WorldWideScience

Sample records for bwr pressure vessel

  1. Service experience of BWR pressure vessels

    International Nuclear Information System (INIS)

    The overall service experience with Boiling Water Reactor (BWR) pressure vessels has been excellent. The only significant factor that impacted the service performance has been thermal fatigue cracking of feedwater inlet nozzle. This concern has been mitigated by eliminating the source of thermal cycling stress through design and operational changes. Although stress corrosion cracking has occurred in early atypical steam generator vessel designs, analysis and field experience has indicated that this mechanism is not expected in the BWR reactor pressure vessel (RPV). Other limited materials related cracking problems have been associated with RPV stainless steel and nickel-base alloy attachments and penetrations. Solutions to these problems have involved design and materials modifications. Finally, due to the low end of life fluence resulting from the large core-to-RPV-wall water annulus, irradiation embrittlement effects are minimal

  2. Inservice inspection of Halden BWR pressure vessel

    International Nuclear Information System (INIS)

    A description is given of how the recertification inspection of the 20 years old Halden Reactor pressure vessel was carried out in accordance with the latest ASME-CODES, despite the fact that inspection accessibility was poor. As no volumetric inspection had been carried out since the preservice radiography in 1957, the ultrasonic inspection included the high flux region of all welds. In total 70% of longitudinal welds and 20% of bottom circumferential welds were inspected as well as the bottom nozzle connection. The vessel was not designed with provisions for inservice inspection, the welds are unaccessible from the outside and removal of the lid is virtually impossible. The ultrasonic probes could only be loaded through 77 mm diameter holes in the top lid and remotely positioned inside the vessel. The inspection was performed using 450C and 60OC 1 MHz angle probes and 2.25 MHz normal probes in immersion technique. In a zone around the welds, small regions with lack of bonding between the stainless steel cladding and the boiler steel were revealed. One root defect known and accepted from the preservice radiographs was examined. The defect was found to be 6x30mm as a maximum and well within acceptable limits according to the fracture mechanics analysis method recommended in ASME X1. The inspection required a period of three weeks' work in the reactor hall. (UK)

  3. A damage tolerance analysis of an old BWR pressure vessel

    International Nuclear Information System (INIS)

    A program has been completed in Sweden where the oldest of the Swedish BWR-units has been subjected to an extensive investigation regarding signs of aging and degradation. The core and all the internal parts were removed and the stripped pressure vessel was decontaminated. Almost every weld in the vessel including all nozzles in the bottom head have then been inspected by aid of UT-inspection. An important part of the programs was the damage tolerance analysis. It has involved postulated surface cracks and embedded cracks (positioned along and across the welds) in all the inspected welds. By using the R6-method the maximum acceptable and critical crack size have then been determined for the most limiting load case and accounting for the individual material properties of each weld, cladding and base material. Of special interest is the core region. The base material is made of A 302 Grade B with rather high Cu- and Ni-contents, which have caused irradiation embrittlement in the beltline region. This implies that during certain cold loading cases, the critical crack size for a postulated surface crack in the core region, can be quite small. However, for load cases during normal operation, the material is on the upper shelf region and the critical crack sizes are large

  4. Analysis of the integrity of the pressure vessel of the BWR type nuclear reactor

    International Nuclear Information System (INIS)

    The presssure vessel of a BWR type reactor was monitored for cracking during alternating events of its in-service life. The monitoring was to determine criticality of fractures catastrophic fractures and the velocity of fracture propagation. Detected cracks were evaluated as specified in ASME code section XI, of a minimum wall thickness of 2.5% crack growths were compared a) of 1/10 of the critical maximum size and b) at in-service inspection intervals according to ASME recommendations to be established at the Laguna Verde nuclear plant. Finally conclusions are made and discussed. (author)

  5. 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).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-02-01

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

  7. Importance of momentum effects in BWR reactor vessel modeling

    International Nuclear Information System (INIS)

    The most severe class of events (other than design-basis LOCA's) for boiling water reactors (BWR's) entails rapid pressurization of the reactor vessel (RV) due to rapid valve closures in the main steamlines. The severity of these events is caused by the fact that the pressurization of the core results in a rapid decrease in the core void fraction and, coupled to a strong negative void coefficient of reactivity, in a rapid power increase if the reactor protective system is slow in responding. The DYNODE-B program models the nuclear steam supply system of BWR's. Early versions only explicitly calculated the RV dome pressure. The latest version permits optional explicit calculation of the dome, core outlet, and core average pressures. The use of the new option has shown the influence of the two-phase momentum effects on the core pressure and flow transients resulting from pressure wave reflections at the liquid (incompressible) core inlet region

  8. Water level measurement system in reactor pressure vessel of BWR and hydrogen concentration monitoring system for severe accident

    International Nuclear Information System (INIS)

    TEPCO's Fukushima Daiichi Nuclear Power Station Accident caused severe accident to lose functions of many instrumentation systems. As a result, many important plant parameters couldn't be monitored. In order to monitor plant parameters in the case of severe accident, new instrumentation systems available in the severe conditions are being developed. Water level in reactor pressure vessel and hydrogen concentration in primary containment vessel are one of the most important parameters. Performance test results about water level measurement sensor and hydrogen sensor in severe environmental conditions are described. (author)

  9. Analysis of the microstructural evolution of the damage by neutron irradiation in the pressure vessel of a nuclear power reactor BWR

    International Nuclear Information System (INIS)

    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

  10. Calculation of a leak in the bottom of a bwr-pressure vessel using THYDE-B1

    International Nuclear Information System (INIS)

    The present report describes the modelling of a boiling water reactor of the Gemeinschaftskraftwerk Tullnerfeld (GKT) plant type. The corresponding input data set for the Japanese computer code THYDE-B1 allows the simulation of the thermohydraulic transient within the reactor coolant system during a loss of coolant accident. The initiating event 'Leak in the Reactor Pressure Vessel Bottom' has been calculated as an application exercise. The results are discussed using graphic representation, where one should be aware that the calculated data do not correspond with any specific plant configuration of the boiling water reactor generic design. 1 ref., 7 figs. (Author)

  11. BWR zero pressure containment

    International Nuclear Information System (INIS)

    This patent describes the operation of a nuclear reactor system, the system including a containment defining a drywall space wherein a nuclear reactor is disposed, there being a suppression pool in the containment with the suppression pool having a wetwell space above a level of the pool to which an non-condensable gases entering the suppression pool can vent. It comprises: continuously exhausting the wetwell space to remove gas mixture therefrom while admitting inflow of air from an atmospheric source thereof to the wetwell during normal operation by blocking off the inflow during a loss-of-coolant-accident whenever a pressure in the wetwell space is above a predetermined value, and subjecting the gas subsequent to its removal from the wetwell to a treatment operation to separate any particulate material entrained therein from the gas mixture

  12. Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessel internals

    International Nuclear Information System (INIS)

    . The guidance reports are directed at technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific plant components addressed in the reports. The report addresses the reactor pressure vessel internals in BWRs. Maintaining the structural integrity of these reactor pressure vessel internals throughout NPP service life, in spite of several ageing mechanisms, is essential for plant safety

  13. Vibration characteristics of BWR primary containment vessel

    International Nuclear Information System (INIS)

    This paper reports on a part of a seismic proving test of a BWR Primary Containment Vessel (PCV), a vibration analysis that is carried out considering the effect of reinforcement around attached masses such as equipment hatches and a personnel airlock. A deflected shape with the local attached masses and reinforcement is expanded using free vibration modes of an axisymmetric PCV without attached masses and reinforcement. The free vibration modes are calculated considering a coupling effect between the PCV shell and suppression pool water using conical shell finite elements and ring fluid elements. The same expanding functions and free vibration modes are used to express strain energy, kinetic energy and work done by inertia forces due to input acceleration considering locally varying thickness of the PCV and attached masses. The equations of motion are obtained by substituting the strain energy, kinetic energy, and work done by inertia forces into the Lagrange's equation. Calculated natural frequencies, free vibration modes and frequency response functions are compared with numerical results obtained by a general shell finite element analysis and with test results using the large-scale high-performance vibration table of Tadotsu Engineering Laboratory, Nuclear Power Engineering Test Center (NUPEC). The comparison shows a reasonable agreement, leading to better understanding of the dynamic characteristics of the BWR PCV

  14. Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessels

    International Nuclear Information System (INIS)

    . The guidance reports are directed at technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific plant components addressed in the reports. This report addresses the reactor pressure vessel (RPV) in BWRs. Maintaining the structural integrity of this RPV throughout NPP service life in spite of several ageing mechanisms is essential for plant safety

  15. Damage by radiation in structural materials of BWR reactor vessels

    International Nuclear Information System (INIS)

    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 1018 n/cm2) in the TRIGA Mark III Salazar reactor and separately with Ni+3 ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A2). (Author)

  16. High Pressure Boiling Water Reactor HP-BWR

    International Nuclear Information System (INIS)

    Some four hundred Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) have been in operation for several decades. The presented concept, the High Pressure Boiling Water Reactor (HP-BWR) makes use of the operating experiences. HP-BWR combines the advantages and leaves out the disadvantages of the traditional BWRs and PWRs by taking in consideration the experiences gained during their operation. The best parts of the two traditional reactor types are used and the troublesome components are left out. HP-BWR major benefits are; 1. Safety is improved; -Gravity operated control rods -Large space for the cross formed control rods between fuel boxes -Bottom of the reactor vessel is smooth and is without penetrations -All the pipe connections to the reactor vessel are well above the top of the reactor core -Core spray is not needed -Internal circulation pumps are used. 2. Environment friendly; -Improved thermal efficiency, feeding the turbine with ∼340 oC (15 MPa) steam instead of ∼285 oC (7MPa) -Less warm water release to the recipient and less uranium consumption per produced kWh and consequently less waste is produced. 3. Cost effective, simple; -Direct cycle, no need for complicated steam generators -Moisture separators and steam dryers are inside the reactor vessel and additional separators and dryers can be installed inside or outside the containment -Well proved simple dry containment or wet containment can be used. (author)

  17. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 1: Plenary sessions; Pressure vessel research; BWR strainer blockage and other generic safety issues; Environmentally assisted degradation of LWR components; Update on severe accident code improvements and applications

    International Nuclear Information System (INIS)

    This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following information: (1) plenary sessions; (2) pressure vessel research; (3) BWR strainer blockage and other generic safety issues; (4) environmentally assisted degradation of LWR components; and (5) update on severe accident code improvements and applications. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database

  18. 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 ...

  19. 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)

  20. Study on reactor vessel replacement (RVR) for 1100 MW class BWR plants in Japan

    International Nuclear Information System (INIS)

    Plant Life Management (PLM) is being studied in Japan, and reactor vessel replacement (RVR) is being considered as one option. Since reactor internals, except for reusable parts, and the reactor pressure vessel (RPV) are replaced, the RVR provides an effective technology for extending the service life of nuclear power plants substantially. At ICONE 7, we reported on the technical viability of the RVR for BWR4-type 800 MWe class plants. This time, we rationalized the RVR method through a study for BWR5-type 1100 MWe class plants to reduce the RVR duration and evaluated the technical viability and the economic efficiency of the method. In addition, we discuss how to dispose of the RPV to complete a scenario of the process from the RVR to its final disposal. (author)

  1. 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

  2. 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

  3. Reliability of BWR high pressure core cooling

    International Nuclear Information System (INIS)

    The high pressure coolant injection system (HPCI), and the reactor core isolation cooling system (RCIC) are steam turbine driven systems that can inject water into a boiling water reactor at full operating pressure. Their purpose is to supply water during any failure that allows water to be lost while the reactor is at pressure and temperature. A large number of BWR plants are not meeting HPCI and RCIC performance goals for core cooling. NSAC considers concurrent failure of NPCI and RCIC to be the most probable potential cause of low reactor water level and possibly fuel damage in a boiling water reactor. Between January 1978 and May 1981, 169 licensee event reports were filed where HPCI or RCIC was inoperable or was declared inoperable. The present effort has shown that at least 40% of NPCI and RCIC problems might be averted by a high quality preventive maintenance program. About half of the plants do not perform cold quick-start surveillance testing of HPCI and RCIC. They do perform routine startup tests, but the equipment is first preheated and the startup is relatively gentle. However, emergency start-ups are abrupt and from the cold condition. Therefore, cold quick-start testing is the only way to assure that all components, control systems, and instruments are functioning correctly for automatic safety initiation. (author)

  4. Reactor pressure vessel materials

    International Nuclear Information System (INIS)

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. Chapter 3 offers a detailed treatment of the selection criteria and properties of reactor pressure vessel materials. The main attention is directed towards steel and ingot making and the subsequent material processing

  5. Reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Research and development activities related to reactor pressure vessel steels during 1997 are reported. The objectives of activities of the Belgian Nuclear Research Centre SCK/CEN in this domain are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate a methodology on a broad database; (3) to achieve regulatory acceptance and industrial use

  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. Pressure vessel design manual

    Energy Technology Data Exchange (ETDEWEB)

    Moss, D.R.

    1987-01-01

    The first section of the book covers types of loadings, failures, and stress theories, and how they apply to pressure vessels. The book delineates the procedures for designing typical components as well as those for designing large openings in cylindrical shells, ring girders, davits, platforms, bins and elevated tanks. The techniques for designing conical transitions, cone-cylinder intersections, intermediate heads, flat heads, and spherically dished covers are also described. The book covers the design of vessel supports subject to wind and seismic loads and one section is devoted to the five major ways of analyzing loads on shells and heads. Each procedure is detailed enough to size all welds, bolts, and plate thicknesses and to determine actual stresses.

  8. 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)

  9. Reactor pressure vessel cooling device

    International Nuclear Information System (INIS)

    A spray pipeline for the head of a pressure vessel is connected to a spray for the pressure vessel head of the reactor pressure vessel. The pipeline is equipped with a spray flow rate control valve for controlling the flow rate of spray jetted from the head spray to the inside of the reactor and a thermometer for spray water. A reactor pressure vessel cooling and controlling portion intakes temperature signals sent from a thermometer for the head of the pressure vessel, a thermometer for the flange of the pressure vessel and a thermometer for the body of the pressure vessel, and a thermometer for spray water. Then, it outputs restriction signal for the opening degree of the spray flow rate control valve to restrict the spray water flow rate within an appropriate range when the difference between the temperature of each portion of the pressure vessel and the temperature of spray water is increased to greater than an allowable temperature difference. With such procedures, spray operation for the head of the pressure vessel can be conducted not only upon emergency but also during reactor cooling operation. (I.N.)

  10. Characteristics of fluctuating pressure generated in BWR main steam lines

    International Nuclear Information System (INIS)

    The BWR-3 steam dryer in the Quad Cities Unit 2 Nuclear Power Plant was damaged by high cycle fatigue due to acoustic-induced vibration. The dryer failure was as attributed to flow-induced acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSLs). The acoustic resonance was considered to be generated by interaction between the sound field and an unstable shear layer across the closed side branches with SRV stub pipes. We have started a research program on BWR dryers to develop their loading evaluation methods. Moreover, it has been necessary to evaluate the dryer integrity of BWR-5 plants which are the main type of BWR in Japan. In the present study, we used 1/10-scale BWR tests and analyses to investigate the flow-induced acoustic resonance and acoustic characteristics in MSLs. The test apparatus consisted of a steam dryer, a steam dome and 4 MSLs with 20 SRV stub pipes. A finite element method (FEM) was applied for the calculation of three-dimensional wave equations in acoustic analysis. We demonstrated that remarkable fluctuating pressures occurred in high and low frequency regions. High frequency fluctuating pressures was generated by the flow-induced acoustic resonance in the SRV stub pipes. Low frequency fluctuating pressure was generated in an MSL with the dead leg. The frequency of the latter almost coincided with the natural frequency of the MSL with the dead leg. The amplitude of the fluctuating pressures in the multiple stub pipes became more intense because of interaction between them compared with that in the single stub pipe. Acoustic analysis results showed that the multiple stub pipes caused several natural frequencies in the vicinity of the natural frequency of the single stub pipe and several modes of the standing wave in the MSLs. (author)

  11. Multilayer Composite Pressure Vessels

    Science.gov (United States)

    DeLay, Tom

    2005-01-01

    A method has been devised to enable the fabrication of lightweight pressure vessels from multilayer composite materials. This method is related to, but not the same as, the method described in gMaking a Metal- Lined Composite-Overwrapped Pressure Vessel h (MFS-31814), NASA Tech Briefs, Vol. 29, No. 3 (March 2005), page 59. The method is flexible in that it poses no major impediment to changes in tank design and is applicable to a wide range of tank sizes. The figure depicts a finished tank fabricated by this method, showing layers added at various stages of the fabrication process. In the first step of the process, a mandrel that defines the size and shape of the interior of the tank is machined from a polyurethane foam or other suitable lightweight tooling material. The mandrel is outfitted with metallic end fittings on a shaft. Each end fitting includes an outer flange that has a small step to accommodate a thin layer of graphite/epoxy or other suitable composite material. The outer surface of the mandrel (but not the fittings) is covered with a suitable release material. The composite material is filament- wound so as to cover the entire surface of the mandrel from the step on one end fitting to the step on the other end fitting. The composite material is then cured in place. The entire workpiece is cut in half in a plane perpendicular to the axis of symmetry at its mid-length point, yielding two composite-material half shells, each containing half of the foam mandrel. The halves of the mandrel are removed from within the composite shells, then the shells are reassembled and bonded together with a belly band of cured composite material. The resulting composite shell becomes a mandrel for the subsequent steps of the fabrication process and remains inside the final tank. The outer surface of the composite shell is covered with a layer of material designed to be impermeable by the pressurized fluid to be contained in the tank. A second step on the outer flange of

  12. BWR Mark II ex-vessel corium interaction analyses

    International Nuclear Information System (INIS)

    This report describes the results of a series of studies conducted to investigate the behavior of core debris within a BWR Mark II containment. These studies focused on the interaction of core debris with concrete and steel structures (downcomers and inpedestal floor drains) within the drywell, the transport of debris through these drains and downcomers into the wetwell, and on debris-water reactions within the wetwell. Estimates of the conditions under which debris would penetrate the in-pedestal drain lines, the time-dependent behavior of the debris within the drain lines, and the amount of debris which might enter the suppression pool via these drain lines are provided. An assessment of the conditions under which the upper lip of the downcomers would be expected to fail (i.e. melt) due to exposure to hot core debris is presented. Finally, the unique characteristics of debris water interactions in Mark II containments are discussed, the existing knowledge base regarding core-concrete debris-water interactions is summarized, and an evaluation of the applicability of the MELCOR 1.80 code's debris-water interaction model to BWR Mark II's is presented

  13. Pressure vessel control

    International Nuclear Information System (INIS)

    The in-service inspection machine (MIS) is robot machine which aim is to inspect, under water, the PWR type vessels (with high radiation level): inspection of the welded joints, of the coatings, detection of the defects situated under the coating, etc... This type of machine utilizes several non destructive testing technologies: ultrasonics with focused transducers, gammagraphy, television, and eddy currents

  14. A Deterministic/probalistic analysis of Ex-Vessel melt risk in a BWR

    OpenAIRE

    Abal López, Javier

    2006-01-01

    The present study is concerned with deterministic and probabilistic analysis of ex-vessel melt risks in a Swedish designed BWR plant. The focus is placed on a station blackout (SBO) scenario, with immediate SCRAM and subsequent activation of the main steam valve isolation (at 52 s). Four sequences were examined in detail to study the effect of two valves systems related to the operation of ADS (Automatic Depressurization System), and cavity flooding by water from suppression po...

  15. 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.

  16. The German pressure vessel code

    International Nuclear Information System (INIS)

    The philosophy and the safety aspects of the German pressure vessel code is summarised. Emphasis is laid upon material selection, design, manufacture and inspection. The AD specifications, which are embedded in the German regulation for the prevention of accidents 'Pressure Vessels' are compiled primarily to give a measure of safety to the average plant. The possible types of failure are discussed in connection with design aspects. It is shown in detail how the AD specifications count for inelastic material behaviour. A comparison is also made between the German code and the ASME boiler and pressure vessel code. Comparatively, the allowable stresses due to the German code are higher, leading to thinner walls in many cases. This is compensated for by rigorous inspection and non-destructive testing during manufacture. Regular examinations round up the system of safety. It is concluded that the German code gives a high degree of safety and flexibility to the manufacturer. (author)

  17. Estimation method of water level behavior in the case of large pressure change in a BWR

    International Nuclear Information System (INIS)

    In a BWR, coolant of core and upper plenum involves so much void volume that free surface level change at downcomer is conspicuous owing to increase and decrease of void volume influenced by pressure change. When mass balance in a reactor vessel becomes non-equilibrium due to steam valve stuck open or feedwater pump trip, difference between liquid level and mixture level becomes very large because of void increased due to depressurization. Therefore, it is very difficult to estimate changes in water level after void exclusion by isolation valve closure etc. So a new parameter ''effective increased void volume'' was contrived to estimate water level in the occurrance of above mentioned phenomena, as a result of consideration about relation between discharged mass and reactor pressure. Degree of water level change under initial operating conditions and reactor pressure change can be estimated by using this parameter. (author)

  18. WWER reactor pressure vessel design

    International Nuclear Information System (INIS)

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. The details of WWER reactor pressure vessel design, as used in the PWR type reactors of the former USSR are given in chapter 4. General design concepts are given as well as specific material requirements. Safety concepts and lifetime assurance are discussed briefly

  19. The calculation of pressure vessels

    International Nuclear Information System (INIS)

    The calculation guidelines of the Arbeitsgemeinschaft Druckbehaelter (task group for pressure vessels) have been revised with the following objective: conversion to international standards (SI), adaption to the latest state of guidelines for production and testing, revision of the contents of individual regulations. Another target of the cooperating interest groups of producers, operators, and supervisory bodies was a harmonization of the approaches for calculation with other German guidelines, in particular the Technische Regeln fuer Dampfkessel (technical regulations for steam boilers). (orig./RW)

  20. Severe accident instrumentation systems for BWR water level and temperature in primary containment vessel measurements

    International Nuclear Information System (INIS)

    The severe accident at TEPCO's Fukushima Daiichi nuclear power station (TF1 accident) in March 2011 brought the lost of the functions of many instrumentation systems. In order to enable the measurements of the important parameters such as reactor water level, temperature and so on even in a case such as the TF1 accident occurs, severe accident instrumentation systems are being developed. In this paper, new system configurations of BWR water level measurement and temperature measurement in primary containment vessels are proposed. Then performance tests for prototype sensors of these measurement systems under high temperature conditions are described. (author)

  1. Study of behavior on bonding and failure mode of pressurized and doped BWR fuel rod

    International Nuclear Information System (INIS)

    The study of transient behavior on the bonding and the failure mode was made using the pressurized/doped 8 x 8 BWR type fuel rod. The dopant was mullite minerals consisted mainly of silicon and aluminum up to 1.5 w/o. Pressurization of the fuel rod with pure helium was made to the magnitude about 0.6 MPa. As a reference, the non-pressurized/non-doped 8 x 8 BWR fuel rod and the pressurized/7 x 7 BWR fuel rod up to 0.6 MPa were prepared. Magnitude of energy deposition given to the tested fuel rods was 248, 253, and 269 cal/g·fuel, respectively. Obtained results from the pulse irradiation in NSRR are as follows. (1) It was found from the experiment that alternation of the fuel design by the adoption of pressurization up to 0.6 MPa and the use of wider gap up to 0.38 mm could avoid the dopant BWR fuel from the overall bonding. The failure mode of the present dopant fuel was revealed to be the melt combined with rupture. (2) The time of fuel failure of the pressurized/doped 8 x 8 BWR fuel defected by the melt/rupture mode is of order of two times shorter than that of the pressurized/ 7 x 7 BWR defected by the rupture mode. Failure threshold of the pressurized/doped 8 x 8 BWR BWR tended to be lower than that of non-pressurized/non-doped 8 x 8 BWR one. Cracked area of the pressurized/doped 8 x 8 BWR was more wider and magnitude of oxidation at the place is relatively larger than the other tested fuels. (3) Failure mode of the non-pressurized/ 8 x 8 BWR fuel rod was the melt/brittle accompanied with a significant bonding at failed location. While, failure mode of the pressurized/ 7 x 7 BWR fuel rod was the cladding rupture accompanied with a large ballooning. No bonding at failed location of the latter was observed. (author)

  2. Introduction of Korean pressure vessel code

    International Nuclear Information System (INIS)

    As of the present date, pressure related components installed in Korea's Nuclear and Thermal Power Plants have been designed, manufactured, and operated in accordance with various foreign Pressure Vessel Codes. Because of the difficulties applying differing Codes, the Korean Engineering Societies have been putting forth the efforts in the development of the Korean Pressure Vessel Code since early 1992. This paper introduces the Korean Pressure Vessel Code, relating to pressure retaining components, and discusses the contents of that Code

  3. 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

  4. Automatic pressure reduction device for BWR type reactor

    International Nuclear Information System (INIS)

    Purpose: To suppress the leakage amount from main steam isolation valves after the closure of the valves upon main steam pipeway rupture at the outside of a reactor container. Constitution: Main steam isolation valves disposed on both sides of the main steam pipeway penetration portion in a reactor container are closed, as well as a safety relief valve of the main steam pipeways disposed in the reactor container is opened by a rupture signal generated by the rupture of the main steam pipeways at the outside of the reactor container. Since the pressure in the reactor pressure vessel is automatically reduced after a predetermined of time even if rupture is caused to the main steam pipeways at the outside of the reactor pressure vessel, it is possible to suppress the leaking amount sufficiently after the entire closure of the main steam isolation valves and contribute to the improvement for the nuclear power plant safety. (Yoshihara, H.)

  5. Pressure vessel and piping codes

    International Nuclear Information System (INIS)

    Section III of the ASME Boiler and Pressure Vessel Code contains simplified design formulas for placing bounds on the plastic deformations in nuclear power plant piping systems. For Class 1 piping a simple equation is given in terms of primary load stress indices (B1 and B2) and nominal pressure and bending stresses. The B1 and B2 stress indices reflect the capacities of various piping products to carry load without gross plastic deformation. In this paper, the significance of the indices, nominal stresses, and limits given in the Code for Class 1 piping and corresponding requirements for Class 2 and Class 3 piping are discussed. Motivation behind recent (1978-1981) changes in the indices and in the associated stress limits is presented

  6. BWR LOCA integral test simulating a 100 % main steam line break outside reactor containment vessel in ROSA-III program, RUN 955

    International Nuclear Information System (INIS)

    This report presents the ROSA-III experimental results of RUN 955, which simulates a 100 % steam line break (SLB) LOCA outside the BWR reactor containment vessel (RCV) with an assumption of high pressure core spray (HPCS) system failure. The ROSA-III test facility simulates a BWR system with volumetric scale of 1/424 and has the principal systems, i.e., four half-length electrically heated fuel bundles, two active recirculation loops, four types of ECCSs, and steam and feedwater systems. The report clarifies that the 100 % SLB LOCA outside the RCV becomes similar to a small SLB LOCA with the safety/relief valve (SRV) operation after the main steam isolation valve (MSIV) closure and that it is analogous to a small recirculation line break (RLB) LOCA with break area less than 2 % of the scaled pipe flow area. (author)

  7. Advanced ultrasonic inspection system for the ID-inspection of reactor pressure vessels of BWRs

    International Nuclear Information System (INIS)

    A newly-developed, modular ultrasonic examination system has been developed by Siemens for the ID inspection of BWR RPV'S. It is based on the phased-array technique with hybrid probes using the latest in manipulator and control equipment technology to allow the often hard-to-access weld areas of older reactor pressure vessels in US BWR plants to be examined within a very short time and with minimal radiation exposure of the examination personnel. New NRC stipulations requiring almost complete ultrasonic examination of all RPV welds can be fully satisfied using this system for the ID inspection of all longitudinal and circumferential welds above the jet pump baffle plate

  8. Reactor pressure vessel thermal annealing

    International Nuclear Information System (INIS)

    The steel plates and/or forgings and welds in the beltline region of a reactor pressure vessel (RPV) are subject to embrittlement from neutron irradiation. This embrittlement causes the fracture toughness of the beltline materials to be less than the fracture toughness of the unirradiated material. Material properties of RPVs that have been irradiated and embrittled are recoverable through thermal annealing of the vessel. The amount of recovery primarily depends on the level of the irradiation embrittlement, the chemical composition of the steel, and the annealing temperature and time. Since annealing is an option for extending the service lives of RPVs or establishing less restrictive pressure-temperature (P-T) limits; the industry, the Department of Energy (DOE) and the Nuclear Regulatory Commission (NRC) have assisted in efforts to determine the viability of thermal annealing for embrittlement recovery. General guidance for in-service annealing is provided in American Society for Testing and Materials (ASTM) Standard E 509-86. In addition, the American Society of Mechanical Engineers (ASME) Code Case N-557 addresses annealing conditions (temperature and duration), temperature monitoring, evaluation of loadings, and non-destructive examination techniques. The NRC thermal annealing rule (10 CFR 50.66) was approved by the Commission and published in the Federal Register on December 19, 1995. The Regulatory Guide on thermal annealing (RG 1.162) was processed in parallel with the rule package and was published on February 15, 1996. RG 1.162 contains a listing of issues that need to be addressed for thermal annealing of an RPV. The RG also provides alternatives for predicting re-embrittlement trends after the thermal anneal has been completed. This paper gives an overview of methodology and recent technical references that are associated with thermal annealing. Results from the DOE annealing prototype demonstration project, as well as NRC activities related to the

  9. 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.

  10. Experimental investigations of BWR pressure suppression pool behavior under loss of coolant accident conditions

    International Nuclear Information System (INIS)

    The experiments discussed in this paper look into different processes which may occur during a loss-of-coolant accident in the pressure suppression pool of a Boiling Water Reactor (BWR). These processes include: a) development of a thermal stratification, b) bubble dynamics and related water flow during continuous release of air and c) air blowdown and associated water slug phenomenon in the water pool. The experiments have been performed in the THAI test facility, which is a cylindrical vessel of 9.2 m height, 3.2 m diameter and with a gas volume of 60 m3. The variation in the investigated test parameters included, steam and air mass flux, initial water pool temperature, blowdown pressures, downcomer submergence, etc. A systematic variation of the test parameters allowed better understanding of the phenomena. Experiments discussed in this paper were performed with a vertical downcomer of 0.1 m diameter and 2 m submergence depth in the water pool. For the blowdown experiments, a separate interconnecting vessel of 1 m3 volume was used to inject air at pressures between 3 bar and 10 bar. A high speed camera (1000 fps) was installed to visualize the formation and propagation of air bubbles in the suppression pool and the resulting pool swelling phenomena. Customized instrumentation applied during the tests included grids of densely spaced thermocouples and of pressure transducers at various locations in order to capture the temperature distribution in the pool and the water slug induced pressure loadings, respectively. The present paper discusses the main outcome of the selected experiments. On the whole the experimental data may be very useful for code validation. (authors)

  11. TRAC-BF1 three-dimensional BWR vessel thermal-hydraulic and ANSYS stress analyses for BWR core shroud cracking

    International Nuclear Information System (INIS)

    TRAC-BF1 transient analysis for a General Electric boiling water reactor (BWR/2) was performed for a main steam line break (MSLB) event and a recirculation line break (RLB) transient to determine the susceptibility of the core shroud to geometric alteration when intergranular stress corrosion cracking (IGSCC) is present at the upper H3 and lower H7 core shroud weld locations. The TRAC-BF 1 model developed for the MSLB transient utilized a 1-D vessel component model with both main steam lines rupturing at transient initiation. A detailed ANSYS model was developed to find the forces imposed at the upper H3 weld positions. The TRAC-BF1 model developed for the RLB transient utilized a 3-D four azimuthal sector vessel component. FORTRAN programs were developed to calculate the forces at the lower (H7) weld location. The results of both transients followed the typical thermal-hydraulic behavior pattern expected during the transients; however, the extreme forces exerted on the core shroud could displace or rotate the shroud during these two postulated transients. Such shroud movement could impede control rod insertion and impair the safe shut down of the reactor. The TRAC-BF 1 pressure response and the ANSYS lift forces demonstrated that when IGSCC is present at the upper H3 core shroud weld, a lift force of 95 925 Pa (13.91 psi) is generated during the double MSLB transient which is great enough to overcome the weight of the vessel internals resting on the core shroud, causing the core shroud to lift. Additionally, the finite element ANSYS stress analysis demonstrated that when IGSCC is not present in the upper core shroud weld, stresses of 1757.3 MPa (254 880.0 psi) are induced in the filler material during a main steam line break. This is great enough to overcome the ultimate strength of 1723.7 Ma (250 000 psi) of the filler material and cause failure of the weld. The TRAC-BF1 RLB transient results determined that during the RLB event a force of 401.7 kN (90 306 lbf

  12. On some problems in pressure vessel analysis

    International Nuclear Information System (INIS)

    Throughout the world the use of pressure vessels has expanded considerably in the last decade. Many different types of vessels and components are used for processing and storage in the chemical and petroleum industries, nuclear and fossil power plants, fertilizer industry, and domestic applications, among others. One of the commonly used pressure vessels is the saddle supported horizontal vessel. When designing such vessels, one of the critical areas to be analyzed for stress is the saddle-shell interface region. A method frequently used by design engineers to accomplish this is that developed by Zick. The authors analysis is based upon the assumption that the support is rigid and not fastened to the vessel. Further, the author uses simple bending theory to calculate the stresses and deflections in the vessel. However, because a large number of the pressure vessels constructed today must be designed to withstand seismic loads, a more accurate analysis is essential. A literature search has revealed that there are no modified design procedures analogous to Zick's analysis which can rapidly predict the stresses at the saddle-shell interface as well as those at the mid-span of the vessel. The ASME Boiler and Pressure Vessel Code, Section VIII, Division 1, gives approximate formulas (based on Zick's analysis), charts and graphs so that the design of saddle supported vessels can be carried out. Analytical results presented by various authors for a particular geometry and support of vessel vary widely

  13. Assessment with coupled thermo-mechanical creep analysis of combined CRGT and external vessel cooling efficiency for a BWR

    International Nuclear Information System (INIS)

    In this paper we consider in-vessel stage of a severe core melt accident in a Nordic design Boiling Water Reactor (BWR). Decay-heated pool of corium melt inflicts thermal and mechanical loads on the lower-head vessel wall. Performed thermo-mechanical creep analysis identified two different modes of vessel wall failure: (i) a ‘ballooning’ of the vessel bottom and (ii) a ‘localized creep’ concentrated within the vicinity of the top surface of the melt pool. Next, given the mechanical and thermal loads from the decay-heated melt, external vessel cooling is applied at a specified time. It is found that combined CRGT and external vessel cooling was able to suppress the creep and subsequently prevent vessel wall failure. (author)

  14. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

  15. Materials for pressure vessels and the strength

    International Nuclear Information System (INIS)

    The viewpoint about safety engineering in design, the selection of materials and the effect of environment on materials are studied, taking pressure vessels as the example. The pressure vessels are defined in the enforcement ordinance of the labor safety and hygiene low, and prescribed in the high pressure gas control low and others. The stress of cylindrical and spherical vessels due to internal or external pressure and the stress concentration at flanges and nozzle portions are explained. The structural materials for pressure vessels are different according to gas or liquid contained, temperature and pressure. In case of the pressure vessels for nuclear reactors, the change of material properties due to neutron irradiation must be considered especially. Usually low alloy steels such as A 302B or A 533B lined with austenitic stainless steel by build-up welding are used. The effect of neutron irradiation is monitored with surveillance test pieces. The inspection of pressure vessels is carried out during the manufacture and in service. The pressure boundary for reactor coolant must be designed so as to endure the leak test and hydrostatic pressure test carried out during the life term of nuclear power generation plants. The statistics on the damage of pressure vessels in USA and UK are given. (Kako, I.)

