T0' was evaluated according to the ASTM E1921 standard. The effect of size and geometry on the T0 was studied and T0 was found to be lower for SENB geometry. In order to understand the fracture behaviour numerically, finite element (FE) simulations were performed using Beremin's model for cleavage and Rousselier's model for ductile failure mechanisms. The simulated fracture behaviour was found to be in good agreement with the experiment.

  19. Contributions of Cu-rich clusters, dislocation loops and nanovoids to the irradiation-induced hardening of Cu-bearing low-Ni reactor pressure vessel steels

    Science.gov (United States)

    Bergner, F.; Gillemot, F.; Hernández-Mayoral, M.; Serrano, M.; Török, G.; Ulbricht, A.; Altstadt, E.

    2015-06-01

    Dislocation loops, nanovoids and Cu-rich clusters (CRPs) are known to represent obstacles for dislocation glide in neutron-irradiated reactor pressure vessel (RPV) steels, but a consistent experimental determination of the respective obstacle strengths is still missing. A set of Cu-bearing low-Ni RPV steels and model alloys was characterized by means of SANS and TEM in order to specify mean size and number density of loops, nanovoids and CRPs. The obstacle strengths of these families were estimated by solving an over-determined set of linear equations. We have found that nanovoids are stronger than loops and loops are stronger than CRPs. Nevertheless, CRPs contribute most to irradiation hardening because of their high number density. Nanovoids were only observed for neutron fluences beyond typical end-of-life conditions of RPVs. The estimates of the obstacle strength are critically compared with reported literature data.

  20. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-3 and TSE-4 and update of TSE-1 and TSE-2 analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.; Bolt, S.E.

    1977-11-04

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and four thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. In the first experiment, initiation was not expected and did not occur, although there was a small amount of subcritical crack growth. In the second experiment, initiation of a semicircular flaw took place as expected; the final length along the surface was about four times the initial length, but there was no radial growth. The third and fourth experiments were similar, and the long axial flaw initiated in good agreement with predictions.

  1. Fracture mechanics characterisation of the beltline welding seam of the decommissioned WWER-440 reactor pressure vessel of nuclear power plant Greifswald Unit 4

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, Hans-Werner, E-mail: H.W.Viehrig@hzdr.de [Helmholz Zentrum Dresden-Rossendorf, P.O. Box 510119, 01314 Dresden (Germany); Altstadt, Eberhard; Houska, Mario [Helmholz Zentrum Dresden-Rossendorf, P.O. Box 510119, 01314 Dresden (Germany); Valo, Matti [VTT Manufacturing Technology, P.O. Box 17042, 02044 VTT (Finland)

    2012-01-15

    The paper presents data measured for trepans sampled from the decommissioned WWER-440 reactor pressure vessel of the NPP Greifswald Unit 4. The main focus of this work is on fracture toughness characterisation according to test standard ASTM E1921. Large variations in the evaluated reference temperature values, T{sub 0}, across the wall of the multilayer beltline welding seam were observed. Generally, the ductile-to-brittle transition temperature shift predicted by the Russian code for the present content of deleterious elements P and Cu and the accumulated neutron fluences lies within the amount of the scatter of the measured T{sub 0} values. Metallographic investigations show that the T{sub 0} values measured with T-S oriented Charpy size SE(B) specimens from different thickness locations of the multilayer welding seams strongly depend on the microstructure at the specimen crack tip, and, consequently, on the initial structure of the multilayer welding seam. The RPV integrity is discussed, taking into account a pressurised thermal shock scenario. - Highlights: Black-Right-Pointing-Pointer The paper presents data of samples from a decommissioned reactor pressure vessel. Black-Right-Pointing-Pointer The main focus is on fracture toughness characterisation of the beltline weld seam. Black-Right-Pointing-Pointer Large variation in the evaluated reference temperatures T{sub 0} was observed. Black-Right-Pointing-Pointer T{sub 0} values strongly depend on the microstructure at the specimen crack tip. Black-Right-Pointing-Pointer RPV integrity is discussed, taking into account a pressurised thermal shock scenario.