  16. Wrapped Wire Detects Rupture Of Pressure Vessel

    Science.gov (United States)

    Hunt, James B.

    1990-01-01

    Simple, inexpensive technique helps protect against damage caused by continuing operation of equipment after rupture or burnout of pressure vessel. Wire wrapped over area on outside of vessel where breakthrough most likely. If wall breaks or burns, so does wire. Current passing through wire ceases, triggering cutoff mechanism stopping flow in vessel to prevent further damage. Applied in other situations in which pipes or vessels fail due to overpressure, overheating, or corrosion.

  17. Computerized reactor pressure vessel materials information system

    International Nuclear Information System (INIS)

    A computerized information system for storage and retrieval of reactor pressure vessel materials data was established, as part of Task Action Plan A-11, Reactor Vessel Materials Toughness. Data stored in the system are necessary for evaluating the resistance of reactor pressure vessels to flaw-induced fracture. This report includes (1) a description of the information system; (2) guidance on accessing the system; and (3) a user's manual for the system

  18. Reactor pressure vessel. Status report

    International Nuclear Information System (INIS)

    This report describes the issues raised as a result of the staffs review of Generic Letter (GL) 92-01, Revision 1, responses and plant-specific reactor pressure vessel (RPV) assessments and the actions taken or work in progress to address these issues. In addition, the report describes actions taken by the staff and the nuclear industry to develop a thermal annealing process for use at U.S. commercial nuclear power plants. This process is intended to be used as a means of mitigating the effects of neutron radiation on the fracture toughness of RPV materials. The Nuclear Regulatory Commission (NRC) issued GL 92-01, Revision 1, Supplement 1, to obtain information needed to assess compliance with regulatory requirements and licensee commitments regarding RPV integrity. GL 92-01, Revision 1, Supplement 1, was issued as a result of generic issues that were raised in the NRC staff's reviews of licensee responses to GL 92-01, Revision 1, and plant-specific RPV evaluations. In particular, an integrated review of all data submitted in response to GL 92-01, Revision 1, indicated that licensees may not have considered all relevant data in their RPV assessments. This report is representative of submittals to and evaluations by the staff as of September 30, 1996. An update of this report will be issued at a later date

  19. Ultrasound periodic inspections of reactor pressure vessels

    International Nuclear Information System (INIS)

    Two versions are described of ultrasonic equipment for periodic inspections of reactor pressure vessels. One uses the principle of exchangeable programmators with solid-state logic while the other uses programmable logic with semiconductor memories. The equipment is to be used for inspections of welded joints on the upper part of the V-1 reactor pressure vessel. (L.O.)

  20. Parametric studies on containment thermal hydraulic loads during high pressure melt ejection in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Silde, A.; Lindholm, I. [VTT Energy, Espoo (Finland)

    1997-12-01

    The containment thermal hydraulic loads during high pressure melt ejection in a Nordic BWR are studied parametrically with the CONTAIN and the MELCOR codes. The work is part of the Nordic RAK-2 project. The containment analyses were divided into two categories according to composition of the discharged debris: metallic and oxidic debris cases. In the base case with highly metallic debris, all sources from the reactor coolant system to the containment were based on the MELCOR/BH calculation. In the base case with the oxidic debris, the source data was specified assuming that {approx} 15% of the whole core material inventory and 34,000 kg of saturated water was discharged from the reactor pressure vessel (RPV) during 30 seconds. In this case, the debris consisted mostly of oxides. The highest predicted containment pressure peaks were about 8.5 bar. In the scenarios with highly metallic debris source, very high gas temperature of about 1900 K was predicted in the pedestal, and about 1400 K in the upper drywell. The calculations with metallic debris were sensititive to model parameters, like the particle size and the parameters, which control the chemical reaction kinetics. In the scenarios with oxidic debris source, the predicted pressure peaks were comparable to the cases with the metallic debris source. The maximum gas temperatures (about 450-500 K) in the containment were, however, significantly lower than in the respective metallic debris case. The temperatures were also insensitive to parametric variations. In addition, one analysis was performed with the MELCOR code for benchmarking of the MELCOR capabilities against the more detailed CONTAIN code. The calculations showed that leak tightness of the containment penetrations could be jeopardized due to high temperature loads, if a high pressure melt ejection occurred during a severe accident. Another consequence would be an early containment venting. (au). 28 refs.

  1. Containment failure time and mode for a low-pressure short-term station blackout in a BWR-4 with Mark-I containment

    International Nuclear Information System (INIS)

    This study investigates containment failure time and mode for a low-pressure, short-term station blackout severe accident sequence in a boiling water reactor (BWR-4) with a Mark-I containment. The severe accident analysis code MELCOR, version 1.8.1, was used in these calculations. Other results using the MELCOR/CORBH package and the BWRSAR and CONTAIN codes are also presented and compared to the MELCOR results. The plant analyzed is the Peach Bottom atomic station, a BWR-4 with a Mark-I containment. The automatic depressurization system was used to depressurize the vessel in accordance with the Emergency Procedure Guidelines. Two different variations of the station blackout were studied: one with a dry cavity and the other with a flooded cavity. For the flooded cavity, it is assumed that a control rod drive (CRD) pump becomes operational after vessel failure, and it is used to pump water into the cavity

  2. 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 GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) SMALL PASSENGER VESSELS CARRYING MORE THAN 150 PASSENGERS OR WITH OVERNIGHT ACCOMMODATIONS FOR MORE THAN 49 PASSENGERS MACHINERY INSTALLATION...

  3. Ten year's experience of in-service inspection on BWR vessel stub tubes

    International Nuclear Information System (INIS)

    The stub tube is a component of the control rod housings in boiling water reactor (BWR) nuclear power plants. In certain cases these tubes may undergo cracking, as a result of which fluid leakage may occur from the reactor vessel. Consequently, these components have to be inspected during service in order to determine whether or not they are affected by such defects. The stainless steel/Inconel stub tubes are welded at the upper end to the control rod housing, and at the lower end to the reactor vessel. Given the geometry, material, welds, stresses and corrosive elements associated with these components, intergranular corrosion cracking may occur in the areas adjacent to the welds. For this reason inspections capable of detecting this type of defect must be performed, with a view to determining the integrity of the component. Since 1981, more than 300 stub tube inspections have been carried out at different Spanish nuclear power plants. Initially, a single ultrasonic technique was used to detect the presence of indications; at present, and after several intermediate stages, various ultrasonic and eddy current techniques are used to dimension the length and depth of indications, determine their evolution and ensure dimensional control of the component for subsequent repair. Parallel to the development of non destructive testing techniques, mechanical scanning equipment has been designed and manufactured for use in test performance. Throughout development of these techniques, and prior to application in the field, different validation tests have been performed, initially using blocks containing artificial reflectors and subsequently blocks with actual crack-type reflectors. (author)

  4. 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...

  5. Technical report on material selection and processing guidelines for BWR coolant pressure boundary piping

    International Nuclear Information System (INIS)

    This report sets forth the NRC staff's revised acceptable methods to reduce the intergranular stress corrosion cracking susceptibility of BWR ASME Code Class 1 and 2 pressure boundary piping and safe end. For plants that cannot fully comply with the material selection, testing, and processing guidelines of this document, varying degrees of augmented inservice inspection and leak detection requirements are presented

  6. Neutron flux calculation and fluence in the encircling of the core and vessel of a reactor BWR

    International Nuclear Information System (INIS)

    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-θ 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-θ, 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, θ 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/cm2s, at a height H 4 (239.07 cm) and angle 32.236o in the core shroud and 4.00 E + 09 n/cm2s at a height H 4 and angle 35.27o in the inner wall of the reactor vessel, positions that are consistent to within ±10% over the ones reported in the literature. (Author)

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

    OpenAIRE

    Javier Ortiz-Villafuerte; Andrés Rodríguez-Hernández; Enrique Araiza-Martínez; Luis Fuentes-Márquez; Jorge Viais-Juárez

    2013-01-01

    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 liqui...

  8. Rupture safety for nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    This safety device for pressure vessels with cooling water inlet and outlet supports is to catch flying parts in the case of a rupture and to convert their energy into forming work without rupturing itself. A steel cladding is used for this purpose which surrounds the lower part of the pressure vessel as protective container like a pot, right up to the coolant supports. In order to keep the possible acceleration paths of the rupture pieces as small as possible, the gap between the two vessels is limited to a small tolerance to compensate for the differences in thermal expansion. In order to take up the compensating pressure after rupture of the reactor pressure vessel, the cylindrical part of the protective vessel requires double wall thickness like the concave floor. The cylindrical part is thus cladded by a cylindrical protective ring of the same wall thickness or by a multilayer sheet cladding which can be easily fabricated, transported and assembled. The cylindrical and concave part of the inner protective vessel have uniform wall thickness. This hence ensures that the elastic form changing is uniformly distributed over the height of the protective vessel. (HP)

  9. Depressurization device for reactor pressure vessel

    International Nuclear Information System (INIS)

    An electromotive valve is interposed to each of a plurality of depressurization pipelines branched from a main steam pipe, which is connected to a reactor pressure vessel. A rupture disk is disposed at the opening end of the depressurization pipeline at the downstream of the electromotive valve. A vent pipeline is disposed between the electromotive valve and the rupture disk, while being branched from the depressurization pipeline, for introducing leaked steams to a suppression pool. Then, a thermometer is disposed in the midway of the vent pipeline. If the electromotive valve is opened, the rupture disk is broken by high pressure steams flowing from the main steam pipe. With such procedures, steams in the reactor pressure vessel are released into the reactor container to depressurize the reactor pressure vessel. Further, the leakage of the electromotive valve during normal operation is easily detected by the thermometer disposed in the midway of the vent tube based on the temperature of the steams. (I.N.)

  10. Evaluation of pressure transitories in BWR type reactors using the BWRDYN code; Evaluacion de transitorios de presion en reactores tipo BWR usando el codigo BWRDYN

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez P, J.A. [ESIME, Unidad Profesional Azcapotzalco, Av. de las Granjas 682, 02550 Mexico D.F. (Mexico)]. e-mail: jrodriguez@ipn.mx

    2007-07-01

    Several simulations of pressure transitory for a nucleo electric power station with BWR/4 type reactor were carried out. The simulated pressure transitories were made for the Peach Bottom 2 Nucleo electric central. Also, it was carried out for the same Plant the simulation of the turbine shot with derivation to the main condenser, of the reference case (benchmark) outlined by the Organization for the Cooperation and the Economic Development and of the Commission Regulatory in Nuclear matter of the United States of America. As tool to carry out the simulations of the transitory ones, the BWRDYN code developed by the Japan Energy Research Institute was used. Among the main suppositions and models that it includes the BWRDYN code its can be mentioned: a) that of punctual kinetics that calculates the neutron flow; for the calculation of the fuel temperature, this it is divided in nodes in the radial and axial directions, the wrapper is considered like a region in the radial direction; c) the pressure is supposed that it is uniform inside the reactor vessel; and d) the thermal hydraulic pattern of the reactor vessel is divided in five regions and the core is divided in several nodes to take into account the distribution of holes in the axial direction. The modeling of the control systems of the feeding water system is also included, of the pressure regulator and of the recirculation system. The systems of what is known as plant balance are also modeled. The numeric results of the simulations provide valuable information of the behavior of the nucleo electric central. The obtained results of the simulation of the reference case agree acceptably with the measurements data, when comparing them with the measurements made in the Peach Bottom 2 Central. The obtained results of each simulation are fundamental to evaluate the transitory one, as well as to delineate the sequence and the impact of diverse events that they happen during the same one transitory. In the case of the

  11. Composite pressure vessels for commercial applications

    OpenAIRE

    Nunes, J. P.; Oliveira, Luís; J. F. Silva; Barros, B.; Amorim, Luís Manuel Machado de; Vasconcelos, M

    2013-01-01

    A new generation of composite pressure vessels for large scale market applications has been studied in this work. The vessels consist on a plastic liner wrapped with a filament winding glass fibre reinforced polymer matrix structure. A polyethylene (PE) was selected as liner for water at room temperatures applications and a thermosetting resin was used as matrices in the glass reinforced filament wound laminate. For applications having higher service temperatures, such as, termal accumulat...

  12. 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.

  13. Mark I 1/12-scale pressure suppression pool swell tests. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    Torbeck, J.E.; Galyardt, D.L.; Walker, J.P.

    1976-05-01

    A second series of 1/12-scale tests of the Mark I BWR containment have been run to define the torus and ring header loads resulting from pressure suppression pool swell following a postulated LOCA. These tests were conducted following modification of the test section used for earlier tests to tighten the sealing and improve control of the test conditions. Forty-two tests were performed, investigating the sensitivity of the pool response to drywell pressurization rate, initial downcomer submergence, initial system pressure and initial drywell/wetwell pressure differential, covering the range of scaled Mark I expected conditions (on the basis of the FSAR).

  14. Internal Friction of Pressure Vessel Steel Embrittlement

    International Nuclear Information System (INIS)

    The contribution consists of an abstract of a PhD thesis. The thesis contains a literature study, a description of the construction details of a new inverted torsion pendulum. This device was designed to investigate pressure-vessel steels at high amplitudes (10-4 to 10-2) and over a wide temperature range (90-700K) at approximately 1 Hz in the irradiated condition. Results of measurements on a variety of reactor pressure vessel steels by means of the torsion penduli are reported and interpreted

  15. Proactive life extension of pressure vessels

    Science.gov (United States)

    Mager, Lloyd

    1998-03-01

    For a company to maintain its competitive edge in today's global market every opportunity to gain an advantage must be exploited. Many companies are strategically focusing on improved utilization of existing equipment as well as regulatory compliance. Abbott Laboratories is no exception. Pharmaceutical companies such as Abbott Laboratories realize that reliability and availability of their production equipment is critical to be successful and competitive. Abbott Laboratories, like many of our competitors, is working to improve safety, minimize downtime and maximize the productivity and efficiency of key production equipment such as the pressure vessels utilized in our processes. The correct strategy in obtaining these objectives is to perform meaningful inspection with prioritization based on hazard analysis and risk. The inspection data gathered in Abbott Laboratories pressure vessel program allows informed decisions leading to improved process control. The results of the program are reduced risks to the corporation and employees when operating pressure retaining equipment. Accurate and meaningful inspection methods become the cornerstone of a program allowing proper preventative maintenance actions to occur. Successful preventative/predictive maintenance programs must utilize meaningful nondestructive evaluation techniques and inspection methods. Nondestructive examination methods require accurate useful tools that allow rapid inspection for the entire pressure vessel. Results from the examination must allow the owner to prove compliance of all applicable regulatory laws and codes. At Abbott Laboratories the use of advanced NDE techniques, primarily B-scan ultrasonics, has provided us with the proper tools allowing us to obtain our objectives. Abbott Laboratories uses B-scan ultrasonics utilizing a pulse echo pitch catch technique to provide essential data on our pressure vessels. Equipment downtime is reduced because the nondestructive examination usually takes

  16. Reactor vessel pressure transient protection for pressurized water reactors

    International Nuclear Information System (INIS)

    During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity

  17. PWR pressure vessel integrity during overcooling accidents

    International Nuclear Information System (INIS)

    Pressurized water reactors are susceptible to certain types of hypothetical accidents that under some circumstances, including operation of the reactor beyond a critical time in its life, could result in failure of the pressure vessel as a result of propagation of crack-like defects in the vessel wall. The accidents of concern are those that result in thermal shock to the vessel while the vessel is subjected to internal pressure. Such accidents, referred to as pressurized thermal shock or overcooling accidents (OCA), include a steamline break, small-break LOCA, turbine trip followed by stuck-open bypass valves, the 1978 Rancho Seco and the TMI accidents and many other postulated and actual accidents. The source of cold water for the thermal shock is either emergency core coolant or the normal primary-system coolant. ORNL performed fracture-mechanics calculations for a steamline break in 1978 and for a turbine-trip case in 1980 and concluded on the basis of the results that many more such calculations would be required. To meet the expected demand in a realistic way a computer code, OCA-I, was developed that accepts primary-system temperature and pressure transients as input and then performs one-dimensional thermal and stress analyses for the wall and a corresponding fracture-mechanics analysis for a long axial flaw. The code is briefly described, and its use in both generic and specific plant analyses is discussed

  18. Problems in Pressure Vessel Design and Manufacture

    International Nuclear Information System (INIS)

    The general desire by the power reactor process makers to increase power rating and their efforts to involve more advanced thermal behaviour and fuel handling facilities within the reactor vessels are accompanied by an increase in both pressure vessel dimensions and various difficulties in giving practical solutions of design materials and fabrication problems. In any section of this report it is emphasized that difficulties and problems already met with will meet again in the future vessels but then in modified forms and in many cases more pertinent than before. As for the increase in geometrical size it can be postulated that with use of better materials and adjusted fabrication methods the size problems can be taken proper care of. It seems likely that vessels of sufficient large diameter and height for the largest power output, which is judged as interesting in the next ten year period, can be built without developing totally new site fabrication technique. It is, however, supposed that such a fabrication technique will be feasible though at higher specific costs for the same quality requirements as obtained in shop fabrication. By the postulated use of more efficient vessel material with principally the same good features of easy fabrication in different stages such as preparation, welding, heat treatment etc as ordinary or slightly modified carbon steels the increase in wall thickness might be kept low. There exists, however, a development work to be done for low-alloy steels to prove their justified use in large reactor pressure vessels

  19. Pressurized wet digestion in open vessels.

    Science.gov (United States)

    Maichin, B; Zischka, M; Knapp, G

    2003-07-01

    The High Pressure Asher (HPA-S) was adapted with a Teflon liner for pressurized wet digestion in open vessels. The autoclave was partly filled with water containing 5% (vol/vol) hydrogen peroxide. The digestion vessels dipped partly into the water or were arranged on top of the water by means of a special rack made of titanium or PTFE-coated stainless steel. The HPA-S was closed and pressurized with nitrogen up to 100 bars. The maximum digestion temperature was 250 degrees C for PFA vessels and 270 degrees C for quartz vessels. Digestion vessels made of quartz or PFA-Teflon with volumes between 1.5 mL (auto sampler cups) and 50 mL were tested. The maximum sample amount for quartz vessels was 0.5-1.5 g and for PFA vessels 0.2-0.5 g, depending on the material. Higher sample intake may lead to fast reactions with losses of digestion solution. The samples were digested with 5 mL HNO(3) or with 2 mL HNO(3)+6 mL H(2)O+2 mL H(2)O(2). The total digestion time was 90-120 min and 30 min for cooling down to room temperature. Auto sampler cups made of PFA were used as digestion vessels for GFAAS. Sample material (50 mg) was digested with 0.2 mL HNO(3)+0.5 mL H(2)O+0.2 mL H(2)O(2). The analytical data of nine certified reference materials are also within the confidential intervals for volatile elements like mercury, selenium and arsenic. No cross contamination between the digestion vessels could be observed. Due to the high gas pressure, the diffusion rate of volatile species is low and losses of elements by volatilisation could be observed only with diluted nitric acid and vessels with large cross section. In addition, cocoa, walnuts, nicotinic acid, pumpkin seeds, lubrication oil, straw, polyethylene and coal were digested and the TOC values measured. The residual carbon content came to 0.2-10% depending on the sample matrix and amount. PMID:12802569

  20. BWR Mark I pressure suppression study: effect of downcomer fill level on the vertical load function

    International Nuclear Information System (INIS)

    The investigation reported forms a part of the BWR Mark I Pressure Suppression Experiment Program and is one of a series of small scale studies designed to evaluate limited aspects of the pool dynamics phenomena prior to conduct of the 1/5 scale air test series. Presented is an experimental study of the effect of downcomer fill level (DFL) on the vertical load function

  1. Local pressure testing of spherical vessels

    International Nuclear Information System (INIS)

    This work pertains to the numerical investigation of the feasibility of a proposed local pressure testing to verify structural integrity of nozzle-to-shell junctions in repaired/altered spherical pressure vessels. The “local pressure testing” involves the use of a small temporary testing closure at the nozzle-to-shell junction to get around the inconvenience of the conventional industry-wide pressure testing of the entire pressure vessel. However, it is essential in deciding on the reliability of such testing approach, to understand the influence of dimensional ratios between the nozzle, vessel and testing closure on achieving equivalent behavior, in terms of stresses and deformations, as compared to those associated with the full conventional testing. The paper presents the findings of a finite element study of the effect of cap size on the stresses near the junction of a cylindrical nozzle with a spherical vessel under internal pressure. The numerical model was verified by comparing its results to available analytical solutions of similar problems. The study focuses on the determination of the minimum required cap radius that will result in a local pressure testing that is equivalent to the conventional full pressure testing, mainly in terms of peak stresses at the junction. Results in the form of plots and empirical equations are presented for a parametric study covering a wide range of dimensions. The results show that the minimum required cap size is linearly related to the nozzle size, but also its value is usually much larger. This leads to the main conclusion that a reliable local pressure testing must use relatively large cap sizes, and that the caps with sizes slightly larger than those of the nozzle may not be acceptable. -- Highlights: • Local pressure testing of spherical vessels with nozzles is studied by the FEM. • It involves use of a “small” cylindrical cap wherein pressure testing is conducted. • It is established that the main

  2. Thermal ratchetting in pressure vessels

    International Nuclear Information System (INIS)

    During its lifetime a nuclear power plant undergoes cyclic thermal loadings, i.e. time variations of temperature spatial distribution; these thermal cycles combined with methanical loads (pressure), can produce dimensional changes for creep, plastic cycling and/or ratchetting. This problem is in special way important for a Liquid Metal Fast Breeder Reactor (LMFBR) because temperatures in it are high (about 5000C) and change either in radial direction or in axial direction. While the plastic behaviour of a structure subjected to mechanical loads and/or mechanical cycle loads can be well studied by Finite Element analyses and several different computer programs, the thermal plastic behaviour was studied for simplified conditions only. Actually, the Bree, Miller, Burgreen model considers a thin cylindrical shell with a radial thermal gradient and in uniaxial stress state while Kalnins, Updike extend this model to a biaxial stress state. To obtain numerical results of the thermo-plastic behaviour of cylindrical shells, a mathematical model was developed and the special purpose computer program PLAST was implemented. (Auth.)

  3. Proceedings of the international specialist meeting on BWR-pressure suppression containment technology. Vol. 2

    International Nuclear Information System (INIS)

    In the frame of R + D-work for BWR-pressure suppression systems the GKSS-Forschungszentrum Geesthacht GmbH organized an international specialist meeting. All important safety relevant aspects of pressure suppression system technology have been included. About 60 experts from USA, Japan, Sweden, Italy, Netherland and the Federal Republic of Germany participated. They came from licensing authorities, vendors, research centers and universities. In 24 papers they have shown the world-wide present status of theoretical and experimental know-how on pressure suppression system behaviour. In discussions and working groups recommendations for future work have been compiled. (orig.)

  4. Determination of K-factors for arbitrarily shaped flaws at pressure vessel nozzle corners

    International Nuclear Information System (INIS)

    Photoelastic and finite element studies are being conducted to determine Mode I stress intensity factor distributions along arbitrarily shaped flaw fronts at pressure vessel nozzle corners. Comparisons of results from NOZ-FLAW, BIGIF, and the photoelastic studies showed that (1) good agreement was obtained between NOZ-FLAW and the photoelastically determined K1's for the deep flaw in an ITV model, (2) good agreement was obtained between NOZ-FLAW BIGIF for shallow and moderately deep flaws in a BWR model, and (3) less satisfactory agreement was obtained between NOZ- FLAW and the photoelastic results for the BWR models, particularly for moderately deep to deep flaws. Attempts are presently being made at understanding and explaining the discrepancies between the two

  5. TAPS pressure vessel surveillance - results and evaluation

    International Nuclear Information System (INIS)

    SA302B (nickel modified) steel cladded with stainless steel is used as the pressure vessel material for the two 210 MWe boiling water reactors of the Tarapur Atomic Power Station. Charpy V-notch impact surveillance specimens representing the pressure vessel belt-line base, weld and the heat affected zone were irradiated at the wall and shroud locations. Some of these specimens were removed after 6.5 effective full power years (EFPY) of reactor operation. The neutron fluences at the locations were 5.31 x 1017 and 4.88 x 1018 n/cm2 (E > 1 MeV). The surveillance data generated from specimens removed after 6.5 EFPY were evaluated on the basis of USNRC Regulatory Guide 1.99, Revision 2, and the results had assured the integrity of the vessel beyond the end of design service life (EOL) of 40 years. The recent evaluation of the additional data generated from specimens removed after 13 EFPY has again confirmed the safety of the pressure vessel beyond EOL by an additional 20 EFPY. (author)

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

    International Nuclear Information System (INIS)

    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)

  7. 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.

  8. Pressure suppression pool mixing in passive advanced BWR plants

    International Nuclear Information System (INIS)

    In the SBWR passive boiling water reactor, the long-term post-accident containment pressure is determined by the combination of noncondensible gas pressure and steam pressure in the wetwell gas space. The suppression pool (SP) surface temperature, which determines the vapor partial pressure, is very important to overall containment performance. Therefore, the thermal stratification of the SP due to blowdown is of primary importance. This work looks at the various phases and phenomena present during the blowdown event and identifies those that are important to thermal stratification, and the scaling necessary to model them in reduced size tests. This is important in determining which of the large body of blowdown to SP data is adequate for application to the stratification problem. The mixing by jets from the main vents is identified as the key phenomena influencing the thermal response of the suppression pool and analytical models are developed to predict the jet influence on thermal stratification. The analytical models are implemented into a system simulation code, TRACG, and used to model thermal stratification behavior in a scaled test facility. The results show good general agreement with the test data

  9. Manipulator for pressure vessel inspection and repairs

    International Nuclear Information System (INIS)

    The manipulator body (a concrete container) consists of three sections. The height of the top, middle and bottom section is 3,780 mm, 3,900 mm, and 3,600 mm, respectively. The outside body diameter is 2,950 mm, the inner diameter is 2,000 mm. The top and middle sections are shielded by lead, the bottom section is shielded by a special concrete. For inspections and repairs of the reactor pressure vessel internals or of the reactor pit two intermediate floors and the floor of the bottom section are provided for the personnel. The manipulator is equipped with a central control and supervision panel located in the reactor hall. The communication with the personnel runs via the closed-circuit TV. The manipulator was proved during number of successful inspections of WWER-440 reactor pressure vessels carried out in Czechoslovak nuclear power plants. (J.B.). 1 fig

  10. Composite Pressure Vessel Including Crack Arresting Barrier

    Science.gov (United States)

    DeLay, Thomas K. (Inventor)

    2013-01-01

    A pressure vessel includes a ported fitting having an annular flange formed on an end thereof and a tank that envelopes the annular flange. A crack arresting barrier is bonded to and forming a lining of the tank within the outer surface thereof. The crack arresting barrier includes a cured resin having a post-curing ductility rating of at least approximately 60% through the cured resin, and further includes randomly-oriented fibers positioned in and throughout the cured resin.

  11. Pressurized thermal shock experiments with thick vessels

    International Nuclear Information System (INIS)

    The authors report a 148-mm-thick pressure vessel containing a long axial flaw subjected to a combined loading of internal pressure and thermal shock. The experiment was performed to investigate the effects of loading sequence on initiation of brittle fracture and the behavior of a crack propagating into ductile regions. Two crack initiation and arrest episodes were generated, in addition to several phases of warm prestressing. Warm prestressing was shown to be effective in inhibiting initiation of cleavage fracture. A crack arrest was observed 88 K above the reference nil-ductility transition temperature

  12. Noninvasive blood pressure measurement in large vessels

    International Nuclear Information System (INIS)

    Pulse pressure in the aorta was evaluated by the measurement of pulse wave velocity (PWV) and blood flow velocity (BFV). PWV reflects the elasticity of the vessel and was determined by a time-of-flight method. BFV was measured by analyzing the change of magnetization decay due to flow in multiecho experiments. If one neglects pulse wave reflections at vascular branch points and flow resistance due to blood viscosity, pulse pressure is proportional to PWV and BFV. Noninvasive MR imaging measurements were obtained in 12 patients, all of whom underwent correlative arterial catheterization. Values varied between 35 and 100 mm Hg. The results demonstrated a high correlation between the two methods

  13. Modeling Scala Media as a Pressure Vessel

    Science.gov (United States)

    Lepage, Eric; Olofsson, A.˚Ke

    2011-11-01

    The clinical condition known as endolymphatic hydrops is the swelling of scala media and may result in loss in hearing sensitivity consistent with other forms of low-frequency biasing. Because outer hair cells (OHCs) are displacement-sensitive and hearing levels tend to be preserved despite large changes in blood pressure and CSF pressure, it seems unlikely that the OHC respond passively to changes in static pressures in the chambers. This suggests the operation of a major feedback control loop which jointly regulates homeostasis and hearing sensitivity. Therefore the internal forces affecting the cochlear signal processing amplifier cannot be just motile responses. A complete account of the cochlear amplifier must include static pressures. To this end we have added a third, pressure vessel to our 1-D 140-segment, wave-digital filter active model of cochlear mechanics, incorporating the usual nonlinear forward transduction. In each segment the instantaneous pressure is the sum of acoustic pressure and global static pressure. The object of the model is to maintain stable OHC operating point despite any global rise in pressure in the third chamber. Such accumulated pressure is allowed to dissipate exponentially. In this first 3-chamber implementation we explore the possibility that acoustic pressures are rectified. The behavior of the model is critically dependent upon scaling factors and time-constants, yet by initial assumption, the pressure tends to accumulate in proportion to sound level. We further explore setting of the control parameters so that the accumulated pressure either stays within limits or may rise without bound.

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

    OpenAIRE

    C. M. Allison; J. K. Hohorst; B. S. Allison; D. Konjarek; Bajs, T.; R. Pericas; Reventos, F.; Lopez, R.

    2012-01-01

    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 re...