  2. 等极孔球形压力容器平面缠绕规律%Flat Winding Principle About Equal Polar Hole Spherical Pressure Vessel

    Institute of Scientific and Technical Information of China (English)

    王洪运; 马国峰; 赵亮; 费春东

    2012-01-01

    Based on the theory of space analytic geometry and kinematics,fiber-touched point' s locus equation of flat winding about FRP spherical pressure vessel is established. The functions of fiber-touched point' s space location, and velocity and acceleration of equal polar hole spherical pressure vessel' s flat winding at any time are deduced. Velocity in various directions and cosine value of angular separation of the fiber-touched point were solved by the analysis of doff point's winding principle. Finally,winding locus is simulated with various initial angles by software of Matlab,the effectiveness of the proposed approach is validated by the comparison results of numerical simulation and actual fiber winding style.%基于空间解析几何理论及运动学分析方法,建立复合材料球形压力容器平面缠绕落纱点轨迹运动方程,推导出等极孔球形容器表面缠绕纤维落纱点空间位置、速度以及加速度的求解函数;通过对落纱点运动规律的分析,解出空间落纱点各方向运动速度的大小及其各方向夹角余弦值.最后利用Matlab软件对不同的丝嘴运动平面倾斜角的缠绕轨迹进行模拟,通过该计算方法模拟的轨迹结果与实际球形容器缠绕线型相符合,验证了该算法的有效性.

  3. Health Monitoring of Composite Overwrapped Pressure Vessels (COPVs) Using Meandering Winding Magnetometer ((MWM(Registered Trademark)) Eddy Current Sensors

    Science.gov (United States)

    Russell, Rick; Grundy, David; Jablonski, David; Martin, Christopher; Washabaugh, Andrew; Goldfine, Neil

    2011-01-01

    There are 3 mechanisms that affect the life of a COPV are: a) The age life of the overwrap; b) Cyclic fatigue of the metallic liner; c) Stress Rupture life. The first two mechanisms are understood through test and analysis. A COPV Stress Rupture is a sudden and catastrophic failure of the overwrap while holding at a stress level below the ultimate strength for an extended time. Currently there is no simple, deterministic method of determining the stress rupture life of a COPV, nor a screening technique to determine if a particular COPV is close to the time of a stress rupture failure. Conclusions: Demonstrated a correlation between MWM response and pressure or strain. Demonstrated the ability to monitor stress in COPV at different orientations and depths. FA41 provides best correlation with bottle pressure or stress.

  4. Experimental Analysis of the Influence of Hydrostatic Stress on the Behaviour of an Adhesive Using a Pressure Vessel

    OpenAIRE

    Cognard, J. Y.; Creac' Hcadec, R; da Silva, L. F. M.; Teixeira, F. G.; Davies, Peter; Peleau, Michel

    2011-01-01

    The modelling of the non-linear behaviour of ductile adhesives requires a large experimental database in order to represent accurately the strains which are strongly dependent on the tensile-shear loading combination. Various pressure-dependent constitutive models can be found in the literature, but only a few experimental results are available, for instance, to represent accurately the initial yield surface taking into account the two stress invariants, hydrostatic stress, and von Mises equi...

  5. Uncertainty analysis of suppression pool heating during an ATWS in a BWR-5 plant. An application of the CSAU methodology using the BNL engineering plant analyzer

    Energy Technology Data Exchange (ETDEWEB)

    Wulff, W.; Cheng, H.S.; Mallen, A.N. [Brookhaven National Lab., Upton, NY (United States); Johnsen, G.W. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Lellouche, G.S. [Technical Data Services, Chicago, IL (United States)

    1994-03-01

    The uncertainty has been estimated of predicting the peak temperature in the suppression pool of a BWR power plant, which undergoes an NRC-postulated Anticipated Transient Without Scram (ATWS). The ATWS is initiated by recirculation-pump trips, and then leads to power and flow oscillations as they had occurred at the LaSalle-2 Power Station in March of 1988. After limit-cycle oscillations have been established, the turbines are tripped, but without MSIV closure, allowing steam discharge through the turbine bypass into the condenser. Postulated operator actions, namely to lower the reactor vessel pressure and the level elevation in the downcomer, are simulated by a robot model which accounts for operator uncertainty. All balance of plant and control systems modeling uncertainties were part of the statistical uncertainty analysis that was patterned after the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology. The analysis showed that the predicted suppression-pool peak temperature of 329.3 K (133{degrees}F) has a 95-percentile uncertainty of 14.4 K (26{degrees}F), and that the size of this uncertainty bracket is dominated by the experimental uncertainty of measuring Safety and Relief Valve mass flow rates under critical-flow conditions. The analysis showed also that the probability of exceeding the suppression-pool temperature limit of 352.6 K (175{degrees}F) is most likely zero (it is estimated as < 5-104). The square root of the sum of the squares of all the computed peak pool temperatures is 350.7 K (171.6{degrees}F).