  15. Heating of reactor pressure vessel bottom head and penetrations in a severe reactor accident

    International Nuclear Information System (INIS)

    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.)

  16. 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...

  17. Technical report on material selection and processing guidelines for BWR coolant pressure boundary piping

    International Nuclear Information System (INIS)

    It is the purpose of this report to set forth acceptable methods to reduce the stress corrosion cracking susceptibility of BWR piping and thereby also provide an increased level of reactor coolant pressure boundary integrity. Because the most straightforward and desirable approach or methods may not be practicable, or even possible, for all plants, the bases for varying degrees of conformance to the guidelines are provided. Augmented inservice inspection and leak detection requirements are established for plants that have not fully implemented the provisions presented

  18. Recent experience reactor pressure vessel manufacture

    International Nuclear Information System (INIS)

    This paper present the Framatome's recent experience in the manufacture of 1300 MWe PWR vessels; one shows how the very high standards of quality have been obtained to meet the stringent requirements. After a description of a pressure vessel, materials and forgings properties are presented. The nature and sequence of the main fabrication operations are reviewed. This paper deals after with the quality of welds, the preheating and post-heating equipment, the submerged arc welding process and procedures, the cladding process, and the under-clad cracking problems. Ultrasonic inspection procedures of the main welds are described with a comparison of RCCM (design and construction rules for mechanical components of PWR units) and Sizewell B specifications. Support of data on the reproductibility and effectiveness of ultrasonic examination and on the reliability given by repetitive inspection are presented

  19. 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)

  20. Refinements to pressure vessel steel embrittlement correlations

    International Nuclear Information System (INIS)

    ASTM E 900-87, Standard Guide for Predicting Neutron Radiation Damage to Reactor Vessel Materials, provides methods for predicting the effects of fast neutron irradiation on the toughness of reactor pressure vessel materials. The prediction methodology is similar but separate for the steel base metal and weld metal. The methodology relies on a correlation between the chemical content of the steel and the fast neutron fluence. The prediction methodology is used to assess the margin of safety against fracture of the reactor vessel during normal operation and postulated transients. The purpose of this paper is to describe the results of an effort to improve the accuracy of the predictions of transition temperature shift. Numerous issues have been identified which could influence the calculational approach. This evaluation was performed using a scrubbed data base of reactor vessel surveillance data coupled with best-estimate irradiation temperatures; the latter were based on the reactor vessel cold leg temperatures rather than the go/no go thermal monitor data. A term for dependence on irradiation temperature was derived and added to the correlation for irradiation induced shift of the transition temperature. The ultimate objective is to contribute to the development of an improved understanding of the empirical relationships which, when coupled with theoretical models of irradiation damage mechanisms, will result in more accurate correlations for inclusion in ASTM E 900. The data base improvements, addition of a temperature term, consideration of saturation effects (with neutron fluence or copper content) and separation by material subsets are intended to assure that the best estimate predictions and their uncertainty are as realistic as possible

  1. Prestressed cast iron pressure vessels as burst-proof pressure vessels for innovative nuclear applications

    International Nuclear Information System (INIS)

    The amendment to the German Atomic Energy Act from July 28, 1994 requires that events 'whose occurrence is practically excluded by the measures against damages', i.e. events of the category residual risk, must not necessitate far reaching protective measures outside the plant. For a conventional reactor pressure vessel, the residual risk consists in the very small probability of a catastrophic failure (formation of a large fracture opening, bursting of the vessel). With a prestressed cast iron vessel (PCIV), the formation of a large fracture opening or bursting of the vessel, respectively, is impossible due to its design properties. Against this background the possibility of the use of this type of pressure vessel for lightwater reactors has been studied in the frame of a 'Working Group for Innovative Nuclear Technology', founded by different research institutes and industrial companies. Furthermore, it has been studied whether the use of the PCIV support the realization of a corecatcher system. The results are presented in this report. Already many years earlier, Siempelkamp has performed industrial development and Forschungszentrum Juelich related experimental and theoretical safety research for the PCIV as an innovative, bust-proof pressure vessel concept. This development of the PCIV as well as its safety properties are also presented in a conclusive manner. (orig.)

  2. Development of new irradiation facility for BWR safety research

    International Nuclear Information System (INIS)

    In JAEA (Japan Atomic Energy Agency), about the irradiation embrittlement of the reactor pressure vessel and the stress corrosion cracking of reactor core composition apparatus concerning the long-term use of the light water reactor (BWR), in order to check the influence of the temperature, pressure, and water quality, etc on BWR condition. The water environmental control facility which performs irradiation assisted stress corrosion-cracking (IASCC) evaluation under BWR irradiation environment was fabricated in JMTR (Japan Materials Testing Reactor). This report is described the outline of manufacture of the water environmental control facility for doing an irradiation test using the saturation temperature capsule after JMTR re-operation. (author)

  3. Control Rod Drive Mechanism Installed in the Internal of Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    This report describes the review results and important technologies related to the in-vessel type control rod drive mechanism. Generally, most of the CRDMs used in the PWR are attached outside of the reactor pressure vessel, and the pernetration of the vessel head can not avoid. However, in-vessel type CRDMs, which are installed inside the reactor vessel, can eliminate the possibility of rod ejection accidents and the penetration of the vessel head, and provide a compact design of the reactor vessel and containment. There are two kinds of in-vessel type CRDM concerning the driving force-driven by a driving motor and by a hydraulic force. Motor driven CRDMs have been mainly investigated in Japan(MRX, IMR, DRX, next generation BWR etc.), and developed the key components such as a canned motor, an integrated rod position indicator, a separating ball-nut and a ball bearing that can operate under the water conditions of a high temperature and pressure. The concept of hydraulically driven CRDMs have been first reported by KWU and Siemens for KWU 200 reactor, and Argentina(CAREM) and China(NHR-5, NHR-200) have been developed the internal CRDM with the piston and cylinder of slightly different geometries. These systems are driven by the hydraulic force which is produced by pumps outside of the reactor vessel and transmitted through a pipe penetrating the reactor vessel, and needs complicated control and piping systems including pumps, valves and pipes etc.. IRIS has been recently decided the internal CRDMs as the reference design, and an analytical and experimental investigations of the hydraulic drive concept are performed by POLIMI in Italy. Also, a small French company, MP98 has been developed a new type of control rods, called 'liquid control rods', where reactivity is controlled by the movement of a liquid absorber in a manometer type device

  4. Control Rod Drive Mechanism Installed in the Internal of Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, M. H.; Choi, S.; Park, J. S.; Lee, J. S.; Kim, D. O.; Hur, N. S.; Hur, H.; Yu, J. Y

    2008-09-15

    This report describes the review results and important technologies related to the in-vessel type control rod drive mechanism. Generally, most of the CRDMs used in the PWR are attached outside of the reactor pressure vessel, and the pernetration of the vessel head can not avoid. However, in-vessel type CRDMs, which are installed inside the reactor vessel, can eliminate the possibility of rod ejection accidents and the penetration of the vessel head, and provide a compact design of the reactor vessel and containment. There are two kinds of in-vessel type CRDM concerning the driving force-driven by a driving motor and by a hydraulic force. Motor driven CRDMs have been mainly investigated in Japan(MRX, IMR, DRX, next generation BWR etc.), and developed the key components such as a canned motor, an integrated rod position indicator, a separating ball-nut and a ball bearing that can operate under the water conditions of a high temperature and pressure. The concept of hydraulically driven CRDMs have been first reported by KWU and Siemens for KWU 200 reactor, and Argentina(CAREM) and China(NHR-5, NHR-200) have been developed the internal CRDM with the piston and cylinder of slightly different geometries. These systems are driven by the hydraulic force which is produced by pumps outside of the reactor vessel and transmitted through a pipe penetrating the reactor vessel, and needs complicated control and piping systems including pumps, valves and pipes etc.. IRIS has been recently decided the internal CRDMs as the reference design, and an analytical and experimental investigations of the hydraulic drive concept are performed by POLIMI in Italy. Also, a small French company, MP98 has been developed a new type of control rods, called 'liquid control rods', where reactivity is controlled by the movement of a liquid absorber in a manometer type device.

  5. Design of pressure vessels. Part 1

    International Nuclear Information System (INIS)

    The equipments and loops of PWR reactors are basically pressure vessels. Their specificities concern the integrity warranties that must be implemented considering their importance for the reactors safety. Thus, stress is put on the exhaustiveness of the prevention of in-service degradation and on the safety scenarios considered. The second specificity concerns the possibility of activation of wear and corrosion products during their flow inside the reactor core. This second aspect leads to some constraints on the choice of the materials used and on the surface coating of the inside wall of big components of the primary circuit. The aim of this document is to develop the general approach adopted for the design of the pressure vessels of PWR fluid loops, and to stress more particularly on the nuclear particularities of these equipments. Some extensions of these rules to high temperature resistant materials (FBR-type reactors) are also evoked. Content: General considerations: design basis of pressure vessels, risk analysis and design conditions, ruining paths and safety coefficients; 2 - damage prevention for excessive deformation: definitions, criteria; 3 - prevention of the plastic instability damage: definition, criteria; 4 - buckling prevention: definition and mechanisms, rules and criteria; 5 - prevention of progressive deformation damage: definitions, plastic adaptation, plastic accommodation, progressive deformation; 6 - prevention of fatigue damage: definitions, general prevention approach, design fatigue curves, analytic approach, particular aspects, analysis of zones with geometrical singularity; 7 - prevention of sudden rupture damage: fragile rupture and ductile tear, general approach, analytic criteria, irradiation and aging effects; 8 - other potential damages; 9 - conclusion. (J.S.)

  6. Dictionary of pressure vessel and piping technology

    International Nuclear Information System (INIS)

    A specialised dictionary has been compiled containing the appropriate English and German terms in the following technical fields: materials science, welding, destructive and non-destructive testing, thermal and mass transfer, the design and construction in particular of pressure vessels, tanks, heat exchangers, piping, expansion joints, valves, and components associated with the above fields. This dictionary is the result of many years spent in evaluating technical terminology from the relevant American and British regulations, technical rules, standards, and specifications (see bibliography) and correlating these with the terminology of comparable German regulations, rules and standards, together with the essential technical literature. (orig.)

  7. Device for extending the life of a pressure vessel, particularly a reactor pressure vessel

    International Nuclear Information System (INIS)

    It is proposed to mount a heating device firmly in the annular space between the outer jacket of the reactor pressure vessel and its insulation, which will from time temper the particularly endangered parts of the wall of the pressure vessel and therefore increase their life expectancy. Some constructional details are given in the proposal. Control will be by temperature measuring devices (thermocouples), which can control the heat output. Induction coils, heating resistors, heat radiators and also hot fluids (i.e. gases or liquids) are mentioned as possible heating elements. (UWI) 891 HP/UWI 892 MB

  8. Dismantling id the reactor pressure vessel insulation and dissecting of the MZFR reactor pressure vessel

    International Nuclear Information System (INIS)

    The MZFR reactor was decommissioned in 1984. The authors describe the dismantling of the reactor pressure vessel insulation that consists of asbestos containing mineral fiber wool. The appropriate remote handling and cutting tools had to be adapted with respect to the restrained space in the containment. The dismantling of the reactor pressure vessel has been completed, the dissected parts have been packaged into 200 containers for the final repository Konrad. During the total project time no reportable events and no damage to persons occurred.

  9. On core debris behaviour in the pressure vessel lower head of nordic boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lindholm, I.; Ikonen, K. [VTT Energy, Espoo (Finland); Hedberg, K. [Vattenfall Energistystem AB, Stockholm (Sweden); Thomsen, K. [Risoe National Lab., Roskilde (Denmark)

    1997-10-01

    In-vessel melt progression in Nordic BWRs has been studied as part of the RAK-2 project within the Nordic Nuclear Safety Programme 1994-1997. A part of the study was the evaluation of the late phase melt progression phenomena and the thermal behaviour of core debris, the pressure vessel wall and the lower head penetrations during a severe accident. The investigations presented here focus on BWR cases. The MELCOR/Bottom Head Package was applied to investigate the core debris bed behaviour and thermal response of structures in the case of the Olkiluoto 1 and 2 reactor vessel lower head. Both low and high pressure scenarious were analysed with sensitivity studies addressing the effects of debris bed porosity, debris particle size and reflooding of the dry debris bed. Lower head failure mechanisms and timing were examined by allowing instrument tube failure (normal case) or by deactivating the penetration failure model with an input option. Due to modelling assumptions in MELCOR, all presented calculations examine thermal behaviour of a rubble bed in the lower head. Calculated results are evaluated against experimental data. Studies using Forsmark 3 input data were carried out with the MAAP4 code. Studied cases covered also low and high pressure sequences, and a number of sensitivity calculations varying a few key parameters were performed. Only creep rupture of the reactor pressure vessel (RPV) was considered in the MAAP4 analyses. 33 refs.

  10. On core debris behaviour in the pressure vessel lower head of nordic boiling water reactors

    International Nuclear Information System (INIS)

    In-vessel melt progression in Nordic BWRs has been studied as part of the RAK-2 project within the Nordic Nuclear Safety Programme 1994-1997. A part of the study was the evaluation of the late phase melt progression phenomena and the thermal behaviour of core debris, the pressure vessel wall and the lower head penetrations during a severe accident. The investigations presented here focus on BWR cases. The MELCOR/Bottom Head Package was applied to investigate the core debris bed behaviour and thermal response of structures in the case of the Olkiluoto 1 and 2 reactor vessel lower head. Both low and high pressure scenarious were analysed with sensitivity studies addressing the effects of debris bed porosity, debris particle size and reflooding of the dry debris bed. Lower head failure mechanisms and timing were examined by allowing instrument tube failure (normal case) or by deactivating the penetration failure model with an input option. Due to modelling assumptions in MELCOR, all presented calculations examine thermal behaviour of a rubble bed in the lower head. Calculated results are evaluated against experimental data. Studies using Forsmark 3 input data were carried out with the MAAP4 code. Studied cases covered also low and high pressure sequences, and a number of sensitivity calculations varying a few key parameters were performed. Only creep rupture of the reactor pressure vessel (RPV) was considered in the MAAP4 analyses

  11. Neural Network Burst Pressure Prediction in Composite Overwrapped Pressure Vessels

    Science.gov (United States)

    Hill, Eric v. K.; Dion, Seth-Andrew T.; Karl, Justin O.; Spivey, Nicholas S.; Walker, James L., II

    2007-01-01

    Acoustic emission data were collected during the hydroburst testing of eleven 15 inch diameter filament wound composite overwrapped pressure vessels. A neural network burst pressure prediction was generated from the resulting AE amplitude data. The bottles shared commonality of graphite fiber, epoxy resin, and cure time. Individual bottles varied by cure mode (rotisserie versus static oven curing), types of inflicted damage, temperature of the pressurant, and pressurization scheme. Three categorical variables were selected to represent undamaged bottles, impact damaged bottles, and bottles with lacerated hoop fibers. This categorization along with the removal of the AE data from the disbonding noise between the aluminum liner and the composite overwrap allowed the prediction of burst pressures in all three sets of bottles using a single backpropagation neural network. Here the worst case error was 3.38 percent.

  12. Cast support for a pressure vessel at high internal pressure

    International Nuclear Information System (INIS)

    The invention concerns a cast support for a pressure vessel, which is sealed with a liner. In order to prevent bulging of the liner which is not fixed to the cast support, a supporting intermediate layer is situated between the liner and the cast support. The cast support is only anchored in the supporting intermediate layer in certain areas. Several types of anchoring are described for the cast support. (orig.)

  13. Predicting crack arrest in reactor pressure vessels

    International Nuclear Information System (INIS)

    The pressurized thermal shock (PTS) issue has provided increased motivation for the search for a reasonably accurate crack arrest prediction methodology. This issue has assumed greater significance recently as a consequence of the imposition of Regulatory Guide 1.99 Revision 2 procedures for determining the effects of radiation embrittlement in the context of the screening criteria in the PTS rule that is used by the United States Nuclear Regulatory Commission to assess the integrity of reactor pressure vessels. The currently accepted procedure for predicting crack arrest is the so-called KIa procedure, which is based on static linear elastic fracture mechanics principles, with a crack being presumed to arrest when the crack tip stress intensity factor KIST falls below a value KIa. The present paper reviews recent EPRI sponsored research, which shows that the static procedure is overly conservative when it is applied to the first arrest of a deep crack in the thickness of a reactor vessel. This conclusion is clearly important when assessing the consequences of the imposition of the procedures of Regulatory Guide 1.99 Revision 2. A more accurate crack arrest prediction procedure, i.e. the Combustion Engineering constrained static procedure or the reflectionless stress intensity factor procedure which are very similar in concept and their arrest prediction, should be considered to assess the impact of its use in the context of the screening criteria limits in the PTS rule. (orig.)

  14. 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.

  15. Low cycle fatigue behavior of pressure vessel steels in high temperature pressurized water

    International Nuclear Information System (INIS)

    Low cycle fatigue behavior of low alloy steels ASTM A508 Cl.3(JIS SFVQ1A) and ASTM A533B Cl.1(JIS SQV2A) for nuclear reactor pressure vessels was investigated in high temperature pressurized water simulating BWR coolant environments. Total strain range, strain rate and dissolved oxygen concentration were varied from 0.5 to 2.2 %, 0.1 to 0.001 %/s and 10 to 8 000 ppb, respectively. Fatigue tests in ambient air and 561 K air were also conducted for comparison. It was found that fatigue lives in high temperature water were shorter than those in ambient air. However, the reduction of fatigue life decreased with decreasing total strain range and rather longer fatigue lives than those in ambient air were observed at lower total strain range. A533B material showed the distinct strain rate dependence of fatigue life compared with A508 material, while they showed the similar dependence on dissolved oxygen concentration. It was found that fatigue cracks initiated at corrosion pits generated by dissolution of MnS inclusions and the low cycle fatigue behavior depended on sulfur content of the material. It can be concluded that the materials tested possess safety margins in reactor coolant environments by judging from the fact that all the present data fell on a region above the design fatigue curves in the ASME Code Sec. III. (author)

  16. Pressure Tube and Pressure Vessel Reactors; certain comparisons

    International Nuclear Information System (INIS)

    In a comparison between pressure tube and pressure vessel type reactors for pressurized D2O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D2O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960

  17. Thermal hydraulics characterization of the core and the reactor vessel type BWR

    International Nuclear Information System (INIS)

    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)

  18. Numerical simulations of pressure fluctuations at branch piping in BWR main steam line

    International Nuclear Information System (INIS)

    The power uprating of a nuclear power plant may increase/accelerate degradation phenomena such as flow-induced vibration and wall thinking. A steam dryer was damaged by a high cycle fatigue due to an acoustic-induced vibration at the branch piping of safety relief valves (SRVs) in main steam lines. In this study, we conducted the numerical simulations of steam/air flow around a simplified branch piping to clarify the basic characteristics of resonance. LES simulations were conducted in ordinary pressure/temperature air and steam under BWR plant conditions. In both cases, the excitation of the pressure fluctuations at the branch was observed under some inlet velocity conditions. These fluctuations and inlet conditions were normalized and the obtained results were compared. The normalized results showed that the range and maximum amplitude of pressure fluctuations were almost the same in low-pressure/temperature air and high-pressure/temperature steam. We found that ordinary pressure/temperature air experiments and simulations can possibly clarify the characteristics of the resonance in high-pressure/temperature steam. (author)

  19. Asymmetric Bulkheads for Cylindrical Pressure Vessels

    Science.gov (United States)

    Ford, Donald B.

    2007-01-01

    Asymmetric bulkheads are proposed for the ends of vertically oriented cylindrical pressure vessels. These bulkheads, which would feature both convex and concave contours, would offer advantages over purely convex, purely concave, and flat bulkheads (see figure). Intended originally to be applied to large tanks that hold propellant liquids for launching spacecraft, the asymmetric-bulkhead concept may also be attractive for terrestrial pressure vessels for which there are requirements to maximize volumetric and mass efficiencies. A description of the relative advantages and disadvantages of prior symmetric bulkhead configurations is prerequisite to understanding the advantages of the proposed asymmetric configuration: In order to obtain adequate strength, flat bulkheads must be made thicker, relative to concave and convex bulkheads; the difference in thickness is such that, other things being equal, pressure vessels with flat bulkheads must be made heavier than ones with concave or convex bulkheads. Convex bulkhead designs increase overall tank lengths, thereby necessitating additional supporting structure for keeping tanks vertical. Concave bulkhead configurations increase tank lengths and detract from volumetric efficiency, even though they do not necessitate additional supporting structure. The shape of a bulkhead affects the proportion of residual fluid in a tank that is, the portion of fluid that unavoidably remains in the tank during outflow and hence cannot be used. In this regard, a flat bulkhead is disadvantageous in two respects: (1) It lacks a single low point for optimum placement of an outlet and (2) a vortex that forms at the outlet during outflow prevents a relatively large amount of fluid from leaving the tank. A concave bulkhead also lacks a single low point for optimum placement of an outlet. Like purely concave and purely convex bulkhead configurations, the proposed asymmetric bulkhead configurations would be more mass-efficient than is the flat

  20. Stress intensities in flawed pressure vessels

    International Nuclear Information System (INIS)

    A technique for determining the stess intensity factor (SIF) near pressure vessel flaws or cracks experimentally from photoelastic data for use in two-dimensional problems was developed in the 1950's. This technique was modified and extended to a variety of two-dimensional problems. The technique has been refined further and what has evolved may be regarded as a hybrid technique which affects a marriage between ''frozen stress'' photoelastic results and a simple least-squares digital computer program for estimating SIF values in three-dimensional problems. This technique, in its original modified form, has been shown to be applicable to a study of surface flaws and the applicability of the method to complex crack body geometries of current technological importance are discussed. The analytical foundations of the method are reviewed

  1. Reactor pressure vessel stud management automation strategies

    International Nuclear Information System (INIS)

    The adoption of hydraulic tensioner technology as the standard for bolting and unbolting the reactor pressure vessel (RPV) head 35 yr ago represented an incredible commitment to new technology, but the existing technology was so primitive as to be clearly unacceptable. Today, a variety of approaches for improvement make the decision more difficult. Automation in existing installations must meet complex physical, logistic, and financial parameters while addressing the demands of reduced exposure, reduced critical path, and extended plant life. There are two generic approaches to providing automated RPV stud engagement and disengagement: the multiple stud tensioner and automated individual tools. A variation of the latter would include the handling system. Each has its benefits and liabilities

  2. Holographic and acoustic emission evaluation of pressure vessels

    International Nuclear Information System (INIS)

    Optical holographic interfereometry and acoustic emission monitoring were simultaneously used to evaluate two small, high pressure vessels during pressurization. The techniques provide pressure vessel designers with both quantitative information such as displacement/strain measurements and qualitative information such as flaw detection. The data from the holographic interferograms were analyzed for strain profiles. The acoustic emission signals were monitored for crack growth and vessel quality

  3. Stress Concentration at Openings in Pressure Vessels – A Review

    OpenAIRE

    AVINASH KHARAT; Kulkarni, V. V.

    2013-01-01

    This paper reviews some of the current developments in the determination of stress concentration factor in pressure vessels at openings. The literature has indicated a growing interest in the field of stress concentration analysis in the pressure vessels. The motivation for this research is to analyze the stress concentration occurring at the openings of the pressure vessels and the means to reduce the effect of the same. Most of the researchers have worked on the stress concentration occurri...

  4. 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,...

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

    International Nuclear Information System (INIS)

    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)

  6. Radiation effects on reactor pressure vessel supports

    International Nuclear Information System (INIS)

    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

  7. Fracture assessment for reactor pressure vessel flaw

    International Nuclear Information System (INIS)

    Background: A flaw is discovered by ultrasonic test at reactor pressure vessel during the service. Purpose: In order to insure the structural integrity of RPV, it is necessary to perform the fracture analysis for RPV flaw. Methods: According to ASME rule, fracture analysis is performed, which the flaw is assumed as a circumferential surface crack and crack depth is 10.1 mm. The analysis work contains the calculation of fatigue crack growth and the assessment of stress intensity factor under several category conditions. The loads are the temperature fluctuation, pressure and weld residual stress. The concerned category conditions include normal and upset conditions, emergency conditions, and faulted conditions. Results: The analysis results show that normal and upset conditions transient loading has little effect on the fatigue crack growth of low inner surface crack. The fatigue crack growth is about 0.228 mm at the end of 40 years service life. Conclusions: The stress intensity factor results of all conditions satisfy the requirement of ASME Rule. The RPV with flaw can continue service without repair. (author)

  8. RELAP5-3D Analysis of Pressure Perturbation at the Peach Bottom BWR During Low-Flow Stability Tests

    International Nuclear Information System (INIS)

    Experimental and theoretical studies about the BWR (Boiling Water Reactor) stability have been performed to design a stable core configuration. BWR instabilities can be caused by inter-dependencies between thermal-hydraulic and reactivity feedback parameters such as the void-coefficient, for example, during a pressure perturbation event. In the present work, the pressure perturbation is considered in order to study in detail this type of transient. To simulate this event, including the strong feedback effects between core neutronic and reactor thermal-hydraulics, and to verify core behavior and evaluate parameters related to safety, RELAP5-3D code has been used in the analyses. The simulation was performed making use of Peach Bottom-2 BWR data to predict the dynamics of a real reactor during this type of event. Stability tests were conducted in the Peach Bottom 2 BWR, in 1977, and were done along the low-flow end of the rated power-flow line, and along the power-flow line corresponding to minimum recirculation pump speed. The calculated results are herein compared against the available experimental data. (authors)

  9. Inductive testing of reactor pressure vessels

    International Nuclear Information System (INIS)

    In Service Inspection of Reactor Pressure Vessels is mostly done with ultrasonics. Using special 2 crystal-probes good detectability is achieved for near surface defects. The problem is to detect closely spaced cracks, to decide if the defects are surface braking and, if not, to decide the remaining ligament. The purpose of this study is to investigate to what extent Eddy Current can solve these problems. Detecting surfacebreaking cracks and fields of cracks can be done using conventional Eddy Current techniques. Mapping of closely spaced cracks requires a small probe and a high frequency. Measurement of depths a larger probe, a lower frequency and knowledge of the crackfield since 2 closely spaced shallow cracks might be mistaken for one deep crack. Depths of singel cracks can be measured down to 7-8 mm. In closely spaced crackfields the depths can not be measured. The measurement is mostly based on amplitude. For not surface breaking defects the problem is to decide the ligament, i.e. the distance between surface and cracktip. To achieve good penetration a large probe, low frequency and high energy or pulsed energy is used. Ligament up to 4 mm can be measured with good accuracy. The measurements is mostly based on phase. Noise, which originates from rough surface, varied material structure and lift off, can be reduced using multi frequency mix, probe design and scanning pattern. (author)

  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

    International Nuclear Information System (INIS)

    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. Quality assurance in fabrication of boilers and pressure vessels

    International Nuclear Information System (INIS)

    Quality assurance for safety and reliability of boilers and pressure vessels is a systematic approach involving various stages right from material identification to final stages of testing, transportation and storage before commissioning. This paper brings out various quality . Assurance aspects to be implemented by manufacturers of boilers and pressure vessels. 1 fig

  12. Variability of mechanical properties of nuclear pressure vessel steels

    International Nuclear Information System (INIS)

    Causes of variability of mechanical properties nuclear pressure vessel steels are reviewed and discussed. The effects of product shape and size, processing history and heat treatment are investigated. Some quantitative informations are given on the scatter of mechanical properties of typical pressure vessel components. The necessity of using recommended or standardized properties for comparing mechanical properties before and after irradiation in pin pointed. (orig.)

  13. Measurement of pressure drops in prototypic BWR and PWR fuel assemblies in the laminar regime - Pressure drop measurement of laminar air flow in prototypic BWR and PWR fuel assemblies

    International Nuclear Information System (INIS)

    Laminar pressure drops in nuclear fuel assemblies are of interest for evaluating complete loss-of-coolant accident scenarios in spent fuel pools and for performance analyses of dry storage casks. To the knowledge of the authors, this study represents the first attempt to directly quantify pressure losses in prototypic fuel assemblies in the laminar regime. Two commercial fuel assemblies were examined including a 17x17 PWR and a 9x9 BWR. The assemblies were tested in the laminar regime with Reynolds numbers ranging from 10 to 1000, based on the average assembly velocity and hydraulic diameter. Pressure drop measurements were made across individual bundle spans and grid spacers in the mock fuel assemblies using high-sensitivity differential pressure gauges. These gauges are capable of detecting extremely small changes in differential pressure with a resolution of ∼0.02 Pa. This level of sensitivity allows meaningful pressure drop measurements across separate fuel components, even at low Reynolds numbers. The fuel assembly mock-ups were constructed from commercial fuel assembly structural components and stainless steel tubing that is within 0.6 pc of the outer diameter of actual fuel. The outer flow boundary in the BWR assembly bundle was defined by the walls of a prototypic canister. In the PWR assembly, the flow was confined by the walls of different stainless steel storage cells. Two of the PWR storage cell sizes represented dimensions spanning pool and cask cells available in industry. Pressure ports were installed along the length of the assemblies at locations corresponding to the entrance and exit of fuel components. Dry, ambient air was metered into the bottom of each assembly through a flow straightener. The geometries of the tube bundles in 17x17 PWR and 9x9 BWR fuel assemblies are fundamentally different. The PWR bundle has a larger flow area and incorporates more grid spacers compared to the BWR bundle. Additionally, eight of the 74 fuel rods in the 9x9

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

    Science.gov (United States)

    2010-10-01

    ... subjected to a hydrostatic test at a pressure of 11/4 times the maximum allowable working pressure twice... normally be subjected to a hydrostatic test: (1) Tubular heat exchangers. (2) Pressure vessels used in... shall be subjected to a hydrostatic test of 11/2 times the maximum allowable working pressure in...

  15. In-service ultrasonic inspection of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Ultrasonic tests of pressure vessels for WWER 440 reactors, type 213 V, are carried out partly manually and partly by test equipment. The inner surface of the pressure vessel is tested using device REACTORTEST TRC which is fully mobile. The outer surface of the cylindrical parts and bottoms of the body is tested using handling equipment permanently in-built under the pressure vessel and dismountable testing heads. A set of these heads may be used for two reactor units. The testing equipment REACTORTEST TRC is equipped with a TRC 800 ultrasound device. The equipment for testing the outer surface of the vessel operates with the UDAR 16 ultrasound apparatus to which may be simultaneously connected 10 ultrasound probes and six probes for acoustic feedback. The whole system of ultrasonic tests makes possible a first-rate and reliable volume control of the whole pressure vessel and all points where cracks may originate and grow. (Z.M.)

  16. Pressurized Thermal Shock Analysis for OPR1000 Pressure Vessel

    International Nuclear Information System (INIS)

    The study provides a brief understanding of the analysis procedure and techniques using ANSYS, such as the acceptance criteria, selection and categorization of events, thermal analysis, structural analysis including fracture mechanics assessment, crack propagation and evaluation of material properties. PTS may result from instrumentation and control malfunction, inadvertent steam dump, and postulated accidents such as smallbreak (SB) LOCA, large-break (LB) LOCA, main steam line break (MSLB), feedwater line breaks and steam generator overfill. In this study our main focus is to consider only the LB LOCA due to a cold leg break of the Optimized Power Reactor 1000 MWe (OPR1000). Consideration is given as well to the emergency core cooling system (ECCS) specific sequence with the operating parameters like pressure, temperature and time sequences. The static structural and thermal analysis to investigate the effects of PTS on RPV is the main motivation of this study. Specific surface crack effects and its propagation is also considered to measure the integrity of the RPV. This study describes the procedure for pressurized thermal shock analysis due to a loss of coolant accidental condition and emergency core cooling system operation for reactor pressure vessel.. Different accidental events that cause pressurized thermal shock to nuclear RPV that can also be analyzed in the same way. Considering the limitations of low speed computer only the static analysis is conducted. The modified LBLOCA phases and simplified geometry can is utilized to analyze the effect of PTS on RPV for general understanding not for specific specialized purpose. However, by integrating the disciplines of thermal and structural analysis, and fracture mechanics analysis a clearer understanding of the total aspect of the PTS problem has resulted. By adopting the CFD, thermal hydraulics, uncertainties and risk analysis for different type of accidental conditions, events and sequences with proper

  17. Guidelines for the pressure and efficient sizing of pressure vessels for compressed air energy storage

    International Nuclear Information System (INIS)

    Highlights: ► In small-scale CAES there are no robust guidelines in choosing an operational pressure for the vessels. ► Through the stress analysis of the vessel, an optimum pressure at minimum cost can be determined. ► One contribution is in determining the optimum pressure is small-scale CAES. ► Another contribution is in determining the shape size, and number of vessels in small-scale CAES designs. - Abstract: The paper reports guidelines for the efficient design and sizing of Small-Scale Compressed Air Energy Storage (SS-CAES) pressure vessels, including guidelines for pressures that should be used in the SS-CAES system to minimize the cost of the pressure vessel. Under a specified energy storage capacity and specified maximum and minimum operating pressures in CAES, the volume of the vessel(s) can be evaluated. The present study provides guidelines for choosing appropriate shape and size for the vessels that minimize material and manufacturing cost for cylindrical vessels. The two main contributions of the paper are that it provides a methodology to determine: (a) an optimum pressure; (b) the shape, size, and number of vessel to be used in a particular application. Results suggest that pressure vessels with a length to diameter ratio of roughly three are the most economical, and that a system should be designed for a pressure of roughly three times the minimum pressure of the expansion device.