  6. The surface chemistry resulting from low-pressure plasma treatment of polystyrene: The effect of residual vessel bound oxygen

    Science.gov (United States)

    Dhayal, Marshal; Alexander, Morgan R.; Bradley, James W.

    2006-09-01

    The surface chemistry of plasma treated polystyrene samples has been studied in a specially designed low-pressure argon discharge system incorporating in situ XPS analysis. By using an electrostatic grid biasing technique, the plasma source can also be used in a mode preventing ion interactions with the sample. The system, which utilizes a vacuum transfer chamber between plasma and XPS analysis has allowed us to differentiate between the level of oxygen incorporated at the polystyrene surface from residual gas during treatment and that from the exposure of the treated sample to the laboratory atmosphere. Using typical base pressures of about 5 × 10 -3 Pa (4 × 10 -5 Torr) the XPS results show that significant oxygen surface incorporation resulted from oxygen containing species in the plasma itself (i.e. water vapour with 2 × 10 -3 Pa partial pressure). The surface concentration of O was measured at 7.6 at.%. Subsequent atmospheric exposure of the treated samples resulted in only a small increase (of 0.6 at.%) in oxygen incorporation in the form of acid anhydride functionalities. XPS measurements of PS samples exposed to plasmas with no ion-surface component (i.e. exposure from VUV, UV and excited neutral species only) showed no appreciable change in oxygen incorporation compared to those with low-energy ion bombardment from the plasma (free radical sites in this discharge regime.

  7. Vertical Drop of 44-BWR Waste Package With Lifting Collars

    Energy Technology Data Exchange (ETDEWEB)

    A.K. Scheider

    2005-08-23

    The objective of this calculation is to determine the structural response of a waste package (WP) dropped flat on its bottom from a specified height. The WP used for that purpose is the 44-Boiling Water Reactor (BWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. The Uncanistered Waste Disposal Container System is classified as Quality Level 1 (Ref. 4, page 7). Therefore, this calculation is subject to the requirements of the Quality Assurance Requirements and Description (Ref. 16). AP-3. 12Q, Design Calculations and Analyses (Ref. 11) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 44-BWR WP considered in this calculation and provides the potential dimensions and materials for that design.

  8. Non-steady-state Gas Leakage Model for Pressure Vessel Failure%压力容器气体非稳态泄漏模型研究

    Institute of Scientific and Technical Information of China (English)

    王大庆; 张鹏

    2012-01-01

    The gas leakage process after pressure vessel failure was researched in order to calculate the gas leakage rate during a non-steady-state leakage and improve the quantification level of consequence assessment. Based on the model of initial instantaneous flow rate and the dynamic variation of the state parameters in the vessel, a non-steady-state leakage model was built. Then, the model was further analyzed and testified in a case. The results show that using the proposed model, the state parameters and average leakage rate could be obtained at any time during the overall unsteady-state release process (including both sonic release period and subsonic release period). Furthermore, two simplified methods for calculating average leakage rate were worked out for high pressure (higher than 3. 0 MPa) vessels.%为计算气体在非稳态泄漏过程中的泄漏率,提高危害后果评估的量化水平,对压力容器失效后气体泄漏过程进行了研究.基于现有的初始泄漏率模型,结合实际泄漏过程中压力容器内各项状态参数的动态变化规律,构建气体非稳态泄漏模型,并通过计算实例进行分析和验证.结果表明,该模型可计算压力容器气体非稳态泄漏过程中(包括音速泄漏阶段和亚音速泄漏阶段)任意时刻容器内的各项状态参数值和孔口处气体的平均泄漏率;同时,对于储存压力较高(大于3.0 MPa)的容器,提出近似计算总平均泄漏率的2种简化方法.

  9. Development of advanced BWR fuel bundle with spectral shift rod - BWR core characteristics with SSR

    Energy Technology Data Exchange (ETDEWEB)

    Hino, T.; Kondo, T.; Chaki, M.; Ohga, Y. [Hitachi-GE Nuclear Energy, Ltd., 1-1, Saiwai-cho, 3-chome, Hitachi-shi, Ibaraki-ken, 317-0073 (Japan); Makigami, T. [Tokyo Electric Power Company Inc., 1-1-3, Uchisaiwai-cho, Chiyoda-ku, Tokyo, 100-0011 (Japan)

    2012-07-01

    The neutron energy spectrum can be varied during an operation cycle to generate and utilize more plutonium from the non-fissile {sup 238}U by changing the void fraction in the core through control of the core coolant flow rate. This operation method, which is called a spectral shift operation, is practiced in BWRs to save natural uranium. A new component called a spectral shift rod (SSR), which is utilized instead of a conventional water rod, has been introduced to amplify the void fraction change and increase the spectral shift effect. In this study, fuel bundle design with the SSR and core design were carried out for the ABWR and the next generation BWR, HP-ABWR (High-Performance ABWR).The core characteristics with the SSR were evaluated and compared with those when using the conventional water rod. Influences of uncertainty of the water level in the SSR on the safety limit minimum critical power ratio (SLMCPR) were considered for evaluation of the uranium saving effect attained by the SSR. As a result, it was found that the amount of natural uranium needed for an operation cycle could be reduced more than 3% with 20% core coolant flow change and more than 5% with 30% core coolant flow change, in the form of increased discharge exposure by using the SSR compared with the conventional water rod use. (authors)

  10. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pilgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on the PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs.; 9 figs.; 30 tabs.

  11. Techniques for Elastic Properties Measurements of Partial Molten Rocks, Hydrous Minerals and Melts in Gas Pressure Vessels and Multi-Anvil Devices

    Science.gov (United States)

    Mueller, H. J.; Roetzler, K.; Schilling, F. R.; Wehber, M.; Lathe, C.

    2008-12-01

    The interpretation of highly resolved seismic data from Earth's deep interior require measurements of the physical properties of Earth materials under experimental simulated mantle conditions. For deep crustal to uppermost mantle conditions high performance gas pressure vessels enable a virtually unrestricted optimization of the measuring configurations for high p-T-conditions [1]. Exhumed high pressure rocks can be used as representative samples. The paper presents transient measurements of elastic wave velocities for granulite facies rocks under partial melting conditions. Despite the compact natural rock samples as a result of long-term experiments exceeding pressures of 1.5 GPa and temperatures of 1,000°C newly-formed garnets, orthopyroxenes and potash feldspars could be found in the samples after the experiments. Discovering the huge water storage capacity of nominally anhydrous minerals (NAMs) under high pressure conditions dramatically changed our image of state and dynamics of Earth's deep interior [2]. The simulation of these in situ conditions require using of diamond anvil cells (DAC) and multi-anvil devices (MAD) as well as mostly synthetical samples. MADs are more limited in pressure, but provide sample volumes 3 to 7 orders of magnitude bigger. They offer small and even adjustable temperature gradients over the whole sample. The bigger samples make anisotropy and structural effects in complex systems accessible for measurements in principle. Using ultrasonic interferometry the measurement of both elastic wave velocities have no limits for opaque and encapsulated samples. Using the 6 to 8 anvils of a MAD as buffers allow the simultaneous recording of acoustic emissions from different directions of space and consequently the localization of the spikes during ongoing phase transitions and dehydration. The recent development of deformation-DIA MADs (D-DIA) make not only deformation measurements under simulated mantle conditions possible, but also the

  12. Recent trends in the mitigation of the IGSCC through modifications in the water chemistry of BWR reactors; Tendencias recientes en la mitigacion del IGSCC mediante modificaciones en la quimica del agua de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Diaz S, A.; Robles, E.F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    During the last years, the Nuclear Power stations had been that to adequate or to modify the parameters and operational conditions, attempting to maintain and to safeguard the integrity and functionality of its components and systems, as well as the personnel safety involved in its operation. In a Boiling water reactor (BWR), the chemical control of the water, constitutes one of the fundamental aspects to get a sure and reliable operation, having as main objectives: (a) The protection of the reactor vessel, of the structural materials of the same one and of the pipes and components of those recirculation systems against the Intergranular stress corrosion phenomena (IGSCC); (b) To guarantee the integrity of the nuclear fuel minimizing the corrosion phenomena in the fuel elements; and (c) The reduction of the operational dose of the personnel involved directly in the operation and maintenance by means of the control of the activated corrosion products. (Author)

  13. Smart Composite Overwrapped Pressure Vessel - Integrated Structural Health Monitoring System to Meet Space Exploration and International Space Station Mission Assurance Needs

    Science.gov (United States)