  18. Construction and operation of pressure vessels. 2. ed.

    International Nuclear Information System (INIS)

    The safe operation and availability of pressure vessels depends on the quality of the components. Quality assurance measures are of particular importance in design, choice of material and material processing. The present state of design, calculations, materials, welding technique of very strong fine grained structural steel, heat treatment of welds, measurement technique in heat treatment and non-destructive testing are all described. Special attention is paid to safety of operation of pressure vessels and assessing faults in pressurized walls. (DG)

  19. Fatigue and fracture mechanics analysis of high pressure vessels

    International Nuclear Information System (INIS)

    A companion document to the ASME Boiler and Pressure Code, section VIII, Division 3 Alternative rules for Construction of High Pressure Vessels, emphasizes that Division 3 is 'intended for vessels in the design pressure range of about 10,000 to 165,000 psi: but no upper or lower limits are given nor are any upper limits implied for Divisions 1 and 2'. Although Division 3 includes much information on welded vessels, attention herein will be focused on the design of vessels that operate above 10,000 psi and are of monobloc or compound construction, using two or more cylindrical shells, that are forged from low- alloy high strength steel ingots. Threaded, pinned, or clamped closures are often used at one or both ends of such vessels; in some cases one end may be forged integrally with the cylindrical body, so-called 'blind end' closure. High pressure piping is connected to such vessels at holes in the end closures or at holes through the cylindrical part of the vessel, usually referred to as cross-bores. These vessels or components thereof may be overstrained (autofrettaged) to enhance static and/or fatigue performance. Division 3 was first published in 1997. In the interim there have been advances in the static, fatigue, and fracture analysis of such vessels and their closures and connections. The aim of this paper is to review some of these advances. (author)

  20. Development of electrical cable penetration for secondary containment vessel of BWR type nuclear power plants

    International Nuclear Information System (INIS)

    The penetration holes in the walls and floors of the secondary containment vessel of the nuclear power plants must be air-tight, shielded against the radiation, and fire-resistant. At present, the penetration holes are air-tightened with iron plates and sealing material after the cables are laid. However, installation of a number of cables and its sealing work now pose a serious problem in nuclear power plant construction in relation to the installation of reactor system components. The authors have recently developed a method for electric wall penetration in an attempt to solve this problem. This method is provided with prefabricated cable portions for wall penetration, reducing field work, saving labor in wiring work through use of multicore cables, and increasing the reliability of the sealing and caulking work. This wall penetration consists of an iron sleeves to be embedded into the wall, a header-plate, and an assembly of modules in which a specified number of insulated conductors are set up, and furthermore termination boxes are installed on both ends of the penetration holes. This paper deals with the design standard and construction of the wall penetration and the results of tests which were performed under various environmental conditions, which has shown excellent properties, such as sealing quality and electric characteristics, of the wall penetration. (author)

  1. Reliable estimation of neutron flux in BWR reactor vessel using the tort code (2) application to neutron and gamma flux estimation

    International Nuclear Information System (INIS)

    A neutron and gamma flux distribution around the core of BWR commercial plant in Japan was calculated, using a three-dimensional transport code, TORT in DOORS32 code system. In the external of the core, the bottom of the model was at an elevation of 150 cm below the bottom of active fuel, the top of the model was at an elevation of the top of the shroud head dome and the radial part of the model was to the outside of the reactor pressure vessel. The top guide beams were modeled explicitly to obtain the neutron and gamma flux distribution both in the beams and outside beams. The each control rod guide tube was also modeled with homogeneous region which included the blade wing and poison tubes so that we could obtain the neutron and gamma flux distribution around the each control rod guide tube. The calculation model mentioned above needed very large memory size which exceeded a few decade giga-bytes. As the using the splicing/coupling method had uncertainly at the splicing/coupling boundary, in this work the calculation was performed without this splicing/coupling method. On the other hand, radioactivity data were measured for a few pieces of the top guide beam, shroud and in-core monitor guide tube in the same plant which was analyzed in the above calculation. So the calculation results were able to be compared with those measured data as benchmarking and at the end of this task, the C/M values at these measured points were obtained and calculation model using TORT was evaluated. (authors)

  2. Firefighter's compressed air breathing system pressure vessel development program

    Science.gov (United States)

    Beck, E. J.

    1974-01-01

    The research to design, fabricate, test, and deliver a pressure vessel for the main component in an improved high-performance firefighter's breathing system is reported. The principal physical and performance characteristics of the vessel which were required are: (1) maximum weight of 9.0 lb; (2) maximum operating pressure of 4500 psig (charge pressure of 4000 psig); (3) minimum contained volume of 280 in. 3; (4) proof pressure of 6750 psig; (5) minimum burst pressure of 9000 psig following operational and service life; and (6) a minimum service life of 15 years. The vessel developed to fulfill the requirements described was completely sucessful, i.e., every category of performence was satisfied. The average weight of the vessel was found to be about 8.3 lb, well below the 9.0 lb specification requirement.

  3. BWR type nuclear reactor

    International Nuclear Information System (INIS)

    Purpose: To simplify the structure of an emergency core cooling system while suppressing the flow out of coolants upon rapture accidents in a coolant recycling device of BWR type reactors. Constitution: Recirculation pumps are located at a position higher than the reactor core in a pressure vessel, and the lower plenum is bisected vertically by a partition plate. Further, a gas-liquid separator is surrounded with a wall and the water level at the outer side of the wall is made higher than the water level in the inside of the wall. In this structure, coolants are introduced from the upper chamber in the lower plenum into the reactor core, and the steams generated in the reactor core are separated in the gas-liquid separator, whereby the separated liquid is introduced as coolants by way of the inner chamber into the lower chamber of the lower plenum and further sent by way of the outer chamber into the reactor core. Consequently, idle rotation of the recycling pumps due to the flow-in of saturated water is prevented and loss of coolants in the reactor core can also be prevented upon raptures in the pipeway and the driving section of the pump connected to the pressure vessel and in the bottom of the pressure vessel. (Horiuchi, T.)

  4. Simulation of a scenario of total loss of external and internal power (Sbo) for different vent pressures of the containment of a BWR-5

    International Nuclear Information System (INIS)

    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/cm2) 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/cm2), 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)

  5. 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.

  6. Strength-toughness requirements for thick walled high pressure vessels

    International Nuclear Information System (INIS)

    The strength and toughness requirements of materials for use in high pressure vessels has been the subject of some discussion in the meetings of the Materials Task Group of the Special Working Group High Pressure Vessels. A fracture mechanics analysis has been performed to theoretically establish the required toughness for a high pressure vessel. This paper reports that the analysis performed is based on the validity requirement for plane strain fracture of fracture toughness test specimens. This is that at the fracture event, the crack length, uncracked ligament, and vessel length must each be greater than fifty times the crack tip plastic zone size for brittle fracture to occur. For high pressure piping applications, the limiting physical dimension is the uncracked ligament, as it can be assumed that the other dimensions are always greater than fifty times the crack tip plastic zone. To perform the fracture mechanics analysis several parameters must be known: these include vessel dimensions, material strength, degree of autofrettage, and design pressure. Results of the analysis show, remarkably, that the effects of radius ratio, pressure and degree of autofrettage can be ignored when establishing strength and toughness requirements for code purposes. The only parameters that enter into the calculation are yield strength, toughness and vessel thickness. The final results can easily be represented as a graph of yield strength against toughness on which several curves, one for each vessel thickness, are plotted

  7. Evaluation of catastrophic failure risk in pressure vessels

    International Nuclear Information System (INIS)

    Within the Nordic countries a four-year research programme in the area of elastic-plastic fracture mechanics was initiated in 1985. Seven laboratories from Denmark, Finland, Norway and Sweden are participating in the programme. The main technical objective of the programme is to clarify how catastrophic fracture can be prevented in pressure vessels and piping by using the leak-before-break concept. The major experimental effort of the programme is destructive pressurization of a large size pressure vessel up to rupture. The vessel has dimensions similar to a nuclear reactor pressure vessel and it has been in operation for 20 years in a Finnish oil refinery plant. The materials characterization of the vessel has been partially carried out within an extensive Nordic round-robin programme. Two pressure tests have been carried out. In both tests an artificial sharp axial surface flaw was made on the inner wall of the vessel. The experimental details of the last test including repair welding of the vessel, flaw prepration, instrumentation and material characterization are described in this report. The fracture behaviour as well as experimental results are reported. The failure pressure is compared to estimates of the analytical pre-test calculations

  8. Design of pressure vessels using shape optimization: An integrated approach

    International Nuclear Information System (INIS)

    Previous papers related to the optimization of pressure vessels have considered the optimization of the nozzle independently from the dished end. This approach generates problems such as thickness variation from nozzle to dished end (coupling cylindrical region) and, as a consequence, it reduces the optimality of the final result which may also be influenced by the boundary conditions. Thus, this work discusses shape optimization of axisymmetric pressure vessels considering an integrated approach in which the entire pressure vessel model is used in conjunction with a multi-objective function that aims to minimize the von-Mises mechanical stress from nozzle to head. Representative examples are examined and solutions obtained for the entire vessel considering temperature and pressure loading. It is noteworthy that different shapes from the usual ones are obtained. Even though such different shapes may not be profitable considering present manufacturing processes, they may be competitive for future manufacturing technologies, and contribute to a better understanding of the actual influence of shape in the behavior of pressure vessels. - Highlights: → Shape optimization of entire pressure vessel considering an integrated approach. → By increasing the number of spline knots, the convergence stability is improved. → The null angle condition gives lower stress values resulting in a better design. → The cylinder stresses are very sensitive to the cylinder length. → The shape optimization of the entire vessel must be considered for cylinder length.

  9. Design of pressure vessels using shape optimization: An integrated approach

    Energy Technology Data Exchange (ETDEWEB)

    Carbonari, R.C., E-mail: ronny@usp.br [Department of Mechatronic Engineering, Escola Politecnica da Universidade de Sao Paulo, Av. Prof. Mello Moraes, 2231 Sao Paulo, SP 05508-900 (Brazil); Munoz-Rojas, P.A., E-mail: pablo@joinville.udesc.br [Department of Mechanical Engineering, Universidade do Estado de Santa Catarina, Bom Retiro, Joinville, SC 89223-100 (Brazil); Andrade, E.Q., E-mail: edmundoq@petrobras.com.br [CENPES, PDP/Metodos Cientificos, Petrobras (Brazil); Paulino, G.H., E-mail: paulino@uiuc.edu [Newmark Laboratory, Department of Civil and Environmental Engineering, University of Illinois at Urbana-Champaign, 205 North Mathews Av., Urbana, IL 61801 (United States); Department of Mechanical Science and Engineering, University of Illinois at Urbana-Champaign, 158 Mechanical Engineering Building, 1206 West Green Street, Urbana, IL 61801-2906 (United States); Nishimoto, K., E-mail: knishimo@usp.br [Department of Naval Architecture and Ocean Engineering, Escola Politecnica da Universidade de Sao Paulo, Av. Prof. Mello Moraes, 2231 Sao Paulo, SP 05508-900 (Brazil); Silva, E.C.N., E-mail: ecnsilva@usp.br [Department of Mechatronic Engineering, Escola Politecnica da Universidade de Sao Paulo, Av. Prof. Mello Moraes, 2231 Sao Paulo, SP 05508-900 (Brazil)

    2011-05-15

    Previous papers related to the optimization of pressure vessels have considered the optimization of the nozzle independently from the dished end. This approach generates problems such as thickness variation from nozzle to dished end (coupling cylindrical region) and, as a consequence, it reduces the optimality of the final result which may also be influenced by the boundary conditions. Thus, this work discusses shape optimization of axisymmetric pressure vessels considering an integrated approach in which the entire pressure vessel model is used in conjunction with a multi-objective function that aims to minimize the von-Mises mechanical stress from nozzle to head. Representative examples are examined and solutions obtained for the entire vessel considering temperature and pressure loading. It is noteworthy that different shapes from the usual ones are obtained. Even though such different shapes may not be profitable considering present manufacturing processes, they may be competitive for future manufacturing technologies, and contribute to a better understanding of the actual influence of shape in the behavior of pressure vessels. - Highlights: > Shape optimization of entire pressure vessel considering an integrated approach. > By increasing the number of spline knots, the convergence stability is improved. > The null angle condition gives lower stress values resulting in a better design. > The cylinder stresses are very sensitive to the cylinder length. > The shape optimization of the entire vessel must be considered for cylinder length.

  10. Neutronic security calculus for the VVER-440 pressure vessel

    International Nuclear Information System (INIS)

    This paper shows how to estimate flow of quick neutrons as well as other structural components received by the vessel during its useful life span. This is done taking into account standard project loads. Values for the Lead Factor at the surface of the vessel are given also and finally critical fragility temperature is evaluated for the VVER-440 pressure vessel at the Juragua nuclear power station

  11. Containment venting sliding pressure venting process for PWR and BWR plants

    International Nuclear Information System (INIS)

    In order to reduce the residual risk associated with hypothetical severe nuclear accidents, nuclear power plants in Germany as well as in certain other European countries have been or will be backfitted with a system for filtered containment venting. During venting system process design, particular importance is attached to the requirements regarding, for example, high aerosol loading capability, provision for decay heat removal from the scrubber unit, the aerosol spectrum to be retained and entirely passive functioning of the scrubber unit. The aerosol spectrum relevant for process design and testing varies depending on aerosol concentrations, the time at which venting is commenced and whether there is an upstream wetwell, etc. Because of this the Reactor Safety Commission in Germany has specified that SnO2 with a mass mean diameter of approximately 0.5 μm should be used as an enveloping test aerosol. To meet the above-mentioned requirements, a combined venturi scrubber system was developed which comprises a venturi section and a filter demister section and is operated in the sliding pressure mode. This scrubber system was tested using a full-scale model and has now been installed in 14 PWR and BWR plants in Germany and Finland

  12. Crashworthy sealed pressure vessel for plutonium transport

    International Nuclear Information System (INIS)

    A rugged transportation package for the air shipment of radioisotopic materials was recently developed. This package includes a tough, sealed, stainless steel inner containment vessel of 1460 cc capacity. This vessel, intended for a mass load of up to 2 Kg PuO2 in various isotopic forms (not to exceed 25 watts thermal activity), has a positive closure design consisting of a recessed, shouldered lid fastened to the vessel body by twelve stainless-steel bolts; sealing is accomplished by a ductile copper gasket in conjunction with knife-edge sealing beads on both the body and lid. Follow-on applications of this seal in newer, smaller packages for international air shipments of plutonium safeguards samples, and in newer, more optimized packages for greater payload and improved efficiency and utility, are briefly presented

  13. Mechanical behaviour of reactor pressure vessel in severe accident

    International Nuclear Information System (INIS)

    The article describes the main achievements in developing methodology for analysing mechanical behaviour of pressure vessels with and without penetrations at high temperatures related to accident scenarios. Validation and applications of the methodology are presented. (orig.)

  14. Water level monitoring and controlling system in reactor pressure vessel

    International Nuclear Information System (INIS)

    Upon controlling of a water level conducted before an operation of removing a head of a reactor pressure vessel performed upon periodical inspection of a power plant, a supersonic displacement sensor capable of conducting measurement on the basis of cm unit is used as a water level indicator. The water level in the reactor pressure vessel can be controlled to a position above a steam dryer accurately by a remote operation while measuring the water level in the pressure vessel at a high accuracy. In addition, a dose equivalent gauge for evaluating the dose equivalent rate of the operation circumstance is previously disposed before the removal of the head of the reactor pressure vessel to reduce the amount of operation and exposure dose of a radiation operation manager. These supersonic displacement sensor and dose equivalent gauge are made detachable so that they can be disposed only when they are required thereby enabling to minimize the deterioration of the device. (N.H.)

  15. Basic conceptions for reactor pressure vessel manipulators and their evaluation

    International Nuclear Information System (INIS)

    The study deals with application fields and basic design conceptions of manipulators in reactor pressure vessels as well as their evaluation. It is shown that manipulators supported at the reactor flange have essential advantages

  16. 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.

  17. Evaluation of pressurized thermal shock in transitional condition for boiling water reactor pressure vessel

    International Nuclear Information System (INIS)

    The structural integrity for Pressurized Thermal Shock (PTS) was evaluated for the RPVs of Japanese Boiling Water Reactors (BWRs). It has been clarified that the BWR RPVs have the sufficient margin of fracture toughness by calculating the stress intensity factor in transitional condition and the acceptance criteria for RPV shell plate which is assumed to be neutron-irradiated in core region for 60 years. (author)

  18. Innovative Impact Protection and Monitoring System for Composite Pressure Vessels

    OpenAIRE

    Kopperud, Paul Andreas

    2015-01-01

    Impact behavior and resistance of composite structures are difficult to predict. For composite pressure vessels, where failure can be fatal, impact protection and detection is particularly important. This thesis aims to render high pressure composite vessels safer to use with regards to impact. Three main objectives were identified; Firstly, finding an effective impact protection method and material. Secondly, developing a low cost impact detection system. Lastly, find an approach to estimate...

  19. Design and fabrication of HTTR reactor pressure vessel

    International Nuclear Information System (INIS)

    The High Temperature Engineering Test Reactor (HTTR) is under construction at the Oarai Research Establishment, JAERI and planned to be critical at the end of 1997. The HTTR is a High Temperature Gas-cooled Reactors (HTGRs) with thermal output of 30MW, inlet coolant temperature of 395degC, and outlet coolant temperature of 850degC at rated operation and 950degC at high temperature test operation. 2.25Cr-1Mo steel is chosen for the reactor pressure vessel of the HTTR because its temperature reaches about 400degC at normal operation. 2.25Cr-1Mo steel has higher creep rupture strength than Mn-Mo steel used for the reactor pressure vessels of Light Water Reactors (LWRs). For the components of the HTTR reactor pressure vessel subjected to low temperatures where creep deformation is negligible, a design guideline based on Japanese structural design standard for LWRs 'Technical standards for LWR power plant components-Ministry of International Trade and Industry Standard No.501' is utilized. On the other hand, design of the components for high temperature application, where creep behavior dominates, is conducted under newly determined high temperature structural design guideline and design material data. The fabrication of the HTTR reactor pressure vessel took about 23 months. It was installed in a reactor containment vessel in August, 1994. After core components had been installed in the reactor pressure vessel, pressure test of the primary and secondary cooling system including the reactor pressure vessel was performed and successfully ended in March, 1996. This paper reports issues of the HTTR reactor pressure vessel such as structure, material, stress analysis, fabrication, examination and testing. (author)

  20. Heavy wall pressure vessels for energy systems

    International Nuclear Information System (INIS)

    Modifications of steels currently accepted in the Code appear to provide improved mechanical properties. These steels may permit the fabrication of larger diameter vessels with thinner section sizes and improved reliability and integrity. Adapting current specifications should expedite Code approval. Finally the challenge of improving welding procedures and adapting processes for field applications will result in higher quality weldments

  1. Simplified system for the pressure control of a Nucleo electric central of the BWR type

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    On-line measurements of the physical state of a prestressed concrete pressure vessel and a comparison of the distribution of temperature, strain and stress within the concrete member to the optimized statical predictions and the criterions of layout yield to an efficient and economical method of operating the vessel with a high potential of safety. The requirements of instrumentation and the comparison with static calculations are discussed on the prototype vessel at Seibersdorf Research Center during the phase of construction and prestressing, the phase of the first thermal treatment (stabilization), the pressure tests and under the operating conditions of a high temperature reactor (1500C/50 bar). (Author)

  3. Jam proof closure assembly for lidded pressure vessels

    International Nuclear Information System (INIS)

    This patent describes a pressure vessel cover adapted for sealing engagement with an outer peripheral lip of an underlying pressure vessel and having a compressible annular gasket overlying the lip, the cover further comprising: a spaced peripheral array of cylindrical bored passageways communicating between outer and inner planar surfaces of the vessel cover; a rigid, comparatively thick-walled, annular sleeve insert; a substantially rigid retainer element configured to seat tightly against the inserted sleeve; bolt threaded to engage an inner surface of the positioned sleeve insert

  4. Light Water Reactor-Pressure Vessel Surveillance project computer system

    International Nuclear Information System (INIS)

    A dedicated process control computer has been implemented for regulating the metallurgical Pressure Vessel Wall Benchmark Facility (PSF) at the Oak Ridge Research Reactor. The purpose of the PSF is to provide reliable standards and methods by which to judge the radiation damage to reactor pressure vessel specimens. Benchmark data gathered from the PSF will be used to improve and standardize procedures for assessing the remaining safe operating lifetime of aging reactors. The computer system controls the pressure vessel specimen environment in the presence of gamma heating so that in-vessel conditions are simulated. Instrumented irradiation capsules, in which the specimens are housed, contain temperature sensors and electrical heaters. The computer system regulates the amount of power delivered to the electrical heaters based on the temperature distribution within the capsules. Time-temperature profiles are recorded along with reactor conditions for later correlation with specimen metallurgical changes

  5. 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.

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

    International Nuclear Information System (INIS)

    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 13. 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.

  7. PWR reactor pressure vessel failure probabilities

    International Nuclear Information System (INIS)

    To evaluate the rupture probability of a LWR vessel a probabilistic method using the fracture mechanics under probabilistic form has been proposed previously, but it appears that more accurate evaluation is possible. In consequence a joint collaboration agreement signed in 1976 between CEA, EURATOM, JRC Ispra and FRAMATOME set up and started a research program covering three parts: a computer code development, data acquisition and processing, and a support experimental program which aims at clarifying the most important parameters used in the COVASTOL computer code

  8. Transportable, small high-pressure preservation vessel for cells

    Energy Technology Data Exchange (ETDEWEB)

    Kamimura, N; Sotome, S; Shimizu, A [Department of Environmental Engineering for Symbiosis, Soka University, 1-236 Tangi-cho, Hachioji, Tokyo 192-8577 (Japan); Nakajima, K [Department of Bioinformatics, Soka University, 1-326 Tangi-cho, Hachioji, Tokyo 192-8577 (Japan); Yoshimura, Y, E-mail: mf_kamimura@yahoo.co.j [Department of Applied Chemistry, National Defence Academy, 1-10-20 Hashirimizu, Yokosuka, Kanagawa 239-8686 (Japan)

    2010-03-01

    We have previously reported that the survival rate of astrocytes increases under high-pressure conditions at 4{sup 0}C. However, pressure vessels generally have numerous problems for use in cell preservation and transportation: (1) they cannot be readily separated from the pressurizing pump in the pressurized state; (2) they are typically heavy and expensive due the use of materials such as stainless steel; and (3) it is difficult to regulate pressurization rate with hand pumps. Therefore, we developed a transportable high-pressure system suitable for cell preservation under high-pressure conditions. This high-pressure vessel has the following characteristics: (1) it can be easily separated from the pressurizing pump due to the use of a cock-type stop valve; (2) it is small and compact, is made of PEEK and weighs less than 200 g; and (3) pressurization rate is regulated by an electric pump instead of a hand pump. Using this transportable high-pressure vessel for cell preservation, we found that astrocytes can survive for 4 days at 1.6 MPa and 4{sup 0}C.

  9. Weld evaluation on spherical pressure vessels using holographic interferometry

    International Nuclear Information System (INIS)

    Waist welds on spherical experimental pressure vessels have been evaluated under pressure using holographic interferometry. A coincident viewing and illumination optical configuration coupled with a parabolic mirror was used so that the entire weld region could be examined with a single hologram. Positioning the pressure vessel at the focal point of the parabolic mirror provides a relatively undistorted 360 degree view of the waist weld. Double exposure and real time holography were used to obtain displacement information on the weld region. Results are compared with radiographic and ultrasonic inspections

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

    International Nuclear Information System (INIS)

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

  11. Definition of welded joints arrangement in pressure vessels

    International Nuclear Information System (INIS)

    The arrangement of welded joints in a pressure vessel is normally a tiresome and lengthy task. This is especially so when long vessels are provided with a great concentration of nozzles, reinforcements, rings and many other accessories. The manual solution is seldom economical from the viewpoint of minimum waste of plates and minimum labor for cutting, bevelling and welding. This paper presents a computer application that allows a more economical solution in a shorter time. (Author)

  12. Reliable analysis for pressure vessel based on ANSYS

    International Nuclear Information System (INIS)

    With the PDS of ANSYS procedure, the ramdomicity of the actually structure design parameters is simulated, by taking the wall thickness, pressure load and elastic module as input random variables. Based on the reliability analysis of the pressure vessel by Monte-Carlo procedure, the stress probability distribution of this finite element analysis model and the sensitivity of the design parameters such as the pressure load and wall thickness to the stress distribution are obtained. (authors)

  13. Simplified distributed parameters BWR dynamic model for transient and stability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco, 186, Col. Vicentina, Mexico, DF 09340 (Mexico)]. E-mail: gepe@xanum.uam.mx; Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Doctor Barragan 779, Col. Narvarte, Mexico, DF 03020 (Mexico); Vazquez-Rodriguez, Alejandro [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco, 186, Col. Vicentina, Mexico, DF 09340 (Mexico)

    2006-09-15

    This paper describes a simplified model to perform transient and linear stability analysis for a typical boiling water reactor (BWR). The simplified transient model was based in lumped and distributed parameters approximations, which includes vessel dome and the downcomer, recirculation loops, neutron process, fuel pin temperature distribution, lower and upper plenums reactor core and pressure and level controls. The stability was determined by studying the linearized versions of the equations representing the BWR system in the frequency domain. Numerical examples are used to illustrate the wide application of the simplified BWR model. We concluded that this simplified model describes properly the dynamic of a BWR and can be used for safety analysis or as a first approach in the design of an advanced BWR.

  14. Crack initiation and arrest in a pressurized thermal shock test for a model pressure vessel made of VVER-440 reactor pressure vessel steel

    International Nuclear Information System (INIS)

    A joint pressure vessel integrity research programme involving three partners is being carried out during 1990-1994. The partners are the Central Research Institute of Structural Materials ''Prometey'' from Russia, IVO International Ltd (IVO) from Finland, and the Technical Research Centre of Finland (VTT). The main objective of the research programme is to increase the reliability of the VVER-440 reactor pressure vessel safety analysis. This is achieved by providing material property data for the VVER-440 pressure vessel steel, and by producing experimental understanding of the crack behaviour in pressurized thermal shock loading for the validation of different fracture assessment methods. The programme is divided into four parts: pressure vessel tests, material characterization, computational fracture analyses, and evaluation of the analysis methods. The testing programme comprises tests on two model pressure vessels with artificial axial outer surface flaws. The first model vessel had circumferential weld seam at the mid-length of the vessel. A special embrittling heat treatment is applied to the vessels before tests to simulate the fracture toughness at the end-of-life condition of a real reactor pressure vessel. The sixth test on the first model led to crack initiation followed by arrest. After the testing phase, material characterization was performed. Comparison of calculated and experimental data generally led to a good correlation, although the work is being continued to resolve the discrepancies between the measured initiation and arrest properties of the material. (orig.)

  15. 46 CFR 54.01-10 - Steam-generating pressure vessels (modifies U-1(g)).

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Steam-generating pressure vessels (modifies U-1(g)). 54... ENGINEERING PRESSURE VESSELS General Requirements § 54.01-10 Steam-generating pressure vessels (modifies U-1(g)). (a) Pressure vessels in which steam is generated are classed as “Unfired Steam Boilers” except...

  16. Beryllium pressure vessels for creep tests in magnetic fusion energy

    International Nuclear Information System (INIS)

    Beryllium has interesting applications in magnetic fusion experimental machines and future power-producing fusion reactors. Chief among the properties of beryllium that make these applications possible is its ability to act as a neutron multiplier, thereby increasing the tritium breeding ability of energy conversion blankets. Another property, the behavior of beryllium in a 14-MeV neutron environment, has not been fully investigated, nor has the creep behavior of beryllium been studied in an energetic neutron flux at thermodynamically interesting temperatures. This small beryllium pressure vessel could be charged with gas to test pressures around 3, 000 psi to produce stress in the metal of 15,000 to 20,000 psi. Such stress levels are typical of those that might be reached in fusion blanket applications of beryllium. After contacting R. Powell at HEDL about including some of the pressure vessels in future test programs, we sent one sample pressure vessel with a pressurizing tube attached (Fig. 1) for burst tests so the quality of the diffusion bond joints could be evaluated. The gas used was helium. Unfortunately, budget restrictions did not permit us to proceed in the creep test program. The purpose of this engineering note is to document the lessons learned to date, including photographs of the test pressure vessel that show the tooling necessary to satisfactorily produce the diffusion bonds. This document can serve as a starting point for those engineers who resume this task when funds become available

  17. Noninvasive liquid level/density measurement in pressure vessels

    International Nuclear Information System (INIS)

    This research investigated and demonstrated the principles of noninvasive detection of liquid level/density variations in a nuclear reactor pressure vessel. The noninvasive signal detection technique is based on using ex-vessel fast neutron detectors to sense variations in the escape rate of fast neutrons with changes in level/density in the pressure vessel. A prototype instrumentation package, deploying four fission chambers in a string, was developed and tested at the Penn State Breazeale Nuclear Reactor, as well as in six loss-of-coolant experiments at the Loss of Fluid Test Facility of the Idaho National Engineering Laboratory, Idaho Falls, Idaho. The six loss of coolant experiments consisted of two large break and four small break simulations. The prototype instrumentation package was microcomputer based, and was designed to operate in both current and pulse models. It tracked, accurately and quickly, the hydraulic conditions in the pressure vessel during these experiments. Analysis of its response data showed clear identification of: (a) downcomer voiding and refilling, (b) core voiding and refilling, (c) combined core and downcomer voiding and refilling, (d) top-down voiding and refilling of the core, (e) bottom-up voiding and refilling of the core, and (f) boiling and frothing in the pressure vessel. A set of algorithms for online detection and tracking of departure from normal hydraulic conditions is presented

  18. Design optimization of a thin walled pressure vessel

    International Nuclear Information System (INIS)

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

  19. A prototype knowledge based system for pressure vessel design

    International Nuclear Information System (INIS)

    The usage of expert system techniques in the area of mechanical engineering design has been studied. A prototype expert system for pressure vessel design has been developed. The work has been carried out in two steps. Firstly, a pre-processor for the finite element system PCFEMP, named INFEMP, was developed. Secondly, an expert supported system for pressure vessel design, named PVES, was developed. Both INFEMP and PVES are integrated to the AutoCAD system, and AutoCAD's language AutoLISP has been used. A practical example has been investigated to demonstrate the principal ideas of the prototype. (au)

  20. Application of fracture mechanics to fatigue in pressure vessels

    International Nuclear Information System (INIS)

    The methods of application of fracture mechanics to predict fatigue crack propagation in welded structures and pressure vessels are described with the following objectives: i) To identify the effect of different variables such as crack tip plasticity, free surface, finite plate thickness, stress concentration and type of the structure, on the magnitude of stress intensity factor K in Welded joint. ii) To demonstrate the use of fracture mechanics for analysing fatigue crack propagation data. iii) To show how a law of fatigue crack propagation based on fracure mechanics, may be used to predict fatigue behavior of welded structures such as pressure vessel. (Author)

  1. Application of acoustic emission in pressure vessel testing

    International Nuclear Information System (INIS)

    Materials of the CrMoV and the NiCrMoV types were tested for integrity using an acoustic emission method developed by SKODA Trust. The materials are used for the production of reactor pressure vessels. The acoustic emission method is employed for determining the beginning of crack formation and crack proliferation. The objective of the tests was to obtain information necessary for evaluating acoustic emission sources in actual components. The use is discussed of a 24-channel system by Trodyne (USA) for testing the integrity of WWER type pressure vessels manufactured by SKODA. (B.S.)

  2. Low Temperature and High Pressure Evaluation of Insulated Pressure Vessels for Cryogenic Hydrogen Storage

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-06-25

    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.

  3. Acoustic emission monitoring and ultrasonic examination correlation on a reactor pressure vessel. Final report

    International Nuclear Information System (INIS)

    The acoustic emission monitoring and corroborative ultrasonic examination of the acoustic emission (AE) locations established during the hydrostatic pressure test of a BWR primary pressure vessel is described. Descriptive information regarding AE is provided as a background and details of the AE and ultrasonic instrumentation, procedures, problems encountered, and test results are discussed. In total, 42 acoustic emission locations were detected, located, and ultrasonically examined during this project. At all 42 AE locations ultrasonic indications were obtained. Of the AE locations, 76% (or 32 of the 42) were confirmed at amplitudes greater than or equal to 2.5% Distance Amplitude Correction (DAC) by either L-wave or shear wave ultrasonic examination, the largest of these being 18% DAC. The remainder of the AE locations were confirmed at amplitudes less than 2.5% DAC. ASME Code requires that ultrasonic examination record for permanent reference indications of 50% DAC or greater. As is to be expected ultrasonic examination detected examinations which were not located by AE monitoring since AE locates only active flaws. Results show the complementary value of AE monitoring to ultrasonic examination in two primary uses: determining the existence and the location of active discontinuities; and assuring that active discontinuities are not overlooked. Results reflect the position that AE monitoring and ultrasonics are supplementary to each other, not replacements for one another

  4. Toughness properties of end of life reactor pressure vessel cladding

    International Nuclear Information System (INIS)

    The inside surface of reactor pressure vessels is protected against corrosion by a clad overlay made of austenic stainless steel. This cladding, applied by automatic submerged arc welding with strip electrode, is constituted by two layers, the first one in 309L (24 CR - 12 Ni) steel and the second one in 308L (18 Cr - 10 Ni) steel. Safety analysis of reactor pressure vessel (RPV) consists to verify the resistance to fracture of the vessel, assumed containing a small crack just under the cladding. That implies the knowledge of the mechanical properties of the cladding. With the objective to evaluate the mechanical properties of the cladding at the end of life of the RPV, a coordinated French experimental programme has been carried out jointly by CEA, EDF, and Framatome. The first experimental results of this programme are given in this paper. 10 figs, 3 tabs

  5. Possible research program on a large scale nuclear pressure vessel

    International Nuclear Information System (INIS)

    The nuclear pressure vessel structural integrity is actually one of the main items in the nuclear plants safety field. An international study group aimed at investigating the feasibility of a ''possible research program'' on a scale 1:1 LWR pressure vessel. This report presents the study group's work. The different research programs carried out or being carried out in various countries of the European Community are presented (phase I of the study). The main characteristics of the vessel considered for the program and an evaluation of activities required for making them available are listed. Research topic priorities from the different interested countries are summarized in tables (phase 2); a critical review by the study group of the topic is presented. Then, proposals for possible experimental programs and combination of these programs are presented, only as examples of possible useful research activities. The documents pertaining to the results of phase I inquiry performed by the study group are reported in the appendix

  6. Monitoring of prestressed concrete pressure vessels for nuclear reactors

    International Nuclear Information System (INIS)

    The paper (1) ''Experience of In-Service Surveillance and Monitoring of Prestressed Concrete Pressure Vessels for Nuclear Reactors'', which was presented by Irving at the York Conference in September 1975, gave details of the statutory requirements for the inspection of prestressed concrete pressure vessels in the United Kingdom, with particular emphasis on the prestressing system. Results were presented of periodic examinations under the Licensing Conditions for the vessels at the gas cooled Magnox reactors at Oldbury and Wylfa, which had been operating since 1967 and 1971 respectively, and these were discussed in relation to design expectations and future requirements. The paper also gave strain, moisture and temperature readings obtained from Oldbury PCPVs over a ten year period and compared these with predictions. The purpose of this paper is to update the information presented in the 1975 York paper by summarizing the results which have been obtained since then up to the present time (1978). The results are summarized below under six headings

  7. Multilayer Pressure Vessel Materials Testing and Analysis. Phase 1

    Science.gov (United States)

    Cardinal, Joseph W.; Popelar, Carl F.; Page, Richard A.

    2014-01-01

    To provide NASA a comprehensive suite of materials strength, fracture toughness and crack growth rate test results for use in remaining life calculations for aging multilayer pressure vessels, Southwest Research Institute (R) (SwRI) was contracted in two phases to obtain relevant material property data from a representative vessel. This report describes Phase 1 of this effort which includes a preliminary material property assessment as well as a fractographic, fracture mechanics and fatigue crack growth analyses of an induced flaw in the outer shell of a representative multilayer vessel that was subjected to cyclic pressure test. SwRI performed this Phase 1 effort under contract to the Digital Wave Corporation in support of their contract to Jacobs ATOM for the NASA Ames Research Center.

  8. Recent experience and new developments in reactor pressure vessel manufacture

    International Nuclear Information System (INIS)

    In this paper, Framatome's recent experience and new developments in the manufacture of pressurized water reactor (PWR) reactor pressure vessel (RPV)s is described to show the very high standards of quality achieved to meet the most stringent requirements. Outstanding new developments include: qualification and utilisation of thick forged shell rings made from large hollow ingots; fully automatic submerged arc narrow gap welding; electroslag stainless steel cladding process; nozzle buttering by automatic hotwire TIG process

  9. Why and how acoustic emission in pressure vessel first hydrotest

    International Nuclear Information System (INIS)

    The main advantages obtained performing the Acoustic Emission (AE) examination during pressure vessel first hydrotest are presented. The characteristics and performance of the AE instrumentation to be used for a correct test are illustrated. The main criteria for AE source characterization (location, typical AE parameters and their correlation with pressure value), the calibration and test procedures are discussed. The ndt post-test examinations and laboratory specimen experiments are also outlined. Personnel qualification requirements are finally indicated. (Author)

  10. Requirements on crack detection in pressurized vessels for ASME authorization

    International Nuclear Information System (INIS)

    The method is presented of training qualified personnel for nondestructive testing of pressure vessels. Personnel is divided according to experience and previous training into three groups of which each has its own educational programme. Written examinations in general knowledge and in specialized subjects and a practical examination in crack detection terminate the training. (E.S.). 1 fig., 4 tabs., 3 refs

  11. 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.

  12. Design and Optimization of Filament Wound Composite Pressure Vessels

    NARCIS (Netherlands)

    Zu, L.

    2012-01-01

    One of the most important issues for the design of filament-wound pressure vessels reflects on the determination of the most efficient meridian profiles and related fiber architectures, leading to optimal structural performance. To better understand the design and optimization of filament-wound pres

  13. Design Guide for glass fiber reinforced metal pressure vessel

    Science.gov (United States)

    Landes, R. E.

    1973-01-01

    Design Guide has been prepared for pressure vessel engineers concerned with specific glass fiber reinforced metal tank design or general tank tradeoff study. Design philosophy, general equations, and curves are provided for safelife design of tanks operating under anticipated space shuttle service conditions.

  14. Vessel Elasticity Estimation by Normalized Blood Pressure Dynamics

    Czech Academy of Sciences Publication Activity Database

    Jurák, Pavel; Halámek, Josef; Vondra, Vlastimil; Leinveber, Pavel; Plachý, M.; Fráňa, P.; Souček, M.; Kára, T.

    Tel-Aviv : Israel Heart Society, 2008. s. 115. ISBN N. [IDSS 2008 - International Dead Sea Symposium on Cardiac Arrhythmias and Device Therapy /9./. 22.09.2008-24.09.2008, Tel-Aviv] Institutional research plan: CEZ:AV0Z20650511 Keywords : hypertension * vessel compliance * blood pressure * dynamic parameters Subject RIV: FA - Cardiovascular Diseases incl. Cardiotharic Surgery

  15. Interpretation of strain measurements on nuclear pressure vessels

    International Nuclear Information System (INIS)

    Selected results from strain measurements on 4 nuclear pressure vessels are presented and discussed. The measurements were made in several different regions of the vessels: transition zones in vessel heads, flanges and bottom parts, nozzels, internal vessel structure and flange bolts. The results presented are based on data obtained by approximately 700 strain-gauges, and a comprehensive knowledge of the quality obtained by such measurements is established. It is shown that a thorough control procedure before and after the test as well as detailed knowledge of the behaviour of the signal from the individual gauges during the test is necessary. If this is omitted, it can be extremely difficult to distinguish between the real structural behaviour and a malfunctioning of a specific gauge installation. In general, most of the measuring results exhibit a very linear behaviour with a negligible zeroshift. However, deviations from linear behaviour are observed in several cases. This nonlinearity can be explained by friction (flange connections) or by gaps (concentrical nozzles) in certain regions, whereas local plastic deformations during the first pressure loadings of the vessel seem to be the reason in other regions. (author)

  16. 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.

  17. Welding of the A1 reactor pressure vessel

    International Nuclear Information System (INIS)

    As concerns welding, the A-1 reactor pressure vessel represents a geometrically complex unit containing 1492 welded joints. The length of welded sections varies between 10 and 620 mm. At an operating temperature of 120 degC and a pressure of 650 N/cm2 the welded joints in the reactor core are exposed to an integral dose of 3x1018 n/cm2. The chemical composition is shown for pressure vessel steel as specified by CSN 413090.9 modified by Ni, Ti and Al additions, and for the welding electrodes used. The requirements are also shown for the mechanical properties of the base and the weld metals. The technique and conditions of welding are described. No defects were found in ultrasonic testing of welded joints. (J.B.)

  18. Assessment of residual life of TAPS pressure vessel

    International Nuclear Information System (INIS)

    SA302B (nickel modified) steel cladded with stainless steel is used as the pressure vessel material for the two 210 MWe boiling water reactors of the Tarapur Atomic Power Station. Charpy V-notch impact surveillance specimens representing the pressure vessel belt-line base, weld and the heat affected zone were irradiated at the wall and shroud locations. Some of these specimens were removed after 6.5 effective full power years (EFPY) of reactor operation. The neutron fluences at the locations were 5.31x1017 and 4.88x1018 n/cm2 (E > 1 MeV). The surveillance data generated from specimens removed after 6.5 EFPY were evaluated on the basis of USNRC Regulatory Guide 1.99, Revision 2, and the results had assured the integrity of the vessel beyond the end of design service life (EOL) of 40 years. The evaluation of the additional data generated from specimens removed after 13 EFPY has again confirmed the safety of the pressure vessel beyond EOL by an additional 20 EFPY. (author)

  19. Fabrication of toroidal composite pressure vessels. Final report

    International Nuclear Information System (INIS)

    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

  20. Modeling irradiation embrittlement in reactor pressure vessel steels

    International Nuclear Information System (INIS)

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 10, numerical modeling of irradiation embrittlement in reactor vessel steels are introduced. Physically-based models are developed and their role in advancing the state-of-the-art of predicting irradiation embrittlement of RPV steels is stressed

  1. Estimation of ex-vessel steam explosion pressure loads

    International Nuclear Information System (INIS)

    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.

  2. 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.

  3. Stable and unstable crack growth in pressure vessel models

    International Nuclear Information System (INIS)

    Three identical steel pressure vessels with 254-mm (10-in.) dia, 38-mm (1.5-in.) wall thicknesses and long, deep machined and sharpened axially oriented flaws were tested at three different temperatures. The vessels were assembled by electron-beam welding cylindrical sections with substantially different toughnesses due to different heat treatments. Crack extension initiated in relatively brittle sections, and the cracks extended both stably and unstably, depending on test temperature, toward the tougher sections where crack arrest did and did not occur. Charpy impact specimens and both slow-bend and dynamic precracked Charpy specimens were used for material characterization. The behavior of the vessels is described and related to the Charpy data

  4. Standard- and extended-burnup PWR [pressurized-water reactor] and BWR [boiling-water reactor] reactor models for the ORIGEN2 computer code

    International Nuclear Information System (INIS)

    The purpose of this report is to describe an updated set of reactor models for pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) operating on uranium fuel cycles and the methods used to generate the information for these models. Since new fuel cycle schemes and reactor core designs are introduced from time to time by reactor manufacturers and fuel vendors, an effort has been made to update these reactor models periodically and to expand the data bases used by the ORIGEN2 computer code. In addition, more sophisticated computational techniques than previously available were used to calculate the resulting reactor model cross-section libraries. The PWR models were based on a Westinghouse design, while the BWR models were based on a General Electric BWR/6 design. The specific reactor types considered in this report are as follows (see Glossary for the definition of these and other terms): (1) PWR-US, (2) PWR-UE, (3) BWR-US, (4) BWR-USO, and (5) BWR-UE. Each reactor model includes a unique data library that may be used to simulate the buildup and deletion of isotopes in nuclear materials using the ORIGEN2 computer code. 33 refs., 44 tabs

  5. 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.

  6. 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.

  7. 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.

  8. 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)

  9. Risk-informed appendices G and E for section XI of the ASME Boiler and Pressure Vessel Code

    Energy Technology Data Exchange (ETDEWEB)

    Carter, B; Spanner, J. [EPRI, (United States); Server, W. [ATI Consulting (United States); Gamble, R. [Sartrex Corporation (United States); Bishop, B.; Palm, N.; Heinecke, C. [Westinghouse Electric Company (United States)

    2011-07-01

    Full text of publication follows: The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, contains two appendices (G and E) related to reactor pressure boundary integrity. Appendix G provides procedures for defining Service Level A and B pressure temperature limits for ferritic components in the reactor coolant pressure boundary. Recently, an alternative risk informed methodology has been developed for ASME Section XI, Appendix G. The alternative methodology provides simple procedures to define risk informed pressure temperature limits for Service Level A and B events, including leak testing and reactor start up and shut down for both pressurized water reactors (PWRs) and boiling water reactors (BWRs). Risk informed pressure temperature limits provide more operational flexibility, particularly for reactor pressure vessels (RPV) with relatively high irradiation levels and radiation sensitive materials. Appendix E of Section XI provides a methodology for assessing conditions when the Appendix G limits are exceeded. A similar risk informed methodology is being considered for Appendix E. The probabilistic fracture mechanics evaluations used to develop the risk informed relationships included appropriate material properties for the range of RPV materials in operating plants in the United States and operating history and system operational constraints in both BWRs and PWRs. The analysis results were used to define pressure temperature relationships that provide an acceptable level of risk, consistent with safety goals defined by the U.S. Nuclear Regulatory Commission. The alternative methodologies for Appendices G and E will provide greater operational flexibility, especially for Service Level A and B events that may adversely affect efficient and safe plant operation, such as low temperature over pressurization for PWRs and BWR leak testing. Overall, application of the risk informed appendices can result in increased plant

  10. Fractographic study of a thick wall pressure vessel failure

    International Nuclear Information System (INIS)

    The pressure vessel described in this paper is identified as Intermediate Test Vessel 1 (ITV-1) and was fabricated of SA508, Class 2 Steel. It was tested to failure at 540C (1300F). The gross failure appeared to be a brittle fracture although accompanied by a measured strain of 0.9%. Seven regions of the fracture were examined in detail and the observed surfaces were compared to Charpy V-notch (C/sub v/) specimens of SA508, Class 2 steel broken at temperatures above and below the ductile to brittle transition temperature. Three samples from the vessel were taken in the region around the fatigue notch and four from areas well removed from the notch. All these were carefully examined both optically and by scanning electron microscopy (SEM). It was established that early crack extension was by ductile mode until a large flaw approximately 500 mm long 83 mm wide was developed. At this point the vessel could no longer contain the internal pressure and final rupture was by brittle fracture

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

  13. Evaluation of the performance of ultrasonic and eddy current testing of austenitic claddings of reactor pressure vessels

    International Nuclear Information System (INIS)

    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.)

  14. 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)

  15. 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 tibial artery 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. PMID:27079693

  16. Automatic flaw detecting device for reactor pressure vessel

    International Nuclear Information System (INIS)

    The device of the present invention is used for detecting flaws in welded portions of a pressure vessel while running along a track formed at the circumference of the pressure vessel, by which cables for transmitting collected data to a detection chamber can be laid around easily. Namely, a power supply conductor for supplying electric power to the device and a leakage coaxial cable as a receiving antenna are disposed to the tracks along the longitudinal direction. A connection terminal is disposed to the automatic flaw detecting device to be in contact with the power supply conductor. In addition, an oscillator is disposed for transmitting signals collected by a supersonic probe to the leakage coaxial cable by way of a transmission antenna. Then, since cables drawn from the automatic flaw detecting device are used only for transporting air and water to the flaw detection portion, the cables can be laid easily to improve the operationability. (I.S.)

  17. Heritability of retinal vessel diameters and blood pressure

    DEFF Research Database (Denmark)

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

    2006-01-01

    , aged 20 to 46 years, interpolated diameter estimates for the central retinal artery (CRAE), the central retinal vein (CRVE), and the artery-to-vein diameter ratio (AVR) were assessed by analysis of digital gray-scale fundus photographs of right eyes. RESULTS: The heritability was 70% (95% CI: 54...... for CRVE, and 0.67 +/- 0.05 microm for AVR. No significant influence on artery or vein diameters was found for gender, smoking, body mass index (BMI), total cholesterol, fasting blood glucose, or 2-hour oral glucose tolerance test values. CONCLUSIONS: In healthy young adults with normal blood pressure...... 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 and the...

  18. 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.

  19. Backstreaming of Impurity Gas Through a Leak in Pressurized Vessel

    CERN Document Server

    Dauvergne, J P

    1998-01-01

    The presence of a leak in a vessel containing pure gas can induce the contamination by atmospheric gas diffusing into the vessel. In order to avoid this, a gas which has to be kept pure also in presen ce of a leak is usually pressurized, to reduce the flow of contaminating gas through the leak owing to the molecular drag by the outstreaming pure gas. In this paper, a simple model calculation of ba ckstreaming based on the solution of the diffusion + drag equation in cylindrical coordinates is presented. It is shown that both the pressure difference and the dimension of the leak are critical in determining the contaminating flow, a maximum in the backstreaming flow appearing when the drag velocity of the outstreaming gas equals the diffusion velocity.

  20. Monitoring test piece for a reactor pressure vessel

    International Nuclear Information System (INIS)

    Purpose: To obtain a test piece capable of measurement for neutron exposure ranging 0.1 -- 2 MeV in a reactor pressure vessel Constitution: Fissionable materials causing nuclear fission by fast neutrons are contained within a sealed container in addition to a test piece for monitoring the change in the mechanical properties and a monitor wire for measuring the neutron dose. If uranium 238 and thorium 232 are selected as the fissionable materials, for instance, they cause nuclear fission by the reaction with neutrons of higher than about 2 MeV and about 0.2 MeV respectively. Then, after the stop of the reactor operation, the monitoring test piece is taken out from the reactor pressure vessel to determine the radioactivity, whereby the neutron dose within the energy range of 0.1 - 2 MeV applied to the fissionable materials of the test piece can be estimated with ease. (Horiuchi, T.)

  1. Glass Fiber Reinforced Metal Pressure Vessel Design Guide

    Science.gov (United States)

    Landes, R. E.

    1972-01-01

    The Engineering Guide presents curves and general equations for safelife design of lightweight glass fiber reinforced (GFR) metal pressure vessels operating under anticipated Space Shuttle service conditions. The high composite vessel weight efficiency is shown to be relatively insensitive to shape, providing increased flexibility to designers establishing spacecraft configurations. Spheres, oblate speroids, and cylinders constructed of GFR Inconel X-750, 2219-T62 aluminum, and cryoformed 301 stainless steel are covered; design parameters and performance efficiencies for each configuration are compared at ambient and cryogenic temperature for an operating pressure range of 690 to 2760 N/sq cm (1000 to 4000 psi). Design variables are presented as a function of metal shell operating to sizing (proof) stress ratios for use with fracture mechanics data generated under a separate task of this program.

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

    International Nuclear Information System (INIS)

    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)

  3. Innovations on life management of VVER reactor pressure vessels

    International Nuclear Information System (INIS)

    The embrittlement rate of the pressure vessel weld material is a dominating factor in life management of VVER reactor pressure vessels. In order to maintain adequate safety level, several backfitting measures have been performed in Loviisa. The neutron flux and embrittlement rate was reduced after receiving the first indications of anticipated problems. An increase of emergency core cooling water temperature and other process related changes followed to eliminate and reduce potential transients. Finally, the core weld of Loviisa 1 was successfully annealed in 1996. A current concern is to verify the post-annealing embrittlement rate in order to enable safe and economic life management of the RPV. Post-annealing re-embrittlement is governed by somewhat different mechanisms than the embrittlement of the first irradiation cycle. A new tentative approach for predicting the re-embrittlement rate has been proposed. (orig.)

  4. 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.

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

    International Nuclear Information System (INIS)

    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.

  6. Large deflection analysis of conical head pressure vessels

    International Nuclear Information System (INIS)

    The present work investigates the nonlinear behaviour at and near the junction of cylindrical and conical shells under internal or external pressure. The results are useful for studying conical head pressure vessels and components of missiles, spacecrafts, underwater vessels, nuclear reactors, etc. Reissner's basic concept for the large deformation of axisymmetric shells, with proper modification and specialization, are used in the derivation of the governing equations for thepresent problem. The nonlinear differential equations are then solved by the multisegment method of integration developed by Kalnins and Lestingi. Results of this analysis indicate that the linear theory is very inadequate and conservative in its prediction of stresses and deformations in the region of the junction of the cylindrical and conical parts. In this analysis the tip of the conical part, where present, is assumed to be replaced by a very small spherical cap meeting tangentially with the conical part, in order to avoid singularity and to render the solution more realistic. (orig.)

  7. Radiation embrittlement in pressure vessels of power reactors

    International Nuclear Information System (INIS)

    It is presented the project to study the effect of lead factors on the mechanical behavior of Reactor Pressure Vessel steels. It is described the facility designed to irradiate Charpy specimens with V notch of SA-508 type 3 steel at power reactor temperature, installed in the RA-1 reactor. The objective is to obtain the fracture behavior of irradiated specimens with different lead factors and to know their dependence with the diffusion of alloy elements. (author)

  8. Two benchmarks for qualification of pressure vessel fluence calculational methodology

    International Nuclear Information System (INIS)

    Two benchmarks for the qualification of the pressure vessel fluence calculational methodology were formulated and are briefly described. The Pool Critical Assembly (PCA) benchmark is based on the experiments performed at the PCA in Oak Ridge. The measured quantities to be compared against the calculated values are the equivalent fission fluxes at several locations in front, behind, and inside the pressure-vessel wall simulator. This benchmark is particularly suitable to test the capabilities of the calculational methodology and cross-section libraries to predict in-vessel gradients because only a few approximations are necessary in the analysis. The HBR-2 benchmark is based on the data for the H.B. Robinson-2 plant, which is a 2,300 MW (thermal) pressurized light-water reactor. The benchmark provides the reactor geometry, the material compositions, the core power distributions, and the power historical data. The quantities to be calculated are the specific activities of the radiometric monitors that were irradiated in the surveillance capsule and in the cavity location during one fuel cycle. The HBR-2 benchmark requires modeling approximations, power-to-neutron source conversion, and treatment of time dependant variations. It can therefore be used to test the overall performance and adequacy of the calculational methodology for power-reactor pressure-vessel flux calculations. Both benchmarks were analyzed with the DORT code and the BUGLE-96 cross-section library that is based on ENDF/B-VI evaluations. The calculations agreed with the measurements within 10%, and the calculations underpredicted the measurements in all the cases. This indicates that the ENDF/B-VI cross sections resolve most of the discrepancies between the measurements and calculations. The decrease of the CIM ratios with increased thickness of iron, which was typical for pre-ENDF/B-VI libraries, is almost completely removed

  9. Ultrasonic testing of electron beam closure weld on pressure vessel

    International Nuclear Information System (INIS)

    One of the special products manufactured at the General Electric Neutron Devices Department (GEND) is a small stainless steel vessel designed to hold a component under high pressure for long periods. The vessel is a thick-walled cylinder with a threaded receptacle into which a plug is screwed and welded after receiving the unit to be tested. The test cavity is then pressurized through a small diameter opening in the bottom and that opening is welded closed. When x-ray inspection techniques did not reveal defective welds at the threaded plug in a pressured vessel, occasional ''leakers'' occurred. With normal equipment tolerances, the electron beam spike tends to wander from the desired path, particularly at the root of the weld. Ultrasonic techniques were used to successfully inspect the weld. The testing technique is based on the observation that ultrasonic energy is reflected from the unwelded screw threads and not from the regions where the threads are completely fused together by welding. Any gas pore or any threaded region outside the weld bead can produce an echo. The units are rotated while the ultrasonic transducer travels in a direction parallel to the axis of rotation and toward the welded end. This produces a helical scan which is converted to a two-dimensional presentation in which incomplete welds can be noted. (U.S.)

  10. Probabilistic study of PWR reactor pressure vessel fracture

    International Nuclear Information System (INIS)

    Different methods are used to evaluate the rupture probability of a nuclear pressure vessel. On of them extrapolates to nuclear pressure vessels, data of failure found in conventional pressure vessels. The disadvantage of such an approach is that the effects of systematic changes in key parameters cannot be taken into account. For example, the influence of irradiation and the use of quality assurance programs encompassing design, fabrication and materials cannot be considered. But the most important disadvantage of this method is the limited size of the representative population and consequently the high value of the upper bound failure rate corresponding to a requested confidence level. The method used in the present work involves the development of physical models based on an understanding of the failure modes and expressing the conventional concepts of fracture mechanics in a probabilistic form; the fatigue crack growth rate, calculated for conditions of cyclic loading, the initiation of unstable crack propagation, and the possibility of crack arrest. The analysis therefore requires the statistical expression of the factors and parameters which appear in the expressions of the law of crack growth and of toughness, and also those which are used in the calculation of the stress intensity factor K1. All input data are entered in COVASTOL code in histogram form. This code takes into account the degree of correlation between the flaw size and the Paris' law coefficients. It computes the propagation of a given defect in a given position, and the corresponding failure probability during accidental loading

  11. Investigation of reactor pressure vessel steels after radiation degradations

    International Nuclear Information System (INIS)

    The degradation of reactor pressure vessel (RPV) steel is a complex process depending on many factors (thermal and radiation treatment, chemical compositions, preparing conditions, ageing, operation environment, etc.). This paper describes tests based on Nondestructive Methods used for evaluation of material characterisation at Slovenske Elektrarne a.s. Positron Annihilation Spectroscopy (PAS), Moessbauer Spectroscopy (MS) and Transmission Electron Microscopy (TEM) investigate microstructure changes of Reactor Pressure Vessel steels caused neutrons irradiation. There are showed results of investigation of reactor pressure vessel steel specimens after five years irradiation in reactor nature by mentioned NDT methods. Investigated specimens has been prepared for the Extended surveillance specimen program which has run out on the 3rd and 4th units of NPP Jaslovske Bohunice and for the Modified surveillance specimen program in the 1st and 2nd unit which is continued in NPP Mochovce (Slovakia). PAS and MS spectra showed that the degradation of the steel properties associated with the effects of neutron irradiation can be well detected. The samples from RPV base metal (15Kh2MFA) and weld metal (Sv 10KhMFT) were measured by PAS and MS before and after irradiation. Samples have been irradiated in VVER-440 reactor (units 3rd and 4th in Bohunice as well as 1st and 2nd units Mochovce) by neutron fluency from 7.8 1023 m-2 up to 2.5 1024 m-2. Measurement results are presented and discussed in detail.(author)

  12. Prevention of catastrophic failure in pressure vessels and pipings

    International Nuclear Information System (INIS)

    The fracture resistance and integrity of pressure-loaded components have been assessed in a Nordic research programme. Experiments were performed to validate the computational fracture assessment analysis. Two tests were also conducted on a large decommissioned pressure vessel from an oil refinery plant. Different fracture assessment methods were developed and subsequently applied to the tested components. Interlaboratory round robin programmes with the participation of several laboratories were arranged to examine elastic-plastic finit element calculations and fracture mechanics testing. The transferability of material parameters derived from small specimens with simple crack geometries to more realistic crack geometries in real components has been verified. (author)

  13. Monitoring of water level inside reactor pressure vessel

    International Nuclear Information System (INIS)

    Up to the TMI accident the water level inside the pressurizer was used to monitor the water inventory inside the primary cooling system of pressurized water reactors. The TMI accident showed that this was not a reliable measurement for the reactor coolant inventory inside the reactor pressure vessel. For this reason there was a demand for a measurement of the water level inside the RVP, independent from the existing one inside the pressurizer and with a diverse measuring method. For WWER reactors a new level measurement system was developed to monitor the water level inside the reactor pressure vessel by means of the KNITU, resp. KITU level probe which meet all the mentioned engineered safeguards and geometric and constructive requirements. First backfitting s of the new level measurement system in the WWER s 440 in Bohunice V1 (Slovakia), unit 1 (1998) and unit 2 (2000), Novovoronezh (Russia), unit 4 (1999) and Kola (Russia), unit 1 and unit 2 (1999) show very good operational results. (Authors)

  14. 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.

  15. Dual shell pressure balanced reactor vessel. Final project report

    International Nuclear Information System (INIS)

    The Department of Energy's Office of Energy Research (OER) has previously provided support for the development of several chemical processes, including supercritical water oxidation, liquefaction, and aqueous hazardous waste destruction, where chemical and phase transformations are conducted at high pressure and temperature. These and many other commercial processes require a pressure vessel capable of operating in a corrosive environment where safety and economy are important requirements. Pacific Northwest Laboratory (PNL) engineers have recently developed and patented (U.S. patent 5,167,930 December 1, 1992) a concept for a novel Dual Shell Pressure Balanced Vessel (DSPBV) which could solve a number of these problems. The technology could be immediately useful in continuing commercialization of an R ampersand D 100 award-winning technology, Sludge-to-oil Reactor System (STORS), originally developed through funding by OER. Innotek Corporation is a small business that would be one logical end-user of the DSPBV reactor technology. Innotek is working with several major U.S. engineering firms to evaluate the potential of this technology in the disposal of wastes from sewage treatment plants. PNL entered into a CRADA with Innotek to build a bench-scale demonstration reactor and test the system to advance the economic feasibility of a variety of high pressure chemical processes. Hydrothermal processing of corrosive substances on a large scale can now be made significantly safer and more economical through use of the DSPBV. Hydrothermal chemical reactions such as wet-air oxidation and supercritical water oxidation occur in a highly corrosive environment inside a pressure vessel. Average corrosion rates from 23 to 80 miles per year have been reported by Rice (1994) and Latanision (1993)

  16. 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.).

  17. Neutron fluence at the pressure vessel of a pressurized water reactor determined by the MCNP code

    International Nuclear Information System (INIS)

    Pressure vessel fluence and reaction rates for dosimetry foils in the cavity surrounding the pressure vessel of a pressurized water reactor were determined with a Monte Carlo calculation using the MCNP code. Source neutrons were sampled from a position probability distribution derived from the utility-provided normalized assembly segment power output. The MCNP model was based on one-eighth core symmetry. Source segment spatial biasing, energy cutoff, spatial importance functions, and weight windows were employed as variance reduction techniques. Computed reaction rates were compared with measured ones and in one case to discrete ordinates transport code calculations. Computed reaction rates matched the measured ones within ±10% for 21 of 33 cases and within ±15% for 26 of 33 cases. Neutron flux and fluence >0.1111 and 1 MeV at the pressure vessel location were computed to 17 n/cm2

  18. 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...

  19. Calculation of reactor pressure vessel fluence using TORT code

    International Nuclear Information System (INIS)

    TORT is employed for fast neutron fluence calculation at the reactor pressure vessel. KORI Unit 1 reactor at cycle 1 is modeled for this calculation. Three-dimensional cycle averaged assembly power distributions for KORI Unit 1 at cycle 1 are calculated by using the core physics code, NESTLE 5.0. The root mean square error is within 4.3% compared with NDR (Nuclear Design Report) for all burnup steps. The C/E (Calculated/Experimental) values for the in-vessel dosimeters distribute between 0.98 and 1.36. The most updated cross-section library, BUGLE-96 based on ENDF/B-VI is used for the neutron fluence calculation. The maximum fast neutron fluence calculated on reactor pressure vessel for KORI Unit 1 operated for 411.41 effective full power days is 1.784x1018n/cm2. The position of the maximum neutron fluence in RPV wall 1/4 T is nearby 60 cm below the midplane at zero degree

  20. Pressurization of Containment Vessels from Plutonium Oxide Contents

    International Nuclear Information System (INIS)

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

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

    DEFF Research Database (Denmark)

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

    2015-01-01

    analysis (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......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...... product development, design and optimization, as well as to minimize the risks and improve the public acceptance. Within the scope of developing different COPV models for a wide range of operating pressures and applications, optical fiber Bragg grating (FBG) sensors were embedded in the liner-composite...