    Saulsberry, Regor; Nichols, Charles; Waller, Jess

    2012-01-01

    Currently there are no integrated NDE methods for baselining and monitoring defect levels in fleet for Composite Overwrapped Pressure Vessels (COPVs) or related fracture critical composites, or for performing life-cycle maintenance inspections either in a traditional remove-and-inspect mode or in a more modern in situ inspection structural health monitoring (SHM) mode. Implicit in SHM and autonomous inspection is the existence of quantitative accept-reject criteria. To be effective, these criteria must correlate with levels of damage known to cause composite failure. Furthermore, implicit in SHM is the existence of effective remote sensing hardware and automated techniques and algorithms for interpretation of SHM data. SHM of facture critical composite structures, especially high pressure COPVs, is critical to the success of nearly every future NASA space exploration program as well as life extension of the International Space Station. It has been clearly stated that future NASA missions may not be successful without SHM [1]. Otherwise, crews will be busy addressing subsystem health issues and not focusing on the real NASA mission

  14. Thermohydraulic stability coupled to the neutronic in a BWR; Estabilidad termohidraulica acoplada a la neutronica en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A. [Comision Federal de Electricidad, Central Nucleoelectrica de Laguna Verde, Carretera Cardel-Nautla Km. 42.5, Mpio. Alto Lucero, Veracruz (Mexico); Castlllo D, R. [ININ, Carretera Mexico-Toluca Km 36.5, La Marquesa, Estado de Mexico (Mexico)]. e-mail: gcm9acpp@cfe.gob.mx

    2006-07-01

    In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the

  15. Construction of an external electrode for determination of electrochemical corrosion potential in normal operational conditions of an BWR type reactor for hot cells; Construccion de un electrodo externo para determinacion del potencial de corrosion electroquimico en condiciones normales de operacion de un reactor tipo BWR para celdas calientes

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar T, J.A.; Rivera M, H.; Hernandez C, R. [Departamento de Sintesis y Caracterizacion de Materiales, Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2001-07-01

    The behavior of the corrosion processes at high temperature requires of external devices that being capable to resist a temperature of 288 Centigrade and a pressure of 80 Kg/cm{sup 2}, to give stable and reproducible results of some variable and resisting physically and chemically the radiation. The external electrode of Ag/AgCl fulfils all the requirements in the determination of the electrochemical corrosion potential under normal operational conditions of a BWR type reactor in hot cells. (Author)

  16. Simulation of Thermal Stratification in BWR Suppression Pools with One Dimensional Modeling Method

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Ling Zou; Hongbin Zhang

    2014-01-01

    The suppression pool in a boiling water reactor (BWR) plant not only is the major heat sink within the containment system, but also provides the major emergency cooling water for the reactor core. In several accident scenarios, such as a loss-of-coolant accident and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; the pool temperature distribution also affects the NPSHa (available net positive suction head) and therefore the performance of the Emergency Core Cooling System and Reactor Core Isolation Cooling System pumps that draw cooling water back to the core. Current safety analysis codes use zero dimensional (0-D) lumped parameter models to calculate the energy and mass balance in the pool; therefore, they have large uncertainties in the prediction of scenarios in which stratification and mixing are important. While three-dimensional (3-D) computational fluid dynamics (CFD) methods can be used to analyze realistic 3-D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, resulting in a long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code (Berkeley mechanistic MIXing code in C++) has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by one-dimensional (1-D) transient partial differential equations and substructures (such as free or wall jets) are modeled with 1-D integral models. This allows very large reductions in computational effort compared to multi-dimensional CFD modeling. One heat-up experiment performed at the Finland POOLEX facility, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, is used for

  17. An efficient modeling method for thermal stratification simulation in a BWR suppression pool

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Ling Zou; Hongbin Zhang; Hua Li; Walter Villanueva; Pavel Kudinov

    2012-09-01

    The suppression pool in a BWR plant not only is the major heat sink within the containment system, but also provides major emergency cooling water for the reactor core. In several accident scenarios, such as LOCA and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; and the pool temperature distribution also affects the NPSHa (Available Net Positive Suction Head) and therefore the performance of the pump which draws cooling water back to the core. Current safety analysis codes use 0-D lumped parameter methods to calculate the energy and mass balance in the pool and therefore have large uncertainty in prediction of scenarios in which stratification and mixing are important. While 3-D CFD methods can be used to analyze realistic 3D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, therefore long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. The POOLEX experiments at Finland, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, are used for validation. GOTHIC lumped parameter models are used to obtain boundary conditions for BMIX++ code and CFD simulations. Comparison between the BMIX++, GOTHIC, and CFD calculations against the POOLEX experimental data is discussed in detail.