  2. Fracture assessment of a BWR pump nozzle

    International Nuclear Information System (INIS)

    Fracture mechanics calculations are performed to support the non-destructive testing (NDT) qualification programs for pump nozzle investigations of boiling water reactor (BWR) nozzles of reactor pressure vessels (RPVs), with the aim of the determination of qualification defects, which are located in the Inconel 182 weld of the pump nozzle at the bottom of the RPV. The ferritic nozzle and housing have an Inconel buttering and each part is cladded with Inconel 182 before it is mounted. All theses weldments are heat treated after welding; only the connecting weldment between pump housing and nozzle, which is also an Inconel 182 weld, performed on site, is in the as welded condition. (author)

  3. 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)

  4. Results of fuel, cladding and reactor pressure vessel computation

    International Nuclear Information System (INIS)

    The cause of the pellet-clad interaction failures are the combined effects of differential thermal expansion driven localized stress, and aggressive fission products, primarily iodine (pitting corrosion). Reactor pressure vessel is also being exposed to high mechanical loads and to an intensive neutron flux irradiation. This leads to a gradual decrease in resistance to brittle fracture. The finite element method has been selected for the solution of the above problems, and results are presented in form of isobars and isotherms respectively in the meridional cross section. (author)

  5. Computing radiation dose to reactor pressure vessel and internals

    International Nuclear Information System (INIS)

    Within the next twenty years many of the nuclear reactors currently in service will reach their design lifetime. One of the key factors affecting decisions on license extensions will be the ability to confidently predict the integrity of the reactor pressure vessel and core structural components which have been subjected to many years of cumulative radiation exposure. This report gives an overview of the most recent scientific literature and current methodologies for computational dosimetry in the OECD/NEA Member countries. Discussion is extended to consider some related issues of materials science, such as the metals, and limitations of the models in current use. Proposals are made for further work. (author)

  6. A high temperature reactor accommodated in a cylindrical pressure vessel

    International Nuclear Information System (INIS)

    This concerns the mixing of hot gases of a high temperature reactor with spherical fuel elements which is placed in a cylindrical pressure vessel. The cooling gas is blown in from above and after passing through the pebble bed it is collected on the bottom where the graphite blocks form a kind of pillard hall. According to this invention, in front of each channel through which the hot gas is blown off there is placed a specially formed replacement body for deviation purpose, and thus the hot gases are thoroughly mixed. Like this the gas flow gets to the components of the primary coolant circuit with a regular temperature. (GL)

  7. Aging and Life Management System of Reactor Pressure Vessel

    OpenAIRE

    Ya-jin Liu; Jiang Guo; Kai-kai Gu

    2011-01-01

    Reactor pressure vessel (RPV), the only key component that can not be replaced in nuclear power plants (NPPs), is the main barrier against the radioactive leakage. The lifetime of NPPs is dependent heavily on the life of RPV, and thus, the aging and life research on a RPV is a key factor in determining the life extension of NPPs. The purpose of this paper is to introduce an aging and life management system for an operating RPV which can be used as a reference of the lifetime extension. In ord...

  8. Automated ultrasonic shop inspection of reactor pressure vessel forgings

    International Nuclear Information System (INIS)

    Automated ultrasonic shop inspection utilizing a computer-controlled system is being applied to each of the forgings for the reactor pressure vessel of the proposed Sizewell B PWR power station. Procedures which utilize a combination of high sensitivity shear wave pulse echo, 0 degrees and 70 degrees angled longitudinal waves, tandem and through-thickness arrays have been developed to provide comprehensive coverage and an overall reliability of inspection comparable to the best achieved in UKAEA defect detection trials and in PISC II. This paper describes the ultrasonic techniques, the automated system (its design, commissioning and testing), validation and the progress of the inspections

  9. Electrode for welding steel for WWER-1000 reactor pressure vessel

    International Nuclear Information System (INIS)

    Of two types of electrodes, ie., with an alloyed core and with an unalloyed core, an electrode was chosen consisting of a basic coat and an unalloyed core. Fluctuations are shown of shear strength, tensile strenght and contraction with the welding mode and annealing temperature. It was found that pre-heating to 250 and 350 degC, respectively, was most suitable for welding a pressure vessel manufactured from material designated SKODA A3/II. Annealing aimed at removing stress was chosen at 650 to 700 degC. (H.S.)

  10. Metallurgy of steels for nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Technology is described of ingot production for pressure vessel parts consisting in the preparation of the alloying master alloy in an electric arc furnace and of a low-alloyed rimmed steel in an open-hearth furnace. The steels are cast into the ladle and are then vacuum cast into a mould. For securing the desired purity, the preparation of pre-molten charge, careful selection of alloying and refining admixtures and stringent observance of the prescribed technology are necessary. The said method allows producing ingots up to 195 tons in weight, showing the desired properties. This was confirmed by tests of the chemical composition and of mechanical properties. (Ha)

  11. High power laser interaction effects with metallic pressure vessel

    Science.gov (United States)

    Das, Nilratan; Mukherji, D.; Kumar, R.; Husain, M.; Zaidi, Z. H.; Kumar, Anil

    2006-05-01

    The paper describes the theoretical and experimental investigation of high power Gas Dynamic Laser(GDL, λ~10.6μm) interaction studies with a pressurized hollow metal(MS) target. The design and development of such type of target which has been shown bursting as well as burning effect at the time of interaction have been carried out. It has been filled by gas mixture of H II and Air in the range of flammability limit. Various parameters like power density, target thickness, filling pressure, mixture ratio etc have been optimized. High mass flow GDL of power level about several KW in unstable mode provides power density about 3.2 KW/Cm2 by a beam delivery system at distance 25m. Since target material is thin and heat diffuses through it rapidly, by maintaining the required power density, rupturing is accomplished by heating an area of the pressure vessel to a temperature at which it will fail under the pressure load. Rupture initiates a propagating crack which spreads the damage over a large fraction of the pressure vessel. The gas mixture ignites due to its contact with atmosphere and explodes with a massive sound level of the order of 130dB. The sound level was measured by a Decibel meter. Temperature distribution along radial and depth have been studied theoretically. Surface temperature during interaction has been measured. Experimental data has been validated with theory. These study shows a very attractive demonstration showing potentiality of scientific applications of High Power CO II Laser.

  12. 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.

  13. 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.

  14. 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.

  15. Test of 6-inch-thick pressure vessels. Series 2. Intermediate test vessels V-3, V-4, and V-6

    International Nuclear Information System (INIS)

    The second series of intermediate vessel tests were crack initiation fracture tests of 6-in.-thick 39-in.-OD steel vessels with sharp surface flaws approximately 21/2 in. deep by 8 in. long in the longitudinal weld seams of the test cylinders. Fracture was initiated by means of hydraulic pressurization. One vessel was tested at each of three temperatures: 75, 130, and 1900F. Pretest analyses were made to predict the failure pressures and strains. Fracture toughness data obtained by equivalent-energy analysis of precracked Charpy-V tests and compact-tension specimen tests were used in the fracture analyses. The vessels behaved generally as had been expected. Posttest fracture analyses were also performed for each vessel. Detailed discussions of the fracture analysis methods developed in support of the vessel tests described are included. 34 references

  16. Finite element analysis of filament-wound composite pressure vessel under internal pressure

    Science.gov (United States)

    Sulaiman, S.; Borazjani, S.; Tang, S. H.

    2013-12-01

    In this study, finite element analysis (FEA) of composite overwrapped pressure vessel (COPV), using commercial software ABAQUS 6.12 was performed. The study deals with the simulation of aluminum pressure vessel overwrapping by Carbon/Epoxy fiber reinforced polymer (CFRP). Finite element method (FEM) was utilized to investigate the effects of winding angle on filament-wound pressure vessel. Burst pressure, maximum shell displacement and the optimum winding angle of the composite vessel under pure internal pressure were determined. The Laminae were oriented asymmetrically for [00,00]s, [150,-150]s, [300,-300]s, [450,-450]s, [550,-550]s, [600,-600]s, [750,-750]s, [900,-900]s orientations. An exact elastic solution along with the Tsai-Wu, Tsai-Hill and maximum stress failure criteria were employed for analyzing data. Investigations exposed that the optimum winding angle happens at 550 winding angle. Results were compared with the experimental ones and there was a good agreement between them.

  17. Finite element analysis of filament-wound composite pressure vessel under internal pressure

    International Nuclear Information System (INIS)

    In this study, finite element analysis (FEA) of composite overwrapped pressure vessel (COPV), using commercial software ABAQUS 6.12 was performed. The study deals with the simulation of aluminum pressure vessel overwrapping by Carbon/Epoxy fiber reinforced polymer (CFRP). Finite element method (FEM) was utilized to investigate the effects of winding angle on filament-wound pressure vessel. Burst pressure, maximum shell displacement and the optimum winding angle of the composite vessel under pure internal pressure were determined. The Laminae were oriented asymmetrically for [00,00]s, [150,-150]s, [300,-300]s, [450,-450]s, [550,-550]s, [600,-600]s, [750,-750]s, [900,-900]s orientations. An exact elastic solution along with the Tsai-Wu, Tsai-Hill and maximum stress failure criteria were employed for analyzing data. Investigations exposed that the optimum winding angle happens at 550 winding angle. Results were compared with the experimental ones and there was a good agreement between them

  18. Pressure vessel codes: Their application to nuclear reactor systems

    International Nuclear Information System (INIS)

    A survey has been made by the International Atomic Energy Agency of how the problems of applying national pressure vessel codes to nuclear reactor systems have been treated in those Member States that had pressurized reactors in operation or under construction at the beginning of 1963. Fifteen answers received to an official inquiry form the basis of this report, which also takes into account some recently published material. Although the answers to the inquiry in some cases data back to 1963 and also reflect the difficulty of describing local situations in answer to standard questions, it is hoped that the report will be of interest to reactor engineers. 21 refs, 1 fig., 2 tabs

  19. BWR internal cracking issues

    International Nuclear Information System (INIS)

    The regulatory issues associated with cracking of boiling water reactor (BWR) internals is being addressed by the Nuclear Regulatory Commission (NRC) staff and is the subject of a voluntary industry initiative. The lessons learned from this effort will be applied to pressurized water reactor (PWR) internals cracking issues

  20. 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.

  1. Reassessment of PWR pressure-vessel integrity during overcooling accidents

    International Nuclear Information System (INIS)

    A continuing analysis of the PTS problem associated with PWR postuated OCA's indicates that the previously accepted degree of conservatism in the fracture-mechanics model needs to be more closely evaluated, and if excessive, reducted. One feature that was believed to be conservative was the use of two-dimensional as opposed to finite-length (three-dimensional) flaws. A flaw of particular interest is one that is located in an axial weld of a plate-type vessel. For those vessels that suffer relatively high radiation damage in the welds, the length of the flaw will be no greater than the length of the weld, and recent calculations indicate that a deep flaw of that length (approx. 2 m) is not effectively infinitely long, contrary to previous thinking. The benefit to be derived from consideration of the 2-m flaw and also a semielliptical flaw with a length-to-depth ratio of 6/1 was investigated by analyzing several postulated transients. In doing so the sensitivity of the benefit to a specified maximum crack arrest toughness and to the duration of the transient was investigated. Results of the analysis indicate that for some conditions the benefit in using the 2-m flaw is substantial, but it decreases with increasing pressure, and above a certain pressure there may be no benefit, depending on the duration of the transient and the limit on crack arrest toughness

  2. Cladding of pressure vessel steel with corrosion resistant filler material

    International Nuclear Information System (INIS)

    Pressure vessels are often on the inside clad with corrosion resistant material. Of the various cladding processes surfacing by welding has proved to be most useful, especially for large thick-walled pressure vessels. Submerged arc welding with strip electrode is the most common method. Rather promising results have also been obtained by plasma hot wire welding. In general, Nb-alloyed austenitic stainless steel, over-alloyed with Cr and Ni, is used as filler material. Henceforth, also nickel alloys, e.g. Inconel 600, are used. The surfacing is made in one or several layers, following the requirements on the clad surface and the welding process used. The most dangerous welding defects in the surface are various types of cracks. The corrosion resistance of the cladding can show rather high local variations, depending on the composition of the filler material and various welding process factors. It is proved that the surface layer comparises areas with low chromium martensite. To ensure the corrosion resistance of the cladding, the generation of low-chromium martensite must be prevented by using suitable welding parameters, welding equipment and filler metal. It is also possible to eliminate the negative influence on the corrosion resistance from the low-chromium martensite, e.g. by welding in two layers. In the case of the high demands on quality a welding procedure test should always be made prior to production welding.(author)

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

    International Nuclear Information System (INIS)

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

  4. Microstructure and embrittlement of VVER 440 reactor pressure vessel steels

    International Nuclear Information System (INIS)

    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)

  5. Remote controlled stud bolt handling device for reactor pressure vessel

    International Nuclear Information System (INIS)

    In nuclear power stations, at the time of regular inspection, the works of opening and fixing the upper covers of reactor pressure vessels are carried out for inspecting the inside of reactor pressure vessels and exchanging fuel rods. These upper covers are fastened with many stud bolts, therefore, the works of opening and fixing require a large amount of labor, and are done under the restricted condition of wearing protective clothings and masks. Babcock Hitachi K.K. has completed the development of a remotely controlled automatic bolt tightenig device for this purpose, therefore, its outline is reported. The conventional method of these works and the problems in it are described. The design of the new device aimed at the parallel execution of cleaning screw threads, loosening and tightening nuts, and taking off and putting on nuts and washers, thus contributing to the shortening of regular inspection period, the reduction of the radiation exposure of workers, and the decrease of the number of workers. The function, reliability and endurance of the new device were confirmed by the verifying test using a device made for trial. The device is composed of a stand, a rail and four stations each with a cleaning unit, a stud tensioner and a nut handling unit. (K.I.)

  6. Fatigue of weldments in nuclear pressure vessels and piping

    International Nuclear Information System (INIS)

    Current (ASME) Code fatigue design rules for nuclear pressure vessels and piping include no special considerations for weldments other than purely geometric factors. Research programs aimed at nonnuclear applications have found weldments to display fatigue behavior inferior to that of pure base material. Available information on fatigue of weldments relevant to nuclear pressure vessels and piping was reviewed and determined changes in the current design rules appear to be dictated by the available information. Information was obtained and summarized and stored in a computerized data management system to facilitate correlation of facts and development of conclusions. Significant areas where development of additional data would substantially increase the ability to judge the adequacy of the current ASME design rules include: a better understanding of the relative importance of crack initiation and crack propagation to fatigue life; additional fatigue data for prototypic commercial weldments, including cumulative damage; properties of repair welds; significance of reheat cracks; quantitative effect of Code-allowable weld defects; and the effect of variable microstructure across the weld joint. Based on the information that is available, there is no evidence that the ASME Code fatigue design procedures need to be changed at this time. The current ASME design procedures, which form the general basis for fatigue evaluation both in the US and abroad are reviewed. Included is a review of various factors that influence the fatigue of weldments and of service experience with nuclear systems regarding fatigue of weldments. Research programs that may contribute to available information are reviewed

  7. Inspecting nuclear pressure vessels: the conundrum of minimizing risk

    International Nuclear Information System (INIS)

    The probability of a sudden, massive release of radioactivity from a light-water nuclear reactor through a breach of the containment is assessed on the basis of statistical data which partly consist of subjective estimates. This breach refers to the existence of crack-like defects remaining after a non-destructive examination of the main pressure vessel surrounding the reactor core. Two studies have recently been made of such sources of information about the effectiveness of non-destructive examination of pressure vessels with respect to defects. The results of these studies indicate that the data used as input in the probabilistic calculations do not possess the reliability that might be assumed from the assessments. This type of failure should therefore no longer be considered a de minimis case. In the present review the overconfidence in the efficiency of non-destructive examination is discussed from psychological, sociological and political science points of view. It is concluded that ingrained professional assumptions and values seem to be the main reason for the trust in the technology of inspection. However, there are also psychological constraints that can be understood only in their social and political contexts. (author)

  8. Reference fracture toughness for irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Transition temperature shift effects due to neutron radiation embrittlement for ferritic nuclear pressure vessel steels are currently evaluated using changes in the Charpy V-notch (CVN) energy curve; i.e., the shift in the 30 ft-lb (41 J) energy level. Estimates of the 30 ft-lb (41 J) transition temperature shifts (including margins for uncertainty) are often utilized based upon Nuclear Regulatory Commission Regulatory Guide 1.99, Revision 2, which includes the effects of material chemistry and fluence exposure. The estimated or measured Charpy shift is then applied to a lower bound reference (KIR) curve moving the curve the same shift amount but leaving the shape of the curve unaltered. Similarly, the flaw evaluation procedures in nonmandatory Appendix A of Section XI of the ASME Boiler and Pressure Vessel Code utilize the shifts in the equivalent of the KIR curve (termed the KIa curve for crack arrest) and a lower bound static crack initiation toughness (KIc) curve. This approach has been reviewed and tested as well as a reference toughness method for estimating statistically-based tolerance bounds. Comparisons of actual, but limited, fracture toughness data and the predicted bounding curves indicate that the shifted KIR/KIc curves are conservative in all cases. The reference toughness approach for 95%-95% tolerance bounds is not as conservative as the Regulatory and ASME Code method and may provide a more realistic bounding method. (author)

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

    International Nuclear Information System (INIS)

    The corrosion behavior of a high-strength steel [Specifications for Uncoated Seven-Wire-Stress-Relieved Strand for Prestressed Concrete (ASTM A 416-74, Grade 270)], typical of those used as tensioning tendons in prestressed concrete pressure vessels was measured in several corrosive environments. The protection obtained by coating the steel with two commercial petroleum-base greases or with Portland cement grout was evaluated. The few reported incidents of prestressing steel failures in concrete pressure vessels used for containment of nuclear reactors were reviewed. The susceptibility of the steel to stress corrosion cracking and hydrogen embrittlement and its general corrosion rate were determined in several salt solutions. Wires coated with the greases and grout were soaked for long periods in the same solutions and changes in their mechanical properties were subsequently determined. All three coatings appeared to give essentially complete protection; however, flaws in the grease coatings could be detrimental, and flaws or cracks less than 1-mm-wide (0.04 in.) in the grout were without effect

  10. 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 78.33-1 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) PASSENGER VESSELS OPERATIONS Reports of Accidents, Repairs, and Unsafe Equipment § 78.33-1 Repairs of boiler and pressure vessels. (a) Before making any repairs to boilers...

  11. Development of internal CRD for next generation BWR

    International Nuclear Information System (INIS)

    In order to develop a competitive and high performance Next Generation BWR with fossil power plant, an internal CRD using a heatproof ceramics insulated coil is under development. In case of a 1700MWe next generation BWR, the internal CRDs are installed in a RPV whose size is equivalent to the 1356 MWe ABWR, and there will be no space required for CRDs and CRD exchange under RPV. These advantages realize a compact PCV and reduced volume of a reactor building. Moreover, the internal CRDs eliminate penetration via a bottom flange of RPV, and lower installation level of RPV in a drywell. This brings further advantages of elimination of RIA (Reactivity Induced Accidents) caused by CR withdrawing under pressure boundary broken, and easy IVR (In Vessel Retention) by vessel bottom cooling in case of a severe accidents. (author)

  12. Development of internal CRD for next generation BWR

    International Nuclear Information System (INIS)

    In order to develop a competitive and high performance Next Generation BWR with fossil power plant, an internal CRD using a heatproof ceramics insulated coil has been developed. In case of a 1700MWe next generation BWR, the internal CRDs are installed in a RPV whose size is equivalent to the 1356 MWe ABWR, and there will be no space required for CRDs and CRD exchange under RPV. These advantages realize a compact PCV and reduced volume of a reactor building. Moreover, the internal CRDs eliminate penetration via a bottom flange of RPV, and lower installation level of RPV in a drywell. This brings further advantages of elimination of RIA (Reactivity Induced Accidents) caused by CR withdrawing under pressure boundary broken, and easy IVR (In Vessel Retention) by vessel bottom cooling in case of severe accidents. (authors)

  13. 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.

  14. Hydrogen attack of thick section pressure vessel steels

    International Nuclear Information System (INIS)

    The thick section pressure vessel steels, A387 2.25Cr-1Mo; A533B Mn-Mo-Ni steel and modifications containing MnNiCr and MnCrSi respectively, have been exposed to hydrogen atmospheres at 550 C, of 10 MPa or 13.8 MPa pressure, for periods of 500 hours and 1000 hours. At the lower pressure, the Mn-Mo-Ni steels, which contained predominantly M3C carbides, showed macroscopic blisters and extensive grain boundary cavities in the as-received and 0.7Mn modified conditions, but revealed very little damage when modified with 1% chromium. No hydrogen attack damage was observed in the 2.25Cr-1Mo modifications at 10 MPa. At higher pressure, the Mn-Mo-Ni steels showed extensive hydrogen attack damage. The 2.25Cr-1Mo steels are considerably less damaged by hydrogen attack than are the Mn-Mo-Ni steel and its modifications. Mechanical properties for a series of alloy modifications are presented and related to the stability of the microstructure and the heat treatment conditions

  15. Transmitted ultrasound pressure variation in micro blood vessel phantoms.

    Science.gov (United States)

    Qin, Shengping; Kruse, Dustin E; Ferrara, Katherine W

    2008-06-01

    Silica, cellulose and polymethylmethacrylate tubes with inner diameters of ten to a few hundred microns are commonly used as blood vessel phantoms in in vitro studies of microbubble or nanodroplet behavior during insonation. However, a detailed investigation of the ultrasonic fields within these micro-tubes has not yet been performed. This work provides a theoretical analysis of the ultrasonic fields within micro-tubes. Numerical results show that for the same tube material, the interaction between the micro-tube and megaHertz-frequency ultrasound may vary drastically with incident frequency, tube diameter and wall thickness. For 10 MHz ultrasonic insonation of a polymethylmethacrylate (PMMA) tube with an inner diameter of 195 microm and an outer diameter of 260 microm, the peak pressure within the tube can be up to 300% of incident pressure amplitude. However, using 1 MHz ultrasound and a silica tube with an inner diameter of 12 microm and an outer diameter of 50 microm, the peak pressure within the tube is only 12% of the incident pressure amplitude and correspondingly, the spatial-average-time-average intensity within the tube is only 1% of the incident intensity. PMID:18395962

  16. Development and operational experiences of an automated remote inspection system for interior of primary containment vessel of a BWR

    International Nuclear Information System (INIS)

    A prototype was developed for an automated remote inspection system featuring continuous monitoring of the working status of major components inside the primary containment vessel of a boiling water reactor. This inspection system consists of four units, or vehicles, which are towed by a trolley chain along a monorail; a complex coaxial cable for data transmission and for power supply; and an operator's console. A TV camera, microphone, thermometer, hygrometer, and ionization chamber are mounted on the various units. After several months' testing under high-ambient temperature, the system was installed in the Tokai-2 power station of Japan Atomic Power Company for in situ tests

  17. Local heat transfer from the corium melt pool to the boiling water reactor pressure vessel wall

    International Nuclear Information System (INIS)

    The present study considers in-vessel accident progression after core melt relocation to the lower head of a Boiling Water Reactor (BWR) and formation of a melt pool containing a forest of Control Rod Guide Tubes (CRGTs) cooled by purging flows. Descending streams of melt that flow along cooled surfaces of CRGT, and impinge on the bottom surface of the vessel wall can significantly increase local heat transfer. The area of enhanced heat transfer enlarges with decreasing of the melt Prandtl (Pr) number, while the peaking value of the heat transfer coefficient is a non-monotone function of Pr number. The melt Pr number depends on the melt composition (fractions of metallic and oxidic melt components) and thus is inherently uncertain parameter of the core melting and relocation scenarios. The effect of Pr number in the range of 1.02 - 0.03 on the local and integral thermal loads on the vessel wall is examined using Computational Fluid Dynamics (CFD). Heat transfer models obtained on the base of CFD simulations are implemented in the Phase-change Effective Convectivity Model (PECM) for simulation of reactor-scale accident progression heat transfer in real 3D geometry of the BWR lower plenum. We found that the influence of the low Pr number on the thermal loads in a big melt pool becomes more significant at later time, than rapid acceleration of the creep in the vessel wall. This result suggests that global vessel failure is insensitive to the melt composition in the considered 0.7 m deep melt pool configuration. However, it is not clear yet if the low Pr number effect has an influence on vessel failure mode in the other possible melt pool configurations. (author)

  18. A Reinforcement Plate for Partially Thinned Pressure Vessel Designed to Measure the Thickness of Vessel Wall Applying Ultrasonic Technique

    International Nuclear Information System (INIS)

    It is very hard to preserve the wall thickness of the vessel because of the erosion or corrosion as time goes by. Therefore, the wall thicknesses of heaters in power plants are periodically measured using ultrasonic test. If the integrity of the wall thickness is estimated not to secure, the reinforcement plate is welled on the thinned area of the vessel. The overlay weld of the reinforcement plate on the thinned vessel is normally the fillet welding. As shown by the references, the reinforcement plate with adequate thickness does its role very well before the vessel wall is perforated due to thinning. However, the integrity of shell cannot insure because the weldment is directly applied by the shell side pressure to after the vessel wall is perforated. Therefore, it is needed to measure the thickness of thinned area under the reinforcement plate continuously for preserving integrity and planning the fabrication of replacement vessel. It is impossible to apply the ultrasonic thickness measurement technique after the reinforcement plate is welded on the shell. In this paper new reinforcement plate, which makes it possible to measure the wall thickness under the reinforcement plate applying the ultrasonic technique, is introduced. A method to evaluate the structural integrity of a fillet weldment for the reinforcement plate welded on a pressure vessel is introduced in this paper. Moreover, new reinforcement plate, which makes it possible to measure the wall thickness of pressure vessels under the reinforcement plate applying the ultrasonic technique, is introduced

  19. Emergency cooling system for the core of a reactor pressure vessel

    International Nuclear Information System (INIS)

    In order to improve the spray distribution in an emergency cooling system for a BWR, the spray nozzles are situated vertically in bores of the pressure containment lid domed towards the inside. The distribution system is therefore situated above the lid and is supported on it. The penetrations for the incoming pipes are situated in the lid. This emergency cooling system is easy to mount and can be backfitted in existing plant. (orig./HP)

  20. Hydrostatic and cyclic pressure tests of a heavy walled pressure vessel

    International Nuclear Information System (INIS)

    A test pressure vessel for high temperature and high pressure use was designed and manufactured, and a series of experimental studies were carried out. It was assumed that the vessel was a reactor for a heavy oil-desulphurizing device, and it was designed in accordance with the ASME Section 8, Div. 2. with the design pressure of 262 kg/cm2, temperature of 4550C, and ASTM-A387-Gr.22, Class 2 as the material. The design and the manufacture of the test vessel are described in the first part of this report. The second part explains the test devices. The third part describes the purpose, the method, and the results of hydrostatic test at the pressure of 400 kg/cm2. The results of measurement were compared with the stress analysis by the finite element method. The fourth part describes about the fatigue test by applying the cyclic pressure ranging from 0 to 300 kg/cm2. The test is not completed yet, but the results of some preliminary experiments are presented in this report. (Aoki, K.)

  1. Pressurized thermal shock probabilistic fracture mechanics sensitivity analysis for Yankee Rowe reactor pressure vessel

    International Nuclear Information System (INIS)

    The Nuclear Regulatory Commission (NRC) requested Oak Ridge National Laboratory (ORNL) to perform a pressurized-thermal-shock (PTS) probabilistic fracture mechanics (PFM) sensitivity analysis for the Yankee Rowe reactor pressure vessel, for the fluences corresponding to the end of operating cycle 22, using a specific small-break-loss- of-coolant transient as the loading condition. Regions of the vessel with distinguishing features were to be treated individually -- upper axial weld, lower axial weld, circumferential weld, upper plate spot welds, upper plate regions between the spot welds, lower plate spot welds, and the lower plate regions between the spot welds. The fracture analysis methods used in the analysis of through-clad surface flaws were those contained in the established OCA-P computer code, which was developed during the Integrated Pressurized Thermal Shock (IPTS) Program. The NRC request specified that the OCA-P code be enhanced for this study to also calculate the conditional probabilities of failure for subclad flaws and embedded flaws. The results of this sensitivity analysis provide the NRC with (1) data that could be used to assess the relative influence of a number of key input parameters in the Yankee Rowe PTS analysis and (2) data that can be used for readily determining the probability of vessel failure once a more accurate indication of vessel embrittlement becomes available. This report is designated as HSST report No. 117

  2. A Study On In-Vessel Severe Accident Progression In The VVER-1000 Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    In a hypothetic severe accident, there is likelihood that a melt pool is formed and the Reactor Pressure Vessel (RPV) fails due to thermal creep. The present paper is concerned with the in-vessel accident progression in the VVER-1000 RPV. The study includes review of simulation results (RELAP, MELCOR), analysis of accident scenarios, determination of core materials relocation, and simulation of heat transfer of the melt pool formed in the lower plenum. The RELAP, MELCOR codes used for safety analysis are capable of simulation of severe accident progression and identification of accident scenarios in the reactor. However, there is limitation in describing turbulent natural convection heat transfer of a melt pool formed in the lower plenum. Computational Fluid Dynamics (CFD) codes which although are capable of melt pool heat transfer simulation, however, are too expensive. The Effective Convectivity Model (ECM) and Phase-change ECM (PECM) which were developed for melt pool heat transfer simulation are applied. The melt pool configuration, initial conditions are determined based on the analysis of accident scenarios, the ECM/PECM is used to simulate melt pool heat transfer. Results of ECM/PECM simulation are analyzed, compared with available RELAP, MELCOR accident progression data, RPV failure mode and timing are discussed. (author)

  3. Recent development for inservice inspection of reactor pressure vessels

    International Nuclear Information System (INIS)

    The German Nuclear Code (KTA-rules) requires a full scope inservice inspection (ISI) of reactor pressure vessels within a period of four years. This has a remarkable influence on plant operation and economy. Therefore, the development of advanced inspection equipment and techniques is directed not only to the enhancement of defect detectability and flaw sizing capabilities but also to reducing inspection times. A new manipulator system for PWR vessels together with fast data processing reduces the time for ISI of modern RPVs to 7 days. A new multichannel UT-system based on ALOK principle offers increased ultrasonic information with comfortable and rapid evaluation and presentation of results together with enhanced sizing capabilities. For specific inspection problems characterized by geometrical complexity the application of phased array probes in connection with UT-tomography provides improved ultrasonic information together with a streamlined manipulator principle and simplification of set up and tear down at the component which results in considerable reduction of radiation exposure. (orig.)

  4. Three-Dimensional Digital Image Correlation of a Composite Overwrapped Pressure Vessel During Hydrostatic Pressure Tests

    Science.gov (United States)

    Revilock, Duane M., Jr.; Thesken, John C.; Schmidt, Timothy E.

    2007-01-01

    Ambient temperature hydrostatic pressurization tests were conducted on a composite overwrapped pressure vessel (COPV) to understand the fiber stresses in COPV components. Two three-dimensional digital image correlation systems with high speed cameras were used in the evaluation to provide full field displacement and strain data for each pressurization test. A few of the key findings will be discussed including how the principal strains provided better insight into system behavior than traditional gauges, a high localized strain that was measured where gages were not present and the challenges of measuring curved surfaces with the use of a 1.25 in. thick layered polycarbonate panel that protected the cameras.

  5. Continuous Cooling Transformations in Nuclear Pressure Vessel Steels

    Science.gov (United States)

    Pous-Romero, Hector; Bhadeshia, Harry K. D. H.

    2014-10-01

    A class of low-alloy steels often referred to as SA508 represent key materials for the manufacture of nuclear reactor pressure vessels. The alloys have good properties, but the scatter in properties is of prime interest in safe design. Such scatter can arise from microstructural variations but most studies conclude that large components made from such steels are, following heat treatment, fully bainitic. In the present work, we demonstrate with the help of a variety of experimental techniques that the microstructures of three SA508 Gr.3 alloys are far from homogeneous when considered in the context of the cooling rates encountered in practice. In particular, allotriomorphic ferrite that is expected to lead to a deterioration in toughness, is found in the microstructure for realistic combinations of austenite grain size and the cooling rate combination. Parameters are established to identify the domains in which SA508 Gr.3 steels transform only into the fine bainitic microstructures.

  6. Neutron exposure for safety limitation for reactor pressure vessel steels

    International Nuclear Information System (INIS)

    The need of improvement in the neutron exposure limitation for reactor pressure vessel steels in the Federal Republic of Germany in metrological view is described. Its consequences from the standpoint of safety technology and of economics are pointed out. Trends in development and international recommendations are reported which are expected to abolish the mentioned deficiencies in the majority. Summarizing three tasks are formulated with respect to the intended improvement: a detailed recommendation for the procedure in radiation damage dosimetry, a test of usefulness by a correlation study for the displacementnumber per atom as the proposed measuring quantity, and compilation of neutron flux density spectra and damage indices for steel irradiation positions, especially also in nuclear power reactors. (orig.)

  7. Effect of aging on properties of pressure vessel steels

    International Nuclear Information System (INIS)

    Manganese-molybdenum-nickel steels are used in nuclear pressure vessels operating at temperatures up to 3500C. The effects of thermal ageing in the temperature range 300-5500C for durations up to 2 x 104 h have been studied in conventionally quenched and tempered and simulated heat-affected-zone (HAZ) microstructural conditions. Quantitative fractography and Auger spectroscopy have been used to relate changes in mechanical properties with changes in fracture mode and grain boundary chemistry. Aging increases the ductile-brittle transition temperature by an amount dependent on material, prior heat treatment, aging temperature and time. Embrittlement is associated with segregation of phosphorus to grain boundaries and is modelled using McLean's approach to equilibrium segregation

  8. 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.

  9. Reidual stresses in weld-clad reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Cladding of low alloy nuclear reactor pressure vessel steel with austenitic stainless steel introduces in heavy section components high residual stresses which may cause microrack formation in stress relief heat treatment. In this investigation an attempt is made to contribute to the solution of the stress relief cracking problem by determining quantitatively the magnitude and distribution of the residual stresses after cladding and after subsequent stress relief heat treatment. Lower residual stresses are caused by reduced thickness of the components. As the heat input is decreased at identical base material thickness, the residual stresses are lowered also. The height of the tensile residual stress peak, however, remains approximately constant. In stress relief annealed condition the residual stresses in the cladding are in tension; in the base material the residual stresses are negligibly small. (Auth.)

  10. 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

  11. ASTM Standards for Reactor Dosimetry and Pressure Vessel Surveillance

    International Nuclear Information System (INIS)

    The ASTM standards provide guidance and instruction on how to field and interpret reactor dosimetry. They provide a roadmap towards understanding the current ''state-of-the-art'' in reactor dosimetry, as reflected by the technical community. The consensus basis to the ASTM standards assures the user of an unbiased presentation of technical procedures and interpretations of the measurements. Some insight into the types of standards and the way in which they are organized can assist one in using them in an expeditious manner. Two example are presented to help orient new users to the breadth and interrelationship between the ASTM nuclear metrology standards. One example involves the testing of a new ''widget'' to verify the radiation hardness. The second example involves quantifying the radiation damage at a pressure vessel critical weld location through surveillance dosimetry and calculation

  12. Fracture probability evaluation of a LWR pressure vessel

    International Nuclear Information System (INIS)

    Fracture probability evaluation, of a LWR pressure vessel have been performed in the past, using statistical data from conventional plant. A more accurate evaluation has been requested in 1976 from the SCSIN to the CEA. With this object, a joint collaboration agreement has been signed between CEA, EURATOM/ISPRA and FRAMATOME. The whole program proceeding from this agreement is managed by a joint board including the three partners. The basic objective of this program is to develop a method which integrates, or makes it possible to integrate at a later stage, the greatest number of significant parameters. Also, in order to prepare the practical applications, a special effort is being made to collect the data corresponding to these parameters. Parallel basic research program have been launched in order to clarify our knowledge on some important parts of the main factors contributing to the evaluation. The results of this research will be progressively introduced into the method or will help checking its validity

  13. Computer system for International Reactor Pressure Vessel Materials Database support

    International Nuclear Information System (INIS)

    This report presents description of the computer tools for support of International Reactor Pressure Vessel Materials Database developed at IAEA. Work was focused on raw, qualified, processed materials data, search, retrieval, analysis, presentation and export possibilities of data. Developed software has the following main functions: provides software tools for querying and search of any type of data in the database; provides the capability to update the existing information in the database; provides the capability to present and print selected data; provides the possibility of export on yearly basis the run-time IRPVMDB with raw, qualified and processed materials data to Database members; provides the capability to export any selected sets of raw, qualified, processed materials data

  14. Nonlinear analysis of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

    Numerical procedures for predicting the nonlinear behaviour of a prestressed concrete reactor vessel over its design life are discussed. The numerical models are constructed by combining three-dimensional isoparametric finite elements which simulate the concrete, thin shell elements which simulate steel liner plates and layers of reinforcement steel, and axial elements for discrete prestressing cables. Nonlinearity under compressive stress, multi-dimensional cracking, shrinkage, and stress/temperature induced creep of concrete are considered in addition to the elastic plastic behaviour of the liner and reinforcing steel. The analysis of an actual PCRV is described. Stress contours and cracking patterns in the region of cutouts corresponding to operational pressure and temperature loads are illustrated. The effects of creep, unloading,and creep recovery are then shown. Lastly, a strategy for assessing the performance over its design life is discussed. (Auth.)

  15. 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.

  16. 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

    International Nuclear Information System (INIS)

    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)

  17. The behavior of shallow flaws in reactor pressure vessels

    International Nuclear Information System (INIS)

    Both analytical and experimental studies have shown that the effect of crack length, a, on the elastic-plastic toughness of structural steels is significant. The objective of this report is to recommend those research investigations that are necessary to understand the phenomenon of shallow behavior as it affects fracture toughness so that the results can be used properly in the structural margin assessment of reactor pressure vessels (RPVs) with flaws. Preliminary test results of A 533 B steel show an elevated crack-tip-opening displacement (CTOD) toughness similar to that observed for structural steels tested at the University of Kansas. Thus, the inherent resistance to fracture initiation of A 533 B steel with shallow flaws appears to be higher than that used in the current American Society of Mechanical Engineers (ASME) design curves based on testing fracture mechanics specimens with deep flaws. If this higher toughness of laboratory specimens with shallow flaws can be transferred to a higher resistance to failure in RPV design or analysis, then the actual margin of safety in nuclear vessels with shallow flaws would be greater than is currently assumed on the basis of deep-flaw test results. This elevation in toughness and greater resistance to fracture would be a very desirable situation, particularly for the pressurized-thermal shock (PTS) analysis in which shallow flaws are assumed to exist. Before any advantage can be taken of this possible increase in initiation toughness, numerous factors must be analyzed to ensure the transferability of the data. This report reviews those factors and makes recommendations of studies that are needed to assess the transferability of shallow-flaw toughness test results to the structural margin assessment of RPV with shallow flaws. 14 refs., 8 figs

  18. Evaluation of the reactor pressure vessel steels by positron annihilation

    International Nuclear Information System (INIS)

    This paper presents a comparison of commercially used German and Russian reactor pressure vessel steels from the positron annihilation spectroscopy (PAS) point of view, having in mind knowledge obtained also from other techniques from the last decades. The second generation of Russian RPV steels seems to be fully comparable with German steels and their quality allows prolongation of NPP operating lifetime over projected 40 years. The embrittlement of CrMoV steels is relatively low due to effect of higher temperature which implies partial in situ annealing of primary microstructural point defects and therefore delays the degradation processes caused by neutron irradiation. PAS techniques can be effectively applied for evaluation of microstructural changes caused by extreme external loads (characterized by high dpa values) by proton implantation, with aim to simulate irradiation and for the evaluation of the effectiveness of post-irradiation thermal treatments. We used our actual and previous results, collected during last 20 years from measurements of different RPV-steels in “as received”, irradiated and post-irradiation annealed state and compare them with the aim to contribute to general knowledge based on experimental PAS data. Actual results from irradiated German and Russian steels confirmed that no large voids or vacancy clusters were formed at defined irradiation conditions stated according to the real operational conditions at nuclear power plants. This indicate the fact that vacancy type defects bear hardly any responsibility for radiation-induced hardening and embrittlement of reactor pressure vessel steels and does not affect significantly the long-term operation of nuclear power plants from safety point of view

  19. Neutron irradiation embrittlement of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    The reliability of nuclear power plants depends on the proper functioning of complex components over the whole life on the plant. Particular concern for reliability is directed to the primary pressure boundary. This report focuses on the portion of the primary system exposed to and significantly affected by neutron radiation. Experimental evidence from research programmes and from reactor surveillance programmes has indicated radiation embrittlement of a magnitude sufficient to raise doubts about reactor pressure vessel integrity. The crucial nature of the primary vessel function heightens the need to be alert to this problem, to which, fortunately, there are positive aspects: for example, steels have been developed which are relatively immune to radiation embrittlement. Further, awareness of such embrittlement has led to designs which can accomodate this factor. The nature of nuclear reactors, of the steels used in their construction, and of the procedures for interpreting embrittlement and minimizing the effects are reviewed with reference to the reactors that are expected to play a major role in electric power production from now to about the turn of the century. The report is intended as a manual or guidebook; the aim has been to make each chapter or major sub-division sufficiently comprehensive and self-contained for it to be understood and read independently of the rest of the book. At the same time, it is hoped that the whole is unified enough to make a complete reading useful and interesting to the several classes of reader that are involved with only specific aspects of the topic

  20. Evaluation of the reactor pressure vessel steels by positron annihilation

    Energy Technology Data Exchange (ETDEWEB)

    Slugeň, V., E-mail: Vladimir.Slugen@elf.stuba.sk [Institute of Nuclear and Physical Engineering, Slovak University of Technology, Ilkovičova 3, 81219 Bratislava (Slovakia); Hein, H. [AREVA NP GmbH, Paul Gossen Strasse 100, 91 001 Erlangen (Germany); Sojak, S.; Simeg Veterníková, J.; Petriska, M.; Sabelová, V.; Pavúk, M.; Hinca, R.; Stacho, M. [Institute of Nuclear and Physical Engineering, Slovak University of Technology, Ilkovičova 3, 81219 Bratislava (Slovakia)

    2013-11-15

    This paper presents a comparison of commercially used German and Russian reactor pressure vessel steels from the positron annihilation spectroscopy (PAS) point of view, having in mind knowledge obtained also from other techniques from the last decades. The second generation of Russian RPV steels seems to be fully comparable with German steels and their quality allows prolongation of NPP operating lifetime over projected 40 years. The embrittlement of CrMoV steels is relatively low due to effect of higher temperature which implies partial in situ annealing of primary microstructural point defects and therefore delays the degradation processes caused by neutron irradiation. PAS techniques can be effectively applied for evaluation of microstructural changes caused by extreme external loads (characterized by high dpa values) by proton implantation, with aim to simulate irradiation and for the evaluation of the effectiveness of post-irradiation thermal treatments. We used our actual and previous results, collected during last 20 years from measurements of different RPV-steels in “as received”, irradiated and post-irradiation annealed state and compare them with the aim to contribute to general knowledge based on experimental PAS data. Actual results from irradiated German and Russian steels confirmed that no large voids or vacancy clusters were formed at defined irradiation conditions stated according to the real operational conditions at nuclear power plants. This indicate the fact that vacancy type defects bear hardly any responsibility for radiation-induced hardening and embrittlement of reactor pressure vessel steels and does not affect significantly the long-term operation of nuclear power plants from safety point of view.

  1. 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

  2. WWER-1000 core loading characteristic influence on irradiation conditions of surveillance specimens and reactor pressure vessel

    International Nuclear Information System (INIS)

    Irradiation conditions of WWER-1000 surveillance specimens and reactor pressure vessel are comparative analyzed for various core loadings. It is proved that the fluences onto specimens don't correlate with ones onto pressure vessel. It is shown that the reconstruction technique using to surveillance specimens of the standard program implemented at the most of the power units with WWER-1000 allows obtaining reliable information on the reactor pressure vessel metal states

  3. FAVOR: A new fracture mechanics code for reactor pressure vessels subjected to pressurized thermal shock

    International Nuclear Information System (INIS)

    Probabilistic fracture mechanics (PFM) analysis is a major element of the comprehensive probabilistic methodology endorsed by the Nuclear Regulatory Commission (NRC) for evaluation of the integrity of pressurized water reactor pressure vessels subjected to pressurized-thermal-shock (PTS) transients. OCA-P and VISA-II are PTS PFM computer codes that are currently referenced in Regulatory Guide 1.154 as acceptable codes for performing plant-specific analyses. These codes perform PFM analyses to estimate the increase in vessel failure probability as the vessel accumulates radiation damage over the operating life of the vessel. Experience with the application of these codes in the last few years has provided insights into areas where they could be improved. As more plants approach the PTS screening criteria and are required to perform plant-specific analyses, there will be an increasing need for an improved and validated PTS PFM code that is accepted by the NRC and utilities. The NRC funded Heavy Section Steel Technology Program (HSST) at the Oak Ridge National Laboratory is currently developing the FAVOR (Fracture Analysis of Vessels: Oak Ridge) code, which is expected to meet this need. The FAVOR code incorporates the most important features of both OCA-P and VISA-II and contains some new capabilities such as (1) a PFM global modeling methodology; (2) the calculation of the axial stress component associated with coolant streaming beneath an inlet nozzle; (3) a library of stress intensity factor influence coefficients, generated by the NQA-1 certified ABAQUS computer code, for an appropriate range of two and three dimensional inner-surface flaws; (4) the flexibility to generate a variety of output reports; and (5) enhanced user friendliness

  4. Acoustic emission method for tracing crack propagation in pressure vessels

    International Nuclear Information System (INIS)

    A hemispherical model and a pressure vessel model were used as samples. In the former model, artificial notches were fabricated at the top. In the latter, two types of artificial notches were fabricated at the points of origin of the maximum stress (corners of the inner surface of the nozzles) in four nozzles with different shapes which were mounted on the model proper. The AE method was used to investigate the process of crack initiation and propagation from these artificial notches by means of repeated loading with internal pressure. It was possible to obtain from the results of these tests much useful data concerning the properties of AE and the points of origin of AE (positional tracking) when cracks are initiated and propagated in structures having complex shapes such as these samples. Simultaneously with the measurements by the AE method, Smek gages and crack gages mounted on the nozzle corners were used to investigate the crack initiation and propagation behavior. It was established that there is a close connection between them

  5. Scaling considerations in a pressure vessel upper head SBLOCA

    International Nuclear Information System (INIS)

    The knowledge of thermalhydraulic phenomena occurring in a Nuclear Power Plant (NPP) during an accident is very important in the assessment of nuclear safety. Due to full-scale testing is usually impossible to perform, small scale Integral Test Facilities (ITFs) are necessary. This paper is focused on the investigation if physical phenomena observed in a small test facility, such as Large Scale Test Facility (LSTF), during a Pressure Vessel (PV) upper head Small Break Loss-Of-Coolant Accident (SBLOCA) can be reliably extrapolated for a scaled-up NPP model. With this aim, the present work details the scaling method used to obtain a scaled-up TRACE5 model from a LSTF TRACE5 model developed and tested by authors in previous works. The scaled-up NPP TRACE5 model has been obtained applying the power-to-volume scaling criterion. The transient considered corresponds to the Test 6-1 of OECD/NEA ROSA Project. This test reproduces a PV upper head SBLOCA with a break size equivalent to 1.9% cold leg break, under the assumption of total failure of High Pressure Injection (HPI) system. Results show that the main physical phenomena is well reproduced in both models, although slight discrepancies are observed due to the bad reproduction of the accumulators injection. (author)

  6. Nonlinear analysis of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

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

  7. Development of a crack monitoring technique for use in a corrosion fatigue study of SA533-B pressure vessel steel

    International Nuclear Information System (INIS)

    At present there does not exist a realistic crack growth law which will provide a good description of the relationship between the alternating stress intensity factor and the crack growth per cycle of stress. Such a law should be applicable to either the pressurized water reactor environment (PWR) or boiling water reactor environmnt (BWR). This project was formulated with the aim of examining the fatigue crack growth rate of SA533-B steel (a nuclear pressure vessel steel) in the threshold region in a simulated PWR environment. The aim of this report is to develop a crack monitoring technique for use in corrosion fatigue studies. Factors affecting fatigue crack propagation include: frequency, stress range, the effect of irradiation, ageing and environment. The mechanisms of crack propagation that are discussed include: slip dissolution, hydrogen assisted cracking, corrosion potential, and morphology studies. D.C. electrical potential, the compliance technique and the back-faced strain gauge method can be used for crack monitoring. Details are also given on the experimental equipment and programme. The results of the experiment has shown that the potential difference technique for monitoring crack length is a valuable one and is well suited for use in fatigue testing applications

  8. Behavior of a corium jet in high pressure melt ejection from a reactor pressure vessel

    International Nuclear Information System (INIS)

    This report provides results from analytical and experimental investigations on the behavior of a gas supersaturated molten jet expelled from a pressurized vessel. Aero-hydrodynamic stability of liquid jets in gas, stream degassing of molten metals and gas bubble nucleation in molten metals are relevant problems which are addressed in this work. Models are developed for jet expansion, primary breakup of the jet and secondary fragmentation of melt droplets resulting from violent effervescence of dissolved gas. The jet expansion model is based on a general relation for bubble growth which includes both inertia-controlled and diffusion-controlled growth phases. The jet expansion model is able to predict the jet void fraction, jet radius as a function of axial distance from the pressure vessel, bubble size and bubble pressure. The number density of gas bubbles in the melt, which is a basic parameter in the model, was determined experimentally and is about 108 per m3 of liquid. The primary breakup of the jet produces a spray of droplets, about 2-3 mm in diameter. Parametric calculations for a TMLB' reactor accident sequence show that the corium jet is disrupted within a few initial jet diameters from the reactor vessel and that the radius of corium spray at the level of the reactor cavity floor is in the range of 0.8 to 2.6 m. (orig./HP)

  9. Structural integrity assessment of reactor pressure vessels during pressurized thermal shock

    International Nuclear Information System (INIS)

    A comparative assessment study is performed for the deterministic fracture mechanics approach of the pressurized thermal shock of a reactor pressure vessel. Round robin problems consisting of two transients and two defects are solved. Their results are compared to suggest some recommendations of best practices and to assure an understanding of the key parameters of this type of approach, which will be helpful not only for the benchmark calculations and results comparisons but also as a part of the knowledge management for the future generation. Seven participants from five organizations solved the problem and their results are compiled in this study

  10. 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

  11. Preservation and management of knowledge on WWER reactor pressure vessels

    International Nuclear Information System (INIS)

    Preservation and management of knowledge are becoming a rising challenge and the importance of collecting all available information about the behaviour of different types of reactors, including WWER, is increasingly being recognized. Several experts are close to retirement, and preservation and use of their knowledge and experience will be nearly impossible in a few years. A new project is in progress at the Joint Research Centre of the European Commission (JRC-EC) Institute for Energy in cooperation with the Nuclear Research Institute Rez with the intention to collect all available information about reactor pressure vessels of WWER type reactors to analyse and summarize the most important items and issues. This project will contribute significantly to the new project Coordinated Action on VVER Safety (COVERS) on WWER Safety under the Framework Programme 6 of the European Commission (EC-FP6), in which all WWER operating countries are also taking part. The results of the project will be useful to young specialists of the new generation in all countries, since they would have access not only to current views and knowledge but also to the history and background on all aspects of WWER power plants. From the very beginning of researched studies and initial experiences with the operation of reactors, a large number of publications were issued in Russian as well as in other national languages (Czech, Slovak, Hungarian, Finnish, etc.). Today, there is no easy access to this substantial amount of information and experience for most foreign experts and WWER reactor operators in different countries. Many publications were also prepared in languages other than Russian for presentation at national conferences or workshops and appearing in national and international journals. Most of this information, experience and knowledge can not be retrieved easily. In the new project, more than 20 specialists, mainly from WWER operating countries, would help with the collection of such

  12. Superheated steam annealing of pressurized water reactor vessel

    International Nuclear Information System (INIS)

    Thermal annealing of an embrittled Reactor Pressure Shell is the only recognized means for recovering material properties lost due to long-term exposure of the reactor walls to radiation. Reduced toughness of the material during operation is a major concern in evaluations of structural integrity of older reactors. Extensive studies performed within programs related to life extension of nuclear plants have confirmed that the thermal treatment of 850 deg. F for 168 hours on irradiated material essentially recovers material properties lost due to neutron exposure. Dry and wet annealing methods have been considered. Wet annealing involves operating the reactor at near design temperatures and pressures. Since the temperature of wet annealing must be limited to vessel design temperature of 650 deg. F, only partial recovery of the lost properties is achieved. Thus dry annealing was selected as an alternative for future development and industrial implementation to extend the safe life of reactors. Dry thermal annealing consists of heating portions of the reactor vessel at a specific temperature for a given period of time using a high temperature heat source. The use of spent fuel assemblies, induction heating and resistance heating elements as well as the circulation of heated fluid were investigated as potential candidate methods. To date the use of resistance heating elements which are lowered into a dry empty reactor was considered to be the preferred method. In-depth research in the United States and practical applications of such a method in Russia have confirmed feasibility of the method. The method of using circulating superheated steam to anneal the vessel at 850 deg. F without complete removal of the reactor internals is described herein. After removing the reactor head and fuel, the core barrel along with the upper and lower core in PWRs is lifted to open an annular space between the reactor shell flange and the core barrel flange. The thermal shield can remain

  13. Structure mechanical analysis of prestressed cast-steel pressure vessels with the finite-element-method

    International Nuclear Information System (INIS)

    The analytical pressure analysis is performed for a vessel with solid bottom and top. The basis of the Finite-Element-Method (FEM) and the criteria for the choice of a suitable element type for use in the computer model was investigated. To investigate the exactness of the FE-program a comparison between the analytical solution and the pressure claculated by FEM at a cylindrical vessel was made. For pressure analyses at the test vessel built of steel sections four different computer models (after FEM) were developed. The pressure analysis of a prestressed cast-steel pressure vessel for the transport and for the storage of burnt HTR fuel elements is performed with the aid of computed models after FEM. The method of developing simple computer models for the prestressed pressure vessel with large dimension is explained with an example. (orig.)

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

    International Nuclear Information System (INIS)

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

  15. The conceptual design of IPR1000 reactor pressure vessel for PWR type

    International Nuclear Information System (INIS)

    The conceptual design of IPR1000 reactor pressure vessel for PWR type has been configured, material selection and dimension parameter are designed to cover the reactor cooling system (RCS), nuclear fuel assembly, and others internals reactor. The reactor pressure vessel consist of closure head assembly, vessel shell upper head assembly, vessel shell lower assembly, and inlet and outlet nozzle. These are designed capable to support weight of RPV, at pressures and temperature of each 2485 psig and 650 °F. The design refers to AP1000 as according to ASME code and industrials standard applicable for Nuclear Power Plants (NPP). (author)

  16. Fabrication techniques of metal liner used for pressure vessels made by composite material

    International Nuclear Information System (INIS)

    Different viable techniques for the manufacturing of metal liner used for pressure vessels are presented. The aim of these metal liner is to avoid the fluid leakage from the pressurized vessel and to serve as a mandreal to be wound by composite material. The studied techniques are described and the practical results are illustrated. Finally a comparative study of the manufacturing techniques is made in order to define the process that furnishes the metal liner with the best characteristics. The advantages offered by these type of pressure vessels when compared with the conventional metallic vessels, are also presented. (Author)

  17. BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

    International Nuclear Information System (INIS)

    This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

  18. Material and fabrication of the HTTR reactor pressure vessel

    International Nuclear Information System (INIS)

    The High Temperature Engineering Test Reactor (HTTR) is under construction at Oarai Research Establishment, Japan Atomic Energy Research Institute (JAERI) and planned to be critical in October 1997. Fabrication of the HTTR reactor pressure vessel (RPV) at Kure Works, Babcock-Hitachi K.K. took about two years, and the RPV was transported to the Oarai site in August 1994. Pressure test of the primary and secondary cooling system including the RPV was performed successfully in March 1996. Because temperature of the HTTR RPV becomes about 400 deg. C at normal operation, 2 1/4 Cr-1 Mo steel is chosen for it. Fluence of the RPV is calculated to be less than 1 X 1017 n/cm2 (E>l MeV), and so irradiation embrittlement, is presumed to be negligible, but temper embrittlement is not. For the purpose of reducing embrittlement, content of some elements is limited on 2 1/4 Cr-1 Mo steel for the HTTR RPV using embrittlement parameters: J-factor and X-bar. In this paper design and structure of the HTTR RPV is briefly reviewed first. Fabrication procedure of the RPV and its special feature is shown. Material data on 2 1/4 Cr-1 Mo steel manufactured for the RPV, especially the embrittlement parameters J-factor and X-bar, and nil-ductility transition temperatures TNDT by drop weight tests, are shown, and increase in the transition temperature is estimated based on data available in literature. Technology of the HTTR RPV is applicable to RPVs of future commercial High Temperature Gas-cooled Reactors (HTGRs). (author)

  19. Radial Body Forces Influence on FGM and Non-FGM Cylindrical Pressure Vessels

    OpenAIRE

    Jacob Nagler

    2016-01-01

    This study deals with the influence of radial body forces on FGM and non-FGM pressure vessels. It contains an extended overview of pressure vessels made from both kinds of material. Furthermore, full mathematical development of stress-strain field for both kinds of cylindrical vessels while being influenced by body forces has been performed. In addition, a new power law model for FGM materials was suggested and discussed. Finally, tables of composed plastic-elastic states are discussed.

  20. Probabilistic Structural Integrity Analysis of Boiling Water Reactor Pressure Vessel under Low Temperature Overpressure Event

    OpenAIRE

    Hsoung-Wei Chou; Chin-Cheng Huang

    2015-01-01

    The probabilistic structural integrity of a Taiwan domestic boiling water reactor pressure vessel has been evaluated by the probabilistic fracture mechanics analysis. First, the analysis model was built for the beltline region of the reactor pressure vessel considering the plant specific data. Meanwhile, the flaw models which comprehensively simulate all kinds of preexisting flaws along the vessel wall were employed here. The low temperature overpressure transient which has been concluded to ...

  1. Residual stresses in weld-clad reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Cladding of low alloy nuclear reactor pressure vessel steel with austenitic stainless steel introduces in heavy section components high residual stresses which may cause microcrack formation in stress relief heat treatment. In this investigation an attempt is made to contribute to the solution of the stress relief cracking problem by determining quantitatively the magnitude and distribution of the residual stresses after cladding and after subsequent stress relief heat treatment. The distribution of residual stresses was determined on the basis of a combined experimental-mathematical procedure. Heavy section plate specimens of low alloy steel as base material were given an austenitic monolayer-cladding using the techniques of strip electrode and plasma hot wire cladding, respectively. A number of plates was stress relief heat treated. Starting from the cladded surface the thickness of the plates was reduced by subsequent removal of layers of material. The elastic strain reaction to the removal of each layer was measured by strain gauges. From the data obtained the biaxial residual stress distribution was computed as a function of thickness using relations which are derived for this particular case. In summary, lower residual stresses are caused by reduced thickness of the components. As the heat input, is decreased at identical base material thickness, the residual stresses are lowered also. The height of the tensile residual stress peak, however, remains approximataly constant. In stress relief annealed condition the residual stresses in the cladding are in tension; in the base material the residual stresses are negligibly small

  2. Pressure vessel steel embrittlement monitoring by magnetic properties measurements

    International Nuclear Information System (INIS)

    The magnetic properties of specimens of one heat of A533B nuclear pressure vessel grade steel have been examined in the as-received condition and after neutron irradiation to various fluence levels up to 4 x 1018 cm-2 (E > 0.1 MeV) in the University of Illinois advanced TRIGA reactor core at two temperatures, approximately 120 and 260 C. The effect of certain heat treatments was also investigated. The magnetic properties were measured by an automated hysteresis curve-tracing method using a miniature transformer which incorporated the specimens in its core. Changes in magnetic hysteresis energy loss were correlated with neutron fluence in the case of certain irradiated specimens and with microhardness measurements in the case of heat-treated specimens. At the higher irradiation temperatures, no significant changes in either the magnetic hysteresis properties or the microhardness were noted for the present fluences. The relationship between the observed magnetic properties response and irradiation-induced embrittlement is discussed

  3. Radiation embrittlement of reactor pressure vessel - BMFT/USNRC cooperation

    International Nuclear Information System (INIS)

    On the basis of comparison materials irradiated in different reactors the integrated effect of all factors accounted for the embrittlement-recovery equilibrium including test and evaluation procedures is to be investigated. A test melt with low upper shelf energy (low shelf test melt LSTM) on the basis of 22 NiMoCr 3 7 (KS 07 A/B) with a chemical composition exceeding the specified limits was selected for this purpose. Together with specimens of material 22NiMoCr 3 7 and 20 MnMoNi 5 5 according to present specification for reactor pressure vessels of light water reactors, the irradiation behavior of the KS 07 material is being investigated in the United States of America in comparison to the research program 'Integrity of components (FKS)' conducted in Germany. This report presents the results from tensile, Charpy impact and fracture mechanics tests obtained within the HSST (heavy section steel technology program) program. First results from the FKS program are given for comparison. A complete evaluation can only be made within the final FKS report 'Irradiation' since at this time not all the necessary results are available. (orig./IHOE)

  4. Microscopic examination of crack growth in a pressure vessel steel

    International Nuclear Information System (INIS)

    A fairly systematic microscopic study concerning ductile and ductile-brittle crack growth in the A508B pressure vessel steel has been performed. The main method of investigation was to subject fracture mechanics specimens (sub-sized three point bend specimens) to predetermined load levels corresponding to different amounts of ductile crack extension. The specimens were then cut perpendicularly to the plane of the crack and the area in front of the crack was examined in a SEM. The object of these examinations was to determine if newly encountered computational results could be correlated to crack extension characteristics and to study whether the mechanism of ductile growth was of the void growth type or of the fast shear mechanism. This is important for further numerical modelling of the process. Both the original material and a specially heat treated piece were investigated. The heat treatment was performed in order to raise the transition temperature to about 60 deg C with the object to provide a more convenient testing situation. Charpy V tests were performed for the specially heat treated material to obtain the temperature dependence of the toughness. This was also studied by performing fracture toughness determination on the same type of specimens as were used for the microscopic study. The heat treatment used fulfilled the above purpose and the microscopic studies provide a good understanding of the micro mechanisms operating in the ductile fracture process for this material

  5. Underwater television camera for monitoring inner side of pressure vessel

    International Nuclear Information System (INIS)

    An underwater television support device equipped with a rotatable and vertically movable underwater television camera and an underwater television camera controlling device for monitoring images of the inside of the reactor core photographed by the underwater television camera to control the position of the underwater television camera and the underwater light are disposed on an upper lattice plate of a reactor pressure vessel. Both of them are electrically connected with each other by way of a cable to rapidly observe the inside of the reactor core by the underwater television camera. The reproducibility is extremely satisfactory by efficiently concentrating the position of the camera and image information upon inspection and observation. As a result, the steps for periodical inspection can be reduced to shorten the days for the periodical inspection. Since there is no requirement to withdraw fuel assemblies over a wide reactor core region, and the device can be used with the fuel assemblies being left as they are in the reactor, it is suitable for inspection of detectors for nuclear instrumentation. (N.H.)

  6. Updated embrittlement trend curve for reactor pressure vessel steels

    International Nuclear Information System (INIS)

    The reactor pressure vessels of commercial nuclear power plants are subject to embrittlement due to exposure to high energy neutrons from the core. Irradiation embrittlement of RPV belt-line materials is currently evaluated using US Regulatory Guide 1.99 Revision 2 (RG 1.99 Rev 2), which presents methods for estimating the Charpy transition temperature shift (ΔT30) at 30 ft-lb (41 J) and the drop in Charpy upper shelf energy (ΔUSE). A more recent embrittlement model, based on a broader database and more recent research results, is presented in NUREG/CR-6551. The objective of this paper is to describe the most recent update to the embrittlement model in NUREG/CR-6551, based upon additional data and increased understanding of embrittlement mechanisms. The updated ΔT30 and USE models include fluence, copper, nickel, phosphorous content, and product form; the ΔT30 model also includes coolant temperature, irradiation time (or flux), and a long-time term. The models were developed using multi-variable surface fitting techniques, understanding of the ΔT30 mechanisms, and engineering judgment. The updated ΔT30 model reduces scatter significantly relative to RG 1.99 Rev 2 on the currently available database for plates, forgings, and welds. This updated embrittlement trend curve will form the basis of revision 3 to Regulatory Guide 1.99. (author)

  7. Advances in crack-arrest technology for reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Bass, B.R.; Pugh, C.E.

    1988-01-01

    The Heavy-Section Steel Technology (HSST) Program at the Oak Ridge National Laboratory (ORNL) under the sponsorship of the US Nuclear Regulatory Commission is continuing to improve the understanding of conditions that govern the initiation, rapid propagation, arrest, and ductile tearing of cracks in reactor pressure vessel (RPV) steels. This paper describes recent advances in a coordinated effort being conducted under the HSST Program by ORNL and several subcontracting groups to develop the crack-arrest data base and the analytical tools required to construct inelastic dynamic fracture models for RPV steels. Large-scale tests are being carried out to generate crack-arrest toughness data at temperatures approaching and above the onset of Charpy upper-shelf behavior. Small- and intermediate-size specimens subjected to static and dynamic loading are being developed and tested to provide additional fracture data for RPV steels. Viscoplastic effects are being included in dynamic fracture models and computer programs and their utility validated through analyses of data from carefully controlled experiments. Recent studies are described that examine convergence problems associated with energy-based fracture parameters in viscoplastic-dynamic fracture applications. Alternative techniques that have potential for achieving convergent solutions for fracture parameters in the context of viscoplastic-dynamic models are discussed. 46 refs., 15 figs., 3 tabs.