  18. Radiation Damage Assessment in the Reactor Pressure Vessel of the Integral Inherently Safe Light Water Reactor (I2S-LWR

    Directory of Open Access Journals (Sweden)

    Flaspoehler Timothy

    2016-01-01

    Full Text Available One of the major limiting factors to nuclear reactors lifetime is the radiation-induced material damage in the Reactor Pressure Vessel (RPV. While older reactors were designed assuming a 40-year operating lifetime, new reactor designs are expected to have lifetimes up to 100 years. For safe operation, the integrity of the RPV must be ensured against significant material property changes. In this work, typical neutron damage indicators are calculated in the RPV of the I2S-LWR (Integral Inherently Safe LWR Power Plant, including DPA (displacements per atom and fast neutron fluence (>1 MeV and >0.1MeV. I2S-LWR is a PWR of integral design, which means that its wider downcomer provides additional shielding to the vessel. However, its higher core power density and longer lifetime may offset this advantage. In order to accurately represent the neutron environment for RPV damage assessment, a detailed model based on the preliminary design specifications of the I2S-LWR was developed to be used in the MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations sequence of the Scale6.1 code package. MAVRIC uses the CADIS (Consistent Adjoint-Driven Importance Sampling methodology to bias a fixed-source MC (Monte Carlo simulation. To establish the upper limit of a bounding envelope, a flat-source distribution was used. For the low limit, a center-peaked source was generated using the KENO-VI criticality sequence assuming uniform fresh fuel core. Results based on the preliminary I2S-LWR model show that DPA rates and fast fluence rates are conservatively 75% lower than in typical PWRs being operated currently in the US.

  19. Radiation Damage Assessment in the Reactor Pressure Vessel of the Integral Inherently Safe Light Water Reactor (I2S-LWR)

    Science.gov (United States)

    Flaspoehler, Timothy; Petrovic, Bojan

    2016-02-01

    One of the major limiting factors to nuclear reactors lifetime is the radiation-induced material damage in the Reactor Pressure Vessel (RPV). While older reactors were designed assuming a 40-year operating lifetime, new reactor designs are expected to have lifetimes up to 100 years. For safe operation, the integrity of the RPV must be ensured against significant material property changes. In this work, typical neutron damage indicators are calculated in the RPV of the I2S-LWR (Integral Inherently Safe LWR) Power Plant, including DPA (displacements per atom) and fast neutron fluence (>1 MeV and >0.1MeV). I2S-LWR is a PWR of integral design, which means that its wider downcomer provides additional shielding to the vessel. However, its higher core power density and longer lifetime may offset this advantage. In order to accurately represent the neutron environment for RPV damage assessment, a detailed model based on the preliminary design specifications of the I2S-LWR was developed to be used in the MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations) sequence of the Scale6.1 code package. MAVRIC uses the CADIS (Consistent Adjoint-Driven Importance Sampling) methodology to bias a fixed-source MC (Monte Carlo) simulation. To establish the upper limit of a bounding envelope, a flat-source distribution was used. For the low limit, a center-peaked source was generated using the KENO-VI criticality sequence assuming uniform fresh fuel core. Results based on the preliminary I2S-LWR model show that DPA rates and fast fluence rates are conservatively 75% lower than in typical PWRs being operated currently in the US.

  20. Particle Imaging Velocimetry Technique Development for Laboratory Measurement of Fracture Flow Inside a Pressure Vessel Using Neutron Imaging

    Energy Technology Data Exchange (ETDEWEB)

    Polsky, Yarom [ORNL; Bingham, Philip R [ORNL; Bilheux, Hassina Z [ORNL; Carmichael, Justin R [ORNL

    2015-01-01

    This paper will describe recent progress made in developing neutron imaging based particle imaging velocimetry techniques for visualizing and quantifying flow structure through a high pressure flow cell with high temperature capability (up to 350 degrees C). This experimental capability has great potential for improving the understanding of flow through fractured systems in applications such as enhanced geothermal systems (EGS). For example, flow structure measurement can be used to develop and validate single phase flow models used for simulation, experimentally identify critical transition regions and their dependence on fracture features such as surface roughness, and study multiphase fluid behavior within fractured systems. The developed method involves the controlled injection of a high contrast fluid into a water flow stream to produce droplets that can be tracked using neutron radiography. A description of the experimental setup will be provided along with an overview of the algorithms used to automatically track droplets and relate them to the velocity gradient in the flow stream. Experimental results will be reported along with volume of fluids based simulation techniques used to model observed flow.