  8. Towards safe long-term operation of reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Rouden, Jenny; Efsing, Pal [Ringhals AB, Vaeroebacka (Sweden); Hein, Hieronymus; May, Johannes [AREVA GmbH, Erlangen (Germany); Planman, Tapio [VTT (Finland); Todeschini, Patrick [EDF, Paris (France); Brumovsky, Milan [UJV Rez, a.s., Hlavni (Czech Republic); Ballesteros, Antonio [JRC Institute for Energy and Transport, Petten (Netherlands). Nuclear Reactor Safety Assessment Unit; Gillemot, Ferenc [MTA-EK, Budapest (Hungary); Chaouadi, Rachid [SCK-CEN, Mol (Belgium); Altstadt, Eberhard [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany)

    2015-05-15

    This publication summarizes the long term operation (LTO) conditions in European NPPs and provides recommendations on reactor pressure vessel (RPV) irradiation surveillance based on the work performed in the Work Package 7 ''Surveillance Guidelines'' of the LONGLIFE international project. The LONGLIFE project ''Treatment of Long Term Irradiation Embrittlement Effects in RPV Safety Assessment'' was 50 % funded by the Euratom 7th Framework Programme of the European Commission. Specific scientific and technical issues addressed in this publication are the following: - Surveillance standards and procedures. - Reuse of tested irradiated surveillance specimens. - Transferability of test reactor results to LWR conditions. - Extension of RPV irradiation surveillance programmes. - Withdrawal scheme for LTO surveillance programmes. The objective of the surveillance guidelines is to support the potential end-user (utilities, nuclear power plants, research institutes, etc.) in selecting the appropriate strategy and technical approaches for RPV irradiation surveillance for LTO. In this way contributing to a reliable monitoring of long-term irradiation effects in RPV, and supporting the European efforts towards harmonisation of procedures for RPV surveillance and safety assessment in the light of LTO.

  9. 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.

  10. Towards safe long-term operation of reactor pressure vessels

    International Nuclear Information System (INIS)

    This publication summarizes the long term operation (LTO) conditions in European NPPs and provides recommendations on reactor pressure vessel (RPV) irradiation surveillance based on the work performed in the Work Package 7 ''Surveillance Guidelines'' of the LONGLIFE international project. The LONGLIFE project ''Treatment of Long Term Irradiation Embrittlement Effects in RPV Safety Assessment'' was 50 % funded by the Euratom 7th Framework Programme of the European Commission. Specific scientific and technical issues addressed in this publication are the following: - Surveillance standards and procedures. - Reuse of tested irradiated surveillance specimens. - Transferability of test reactor results to LWR conditions. - Extension of RPV irradiation surveillance programmes. - Withdrawal scheme for LTO surveillance programmes. The objective of the surveillance guidelines is to support the potential end-user (utilities, nuclear power plants, research institutes, etc.) in selecting the appropriate strategy and technical approaches for RPV irradiation surveillance for LTO. In this way contributing to a reliable monitoring of long-term irradiation effects in RPV, and supporting the European efforts towards harmonisation of procedures for RPV surveillance and safety assessment in the light of LTO.

  11. Investigation on reconstitution of reactor pressure vessel surveillance specimen

    International Nuclear Information System (INIS)

    Since the reconstitution of surveillance specimens became an issue due to the shortage of surveillance test specimens for reactor pressure vessels (RPVs) for long-term operation of nuclear power plants, investigation research had been conducted for seven years until March 2006. As the standard Charpy and CT specimen reconstitution, minimum insert length was obtained from hardness distribution and twice of the sum of plastic zone width plus maximum of heat affected zone width and heat recovery zone width. Plastic zone width was correlated with absorbed energy (J) for Charpy impact test specimen (5.3 mm maximum) and J-integral (kJ/m2) for CT test specimen. Heat affected zone was checked by etching, and 1.2 mm for Charpy specimen of surface activated joining reconstitution and 1.6 mm for CT specimen of laser welding reconstitution. Heat recovery width was obtained by test measurement or thermal analysis of temperature history of inserts under the joining condition, and 1.9 mm for Charpy specimen of surface activated joining reconstitution and 2.5 mm for CT specimen of laser welding reconstitution. Standard surveillance specimen reconstitution could contribute to assessment of the integrity of aged RPVs. (T. Tanaka)

  12. HEXANN-EVALU, Neutron Irradiation of Reactor Pressure Vessels

    International Nuclear Information System (INIS)

    1 - Description of program or function: HEXANN-EVALU calculates the neutron irradiation of a pressure vessel surrounding a nuclear reactor core composed of hexagonal assemblies. The area outside the core may contain hexagonal shielding assemblies of non-multiplying materials, a core liner and annular material zones. 2 - Method of solution: The Monte Carlo method is used. The neutrons start at the core boundary with a given distribution in space, angle and energy. The angular distribution is calculated by HEXANN itself from data describing the spatial distribution of the neutron source in the core. Survival biasing is used in all collisions. To increase efficiency the following options are included: region-wise importance, Russian roulette, low energy Russian roulette, splitting, path stretching with explicit exponential transform and automatic importance correction. 3 - Restrictions on the complexity of the problem: 30-degree symmetry assumed; Maximum number of energy groups = 30; Maximum group number change in down-scattering = 19; No up-scattering; Maximum number of materials = 10; Maximum number of cylindrical surfaces = 20; Maximum number of source faces (hexagon faces at edge of core in 30-degree sector)= 20; Vacuum boundary conditions are used at the top, bottom and outer boundary

  13. Assessment of reactor pressure vessel multilayer weld metal

    International Nuclear Information System (INIS)

    Fracture toughness of the multi-layer belt-line welding seam of the Biblis C reactor pressure vessel was characterized by the test standard ASTM E1921. The reference temperature, T0, was determined for different thickness positions of the multi-layer welding seam. Additionally, the influence of the specimen orientation on T0 was investigated. In contrast to the T-S orientation (crack extension through the thickness) the crack front of the T-L oriented specimens (crack extension in welding direction) penetrates several welding beads. By means of fractographic and metallographic analyses it was shown that the distribution of the crack initiation sites is not necessarily correlated to the structure of the different welding beads along the crack front. Furthermore, it was found that the scatter of the KJc values determined with T-S specimens is significantly higher than in case of the T-L specimens. T0 values measured at different thickness locations of the multi-layer welding seam vary in a range of about 40 K. (authors)

  14. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Requirements for thermal annealing of the reactor pressure... Requirements for thermal annealing of the reactor pressure vessel. (a) For those light water nuclear power... thermal annealing may be applied to the reactor vessel to recover the fracture toughness of the...

  15. PWR 900 MWe pressure vessel surveillance - analysis of neutron field characteristics and damage function experimental determination

    International Nuclear Information System (INIS)

    This paper presents an overview of the studies performed by CEA and EDF in the scope of the pressure vessel surveillance of the nuclear installations in France. The power plants are equipped with specimens for monitoring the effects of the irradiation to the pressure vessel material. (TEC)

  16. Simulation of the aspersion system of the core low pressure (LPCS) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

    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)

  17. Influence of gamma irradiation on the deterioration of reactor pressure vessel materials and on reactor dosimetry measurements. Final report

    International Nuclear Information System (INIS)

    Radiation embrittlement of pressure vessel steel in mixed neutron-gamma fields is mostly determined by neutrons, but in some cases also by gamma-radiation. Depending on the reactor type, gamma radiation can influence evaluations of lead factors of surveillance specimens, effect the interpretation of results of irradiation experiments and finally, it can result in changed pressure vessel lifetime evaluations. The project aimed at the evaluation of the importance of gamma radiation for RPV steel damage for several types of light-water reactors. Absolute neutron and gamma fluence rate spectra had been calculated for the Russian PWR types VVER-440 and two core loading variants of VVER-1000, for a German 1300 MW PWR and a German 900 MW BWR. Based on the calculated spectra several flux integrals and radiation damage parameters were derived for the region of the azimuthal flux maxima in the mid-planes for different radial positions between core and biological shield, especially, at the inner and outer surfaces of the PV walls, at the (1/4)-PV-thickness and at the surveillance positions. Fissionable materials are often used as activation detectors for neutron fluence measurements. To get the real value the analysis demands to take into account the gamma induced fissions. Therefore, the part of these fissions in the total number of fissions was estimated for the detector reactions 237Np(n,f) and 238U(n,f) in the calculated neutron/gamma fields. It has been found that considerable corrections of the neutron fluence measurements can be necessary, especially in case of 238U(n,f). Most of the calculations were performed using a three-dimensional synthesis of 2D/1D-flux distributions obtained by the SN-code DORT with the BUGLE-96T group cross-section library. (orig.)

  18. 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.

  19. 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.

  20. Status of reactor pressure vessel embrittlement study in Japan

    International Nuclear Information System (INIS)

    Since the construction of Japanese first commercial nuclear power plant in 1966, 52 nuclear power plants have been commissioned in Japan to commercial operation. Japanese first nuclear power plant has now been service for 30 years and the aging of nuclear power plants is steadily progressing in general. Under these circumstances, the Japan Power Engineering and Inspection Corporation (JAPEIC) is executing, under consignment by the Ministry of International Trade and Industry (MITI), the development and verification test programs for plant integrity evaluation technology by which nuclear power plant aging can be appropriately handled. This paper shows the outline of study dealing with embrittlement of RPV caused by neutron irradiation, as one of the activity of JAPEIC. The embrittlement of RPV caused by neutron irradiation is manifested as a shift of transition temperature and as a reduction in Upper Shelf Energy (USE). In JAPEIC, the study dealing with a shift of transition temperature was conducted in the ''Reactor Pressure Vessel Pressurized Thermal Shock Test Project (the PTS Project)'', and the study dealing with a reduction in USE has been conducted in the ''Nuclear Power Plant Life Management Technology (the PLIM Project)''. And the reconstitution technology of surveillance test specimen has been conducted in PLIM Project as one of the measures to improve monitoring above material characteristic changes. The integrity evaluation under the Pressurized Thermal Shock (PTS) events including the effect of neutron irradiation embrittlement was initiated in 1983 FY as the PTS Project and was completed in the 1991 FY. The study verified that plant integrity could be assured at not only the end of design life, but also an extended service life even when the severest PTS events were postulated. The PLIM Project, designed to develop and verify the integrity evaluation technology dealing with reduction of USE by neutron irradiation, was started in the 1996 FY as a 10

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

    International Nuclear Information System (INIS)

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

  2. Prediction of BWR performance under the influence of Isolation Condenser-using RAMONA-4 code

    International Nuclear Information System (INIS)

    The purpose of the Boiling Water Reactor (BWR) Isolation Condenser (IC) is to passively control the reactor pressure by removing heat from the system. This type of control is expected to reduce the frequency of opening and closing of the Safety Relief Valves (SRV). A comparative analysis is done for a BWR operating with and without the influence of an IC under Main Steam Isolation Valve (MSIV) closure. A regular BWR, with forced flow and high thermal power, has been considered for analysis. In addition, the effect of ICs on the BWR performance is studied for natural convection flow at lower power and modified riser geometry. The IC is coupled to the steam dome for the steam inlet flow and the Reactor Pressure Vessel (RPV) near the feed water entrance for the condensate return flow. Transient calculations are performed using prescribed pressure set points for the SRVs and given time settings for MSIV closure. The effect of the IC on the forced flow is to reduce the rate of pressure rise and thereby decrease the cycling frequency ofthe SRVS. This is the primary objective of any operating IC in a BWR (e.g. Oyster Creek). The response of the reactor thermal and fission power, steam flow rate, collapsed liquid level, and core average void fraction are found to agree with the trend of pressure. The variations in the case of an active IC can be closely related to the creation of a time lag and changes in the cycling frequency of the SRVS. An analysis for natural convection flow in a BWR indicates that the effect of an IC on its transient performance is similar to that for the forced convection system. In this case, the MSIV closure, has resulted in a lower peak pressure due to the magnitude of reduced power. However, the effect of reduced cycling frequency of the SRV due to the IC, and the time lag between the events, are comparable to that for forced convection

  3. BWR MARK I pressure suppression pool mixing and stratification analysis using GOTHIC lumped parameter modeling methodology

    International Nuclear Information System (INIS)

    As a part of the GOTHIC (GOTHIC incorporates technology developed for the electric power industry under the sponsorship of EPRI.) Fukushima Technical Evaluation project (EPRI, 2014a, b, 2015), GOTHIC (EPRI, 2014c) has been benchmarked against test data for pool stratification (EPRI, 2014a, b, Ozdemir and George, 2013). These tests confirmed GOTHIC’s ability to simulate pool mixing and stratification under a variety of anticipated suppression pool operating conditions. The multidimensional modeling requires long simulation times for events that may occur over a period of hours or days. For these scenarios a lumped model of the pressure suppression chamber is desirable to maintain reasonable simulation times. However, a lumped model for the pool is not able to predict the effects of pool stratification that can influence the overall containment response. The main objective of this work is on the development of a correlation that can be used to estimate pool mixing and stratification effects in a lumped modeling approach. A simplified lumped GOTHIC model that includes a two zone model for the suppression pool with controlled circulation between the upper and lower zones was constructed. A pump and associated flow connections are included to provide mixing between the upper and lower pool volumes. Using numerically generated data from a multidimensional GOTHIC model for the suppression pool, a correlation was developed for the mixing rate between the upper and lower pool volumes in a two-zone, lumped model. The mixing rate depends on the pool subcooling, the steam injection rate and the injection depth

  4. Experimental studies on heat transfer in external cooling of the reactor pressure vessel

    International Nuclear Information System (INIS)

    The filling of the reactor cavity by accidental initiation of the containment spray system, while reactor is in full power, could have severe consequences. If the relatively cold water suddenly cools down the wall of fully pressurized reactor pressure vessel and a crack is assumed to be located on the outer surface of the vessel, the induced thermal stresses might damage the pressure vessel wall. The effects of the inadvertent cooling and pressurized thermal shock (PTS) were studied experimentally at Lappeenranta University of Technology and the heat transfer coefficients gained from the experimental results were compared with calculations. (author)

  5. FAVOR: A new fracture mechanics code for reactor pressure vessels subjected to pressurized thermal shock

    International Nuclear Information System (INIS)

    This report discusses probabilistic fracture mechanics (PFM) analysis which is a major element of the comprehensive probabilistic methodology endorsed by the NRC for evaluation of the integrity of Pressurized Water Reactor (PWR) pressure vessels subjected to pressurized-thermal-shock (PTS) transients. It is anticipated that there will be an increasing need for an improved and validated PTS PFM code which is accepted by the NRC and utilities, as more plants approach the PTS screening criteria and are required to perform plant-specific analyses. The NRC funded Heavy Section Steel Technology (HSST) Program at Oak Ridge National Laboratories is currently developing the FAVOR (Fracture Analysis of Vessels: Oak Ridge) PTS PFM code, which is intended to meet this need. The FAVOR code incorporates the most important features of both OCA-P and VISA-II and contains some new capabilities such as PFM global modeling methodology, the capability to approximate the effects of thermal streaming on circumferential flaws located inside a plume region created by fluid and thermal stratification, a library of stress intensity factor influence coefficients, generated by the NQA-1 certified ABAQUS computer code, for an adequate range of two and three dimensional inside surface flaws, the flexibility to generate a variety of output reports, and user friendliness

  6. Deterministic and probabilistic analysis of a reactor pressure vessel subjected to pressurized thermal shocks

    International Nuclear Information System (INIS)

    Highlights: • Deterministic and probabilistic methods are used to analyze a reactor pressure vessel. • Assuming shallower cracks can be more conservative than assuming deeper ones. • Master Curve methods are implemented in FAVOR for fracture toughness analysis. • Master Curve method is more realistic in modeling fracture toughness. • Warm prestressing effect decreases failure probability significantly. - Abstract: Both deterministic and probabilistic methods are used to analyze a reference reactor pressure vessel (RPV) subjected to pressurized thermal shocks (PTSs). The FAVOR code was applied to calculate the probabilities for crack initiation and failure of a RPV subjected to two PTS transients, by considering different crack types, sizes and orientations. The Master Curve methods are implemented in the FAVOR code for a more realistic consideration of fracture toughness of the irradiated RPV. The analysis shows that a postulated underclad crack is the most conservative crack assumption. Assuming shallower cracks can be more conservative than deeper ones due to the fact that both KI and KIC at the crack tip increase with crack depth. Considering the warm prestressing effect (WPS) reduces the failure probability by more than two orders of magnitude. In this analysis, the FAVOR model for the calculation of fracture toughness is more conservative than the Master Curve method. But the Master Curve method is more realistic than the FAVOR model and thus its application is recommended

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

    International Nuclear Information System (INIS)

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

  8. 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)

  9. The use of pressure testing to ensure the safety of fatigue loaded pressure vessels

    International Nuclear Information System (INIS)

    An attempt is made to analyse the value of the pressure test as a proof test in the presence of undetected flaws. The R curve concept has been applied as follows: (i) The crack driving force due to test pressure is computed as a function of the postulated crack size. The value of crack driving force is conservatively adopted as a material resistance value for vessels which did not fail during testing in spite of having a crack with an assumed size. (ii) The crack growth during the operation corresponding to a given fatigue loading is estimated. The crack driving force for the extended crack under maximum operational loads is compared to the above-mentioned resistance force. (iii) The ratio of test pressure to operational pressure, p√p, needed for stability of the crack, is computed as a function of initial crack size and fatigue loading. In numerical examples axial semi-elliptical surface cracks in cylinders are analysed by the application of LEFM. Models have been developed to simulate fatigue crack growth generated by cyclic pressure and thermal transients. In both cases a non-failure in a pressure test seems to provide a proof of safety during subsequent service periods, if the size of the initial flaw during the test is within certain limits, which can be determined as a function of the fatigue loading and the ratio p√p. (author)

  10. 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.

  11. Residual stresses in a weldment of pressure vessel steel

    International Nuclear Information System (INIS)

    A study was made of the distribution of residual stresses around a typical weld from a light water reactor pressure vessel by an X-ray double-exposure camera technique. So that the magnitude, sign, and distribution of the residual stresses were as similar as possible to those found in practice, a wide, full-thickness specimen of A533B Cl 1 steel containing a submerged-arc weld was stress-relief annealed. To obtain a three-dimensional distribution of the stresses the specimen was examined at different levels through the thickness. Following the removal of material by milling, the specimen surface was electropolished to free it from cold work. Corrections have been made to take into account specimen relaxation. To completely define the original stress system it is desirable also to measure the change in curvature on removing a layer of material. Unless this is done assumptions must be made which complicate the calculations unnecessarily. This became apparent after the experimental work was completed. In the centre of the plate the methods of correction which can be used are sensitive to errors in the measurements. The corrected results show that the dominant residual stress is perpendicular to the weld. It is positive at the surfaces and negative in the centre of the plate. The maximum value can reach the yield stress. The residual stresses in the weld metal can locally vary considerably: from 100 to 350N/mm2 over a distance of 5mm. Such large variations have been found to coincide with the heat-affected zones of the individual weld runs. (author)

  12. Fatigue properties of reactor pressure vessel steel and damage evaluation

    International Nuclear Information System (INIS)

    Low-cycle fatigue tests were carried out on a reactor pressure vessel steel at 290degC in order to investigate an evaluation method for fatigue damage and fatigue straining histories. The initiation behavior of surface microcracks was examined, and the following results were obtained. (1) The initiation of surface cracks begins from the early stage of fatigue life and the crack density (number of cracks per unit surface area) continues to increase throughout the fatigue life. (2) The increasing behavior of the crack density can be correlated with the Coffin-Manson type parameter ΔεP·N0.6 irrespective of an applied strain range (Δεt). Therefore, fatigue straining histories represented by ΔεP·N0.6 can be estimated from the measurement of surface crack density. (3) The grain boundary is one of the preferential sites of crack initiation, and micro-cracks which just initiated have a tendency to propagate along the grain boundary. Etching properties of fatigue damaged specimens were alos studied and were found to change with damage accumulation. The results obtained can be summarized as follows. (4) Many random-shape pits appeared only on the surface of the fatigue damaged material by the electrochemical etching. These random-shape pits initiated mainly at the triple point of grain boundaries and grew preferentially along the grain boundaries. These pits can be considered as a result from degradation of corrosion resistance of the demaged material cuased by microstructural changes at or near grain boundaries during fatigue. (5) The area fraction of corroded region on the electrochemically etched surface shows a single correlation with ΔεP· N0.6, and quantitative evaluation of fatigue damage accumulation can be made nondestructively by the electrochemical etching. (author)

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

    International Nuclear Information System (INIS)

    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

  14. EASY 5 BWR simulation model for digital feedwater control design

    International Nuclear Information System (INIS)

    The development of a BWR simulation model in support of a program to design and evaluate the digital feedwater control system for the Monticello Boiling Water Reactor (BWR) is described. This model was developed in the EASY5 simulation language in conjunction with EPRI's Modular Modeling System (MMS) two-phase Library. The model consists of three main elements: the BWR reactor vessel module, the feedwater system model, and the steamline model. Transient results for the BWR vessel module and the feedwater system model are presented

  15. The use of surfacing in the construction of pressure vessels for nuclear reactors

    International Nuclear Information System (INIS)

    The design of the pressure vessel of the VVER 440 reactor is based on the extensive application of surfacing; the weight of the deposits amounts to 7.6% of the weight of the shell and the top end of the vessel. Corrosion-resisting austenitic deposits of type 19%Cr-10%Ni (stabilized with Nb) are made on the entire internal surface of the pressure vessel. 'Strong' austenitic deposits are made on the weld areas of the actual parts of the pressure vessel (made of the 15Kh2MFA steel) to which components of austenitic or carbon steels are welded. The austenitic deposits which represent the dominant proportion of the total weight of the deposited metal are made primarily by automatic submerged-arc surfacing using strip electrodes. The quality of the deposits satisfies the requirements imposed on the pressure vessels of nuclear reactors. (author)

  16. Investigations for WWER reactor pressure vessel neutron exposure evaluation in CSFR

    International Nuclear Information System (INIS)

    The self-consistent results of WWER-440 reactor pressure vessel monitoring correspond to the mockup methodology. The calculations underestimate the density of neutron flux impinging on the pressure vessel inner wall in comparison with the experimental data evaluated by means of the mockup. In results adopted for reactor pressure vessel dosimetry, the effects of the displacer (effective decreasing of the water density in the mockup) and of air gaps on measurements over the pressure vessel thickness were neglected. The influence of the LR-0 tank (16 mm thick aluminium vessel) was disregarded, it was also estimated by experiment. In the mockups of WWER-1000 geometry a background of albedo neutrons was found. The effect was suppressed using a thick axial reflector. (author) 12 figs., 5 tabs., 9 refs

  17. Computational methods for fracture analysis of heavy-section steel technology (HSST) pressure vessel experiments

    International Nuclear Information System (INIS)

    This paper summarizes the capabilities and applications of the general-purpose and special-purpose computer programs that have been developed for use in fracture mechanics analyses of HSST pressure vessel experiments. Emphasis is placed on the OCA/USA code, which is designed for analysis of pressurized-thermal-shock (PTS) conditions, and on the ORMGEN/ADINA/ORVIRT system which is used for more general analysis. Fundamental features of these programs are discussed, along with applications to pressure vessel experiments

  18. Thermal-hydraulic instability of the natural circulation BWR. 6. Occurrence condition and mechanism of the instability at the higher system pressure

    International Nuclear Information System (INIS)

    Experiments have been conducted to investigate thermal-hydraulic instabilities at the higher system pressure ranging from 1.0 to 7.2 MPa in a boiling natural circulation loop with a chimney. A test facility used in this experiments was designed and constructed to have non-dimensional values which are nearly equal to those of natural circulation BWR flow dynamics. Stability maps in reference to the system pressure, the channel inlet subcooling, and heat flux are presented. This instability mechanism is classified into the density wave oscillations that oscillation period is one to two times the time required for a bubble generated in the channel to travel through the chimney, and different from the flashing induced instability at the lower system pressure. The difference from other phenomena such as flow pattern transient oscillations and natural circulation oscillations are discussed by investigating the transient flow pattern and the response of momentum energy to driving force of the circulation. (author)

  19. Investigation of the failure of a reactor pressure vessel by plastic instability

    International Nuclear Information System (INIS)

    A possible consequence of a core meltdown accident in a pressurized water reactor is the failure of the reactor pressure vessel under high internal pressure. With the aid of the finite element program ABAQUS and using a material model of the thermo-plasticity for large deformation, the failure of the reactor pressure vessel due to plastic instability was examined. It was apparent from the finite element calculations that solely due to reduction in strength of the material, even for internal wall temperatures clearly below the core melt; of about 2000 C, the critical internal pressure can fall to values which are lower than the working pressure. With the aid of simplified geometry, a lower limit for the pressure at failure of the reactor pressure vessel can be calculated. (orig./HP)

  20. Pressurized thermal shock. Thermo-hydraulic conditions in the CNA-I reactor pressure vessel

    International Nuclear Information System (INIS)

    In this paper we analyze several reports issued by the Utility (Nucleo Electrica S.A.) and related to Reactor Pressure Vessel (RPV) phenomena in the CNA-I Nuclear Power Plant. These analyses are aimed at obtaining conclusions and establishing criteria ensuring the RPV integrity. Special attention was given to the effects ECCS cold-water injection at the RPV down-comer leading to pressurized thermal shock scenarios. The results deal with hypothetical primary system pipe breaks of different sizes, the inadvertent opening of the pressurizer safety valve, the double guillotine break of a live steam line in the containment and the inadvertent actuation pressurizer heaters. Modeling conditions were setup to represent experiments performed at the UPTF, under the hypothesis that they are representative of those that, hypothetically, may occur at the CNA-I. No system scaling analysis was performed, so this assertion and the inferred conclusions are no fully justified, at least in principle. The above mentioned studies, indicate that the RPV internal wall surface temperature will be nearly 40 degree. It was concluded that they allowed a better approximation of PTS phenomena in the RPV of the CNA-I. Special emphasis was made on the influence of the ECCS systems on the attained RPV wall temperature, particularly the low-pressure TJ water injection system. Some conservative hypothesis made, are discussed in this report. (author)

  1. Response of a LWR pressure vessel to severe-accident loadings

    Energy Technology Data Exchange (ETDEWEB)

    Ju, F.D.; Bennett, J.G.; Anderson, C.A.

    1982-01-01

    In the recent emphasis on nuclear safety, structural studies of nuclear reactor vessels have been directed toward evaluating their response during severe loading incidents or accidents including even core meltdown - however improbable these accidents may be. The present paper will address some of these problems. The ultimate load carrying capacity of an unflawed nuclear pressure vessel is estimated. The measure of the maximum pressure that the vessel can resist during quasistatic loading is a useful quantitative estimate of overall vessel strength. The paper than analyzes two structural problems during a hypothetical meltdown. In the initial stage, the molten core mixture drops into the lower portion of the pressure vessel, resulting in both temperature and pressure rises. Subsequently, a vapor explosion may occur as a result of the molten metal coming in sudden contact with the water in the lower portion of the vessel. The explosion is postulated to propel a slug of molten metalup the vessel barrel that eventually impacts the upper head of the vessel potentially generating missiles in the containment building. The reactor vessel at Indian Point, New York is used as a prototype of this analysis.

  2. Response of a LWR pressure vessel to severe-accident loadings

    International Nuclear Information System (INIS)

    In the recent emphasis on nuclear safety, structural studies of nuclear reactor vessels have been directed toward evaluating their response during severe loading incidents or accidents including even core meltdown - however improbable these accidents may be. The present paper will address some of these problems. The ultimate load carrying capacity of an unflawed nuclear pressure vessel is estimated. The measure of the maximum pressure that the vessel can resist during quasistatic loading is a useful quantitative estimate of overall vessel strength. The paper than analyzes two structural problems during a hypothetical meltdown. In the initial stage, the molten core mixture drops into the lower portion of the pressure vessel, resulting in both temperature and pressure rises. Subsequently, a vapor explosion may occur as a result of the molten metal coming in sudden contact with the water in the lower portion of the vessel. The explosion is postulated to propel a slug of molten metalup the vessel barrel that eventually impacts the upper head of the vessel potentially generating missiles in the containment building. The reactor vessel at Indian Point, New York is used as a prototype of this analysis

  3. 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

  4. Improvement in reactor pressure vessel reliability through assembly production

    International Nuclear Information System (INIS)

    The importance of the Framatome nuclear programme requires the implementation of significant human and equipment resources for the manufacturing of a large number of reactor vessels, at a rate of six vessels per year. The time needed to fabricate one vessel is approximately three years and as many as eighteen vessels can be present, at the same time, on the assembly line in Framatome workshops. In order to cope with this mass-type production plan, Framatome is geared to transform most of the original manual welding operations into automatic welding processes, which result in a reduction of the number of weld defects and therefore in the number of required weld repairs. Another benefit is the marked improvement in the welders' working conditions. Both resulted in improving component reliability. These developments are described. (author)

  5. Non-invasive method and apparatus for measuring pressure within a pliable vessel

    Science.gov (United States)

    Shimizu, M. (Inventor)

    1983-01-01

    A non-invasive method and apparatus is disclosed for measuring pressure within a pliable vessel such as a blood vessel. The blood vessel is clamped by means of a clamping structure having a first portion housing a pressure sensor and a second portion extending over the remote side of the blood vessel for pressing the blood vessel into engagement with the pressure sensing device. The pressure sensing device includes a flat deflectable diaphragm portion arranged to engage a portion of the blood vessel flattened against the diaphragm by means of the clamp structure. In one embodiment, the clamp structure includes first and second semicylindrical members held together by retaining rings. In a second embodiment the clamp structure is of one piece construction having a solid semicylindrical portion and a hollow semicylindrical portion with a longitudinal slot in the follow semicylindrical portion through which a slip the blood vessel. In a third embodiment, an elastic strap is employed for clamping the blood vessel against the pressure sensing device.

  6. Structural Integrity of Gas-Filled Composite Overwrapped Pressure Vessels Subjected to Orbital Debris Impact

    Science.gov (United States)

    Telichev, Igor; Cherniaev, Aleksandr

    Gas-filled pressure vessels are extensively used in spacecraft onboard systems. During operation on the orbit they exposed to the space debris environment. Due to high energies they contain, pressure vessels have been recognized as the most critical spacecraft components requiring protection from orbital debris impact. Major type of pressurized containers currently used in spacecraft onboard systems is composite overwrapped pressure vessels (COPVs) manufactured by filament winding. In the present work we analyze the structural integrity of vessels of this kind in case of orbital debris impact at velocities ranging from 2 to 10 km/s. Influence of such parameters as projectile energy, shielding standoff, internal pressure and filament winding pattern on COPVs structural integrity has been investigated by means of numerical and physical experiments.

  7. Problems in manufacturing and transport of pressure vessels of integral reactors

    International Nuclear Information System (INIS)

    Integral water-cooled reactors are typical with eliminating large-diameter primary pipes and placing primary components, i.e. steam generators and pressurizers in reactor vessels. This arrangement leads to reactor pressure vessels of large dimensions: diameters, heights and thick walls and subsequently to great weights. Thus, even medium power units have pressure vessels which are on the very limit of present manufacturing capabilities. Principal manufacturing and inspection operations as well as pertinent equipment are concerned: welding, cladding, heat treatment, machining, shop-handling, non-destructive testing, hydraulic pressure tests etc. Tile transport of such a large and heavy component makes a problem which effects its design as well as the selection of the plant site. Railway, road and ship are possible ways of transport each of them having its advantages and limitations. Specific features and limits of the manufacture and transport of large pressure vessels are discussed in the paper. (author)

  8. Preliminary study of an expert system for mechanical design of a pressure vessel

    International Nuclear Information System (INIS)

    This paper describes a preliminary study of an expert system for mechanical design of a pressure vessel. The system supports the framework for the conceptual mechanical design from the initial stages within the design procedures. ASME Boiler and Pressure Vessel Code Section VIII Division 1 were applied as a design rule. The proposed methodology facilitates the development of knowledge base acquisition, knowledge base construction and the prototype implementation. This study characterizes a knowledge base (procedure) of mechanical design of a pressure vessel subjected to internal pressure including all design parameters; i.e. temperature, shell thickness, selection of materials of constructions, stress analysis procedure, support and ancillary items. The rationalization of the mechanical design is shown in the form of a schematic flow diagram. A Kappa PC expert system shell is used as a tool to develop the prototype software. It provides graphical representation for creating objects, hierarchies and rules for knowledge base used in pressure vessel design. (Author)

  9. 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

  10. Evaluation of Progressive Failure Analysis and Modeling of Impact Damage in Composite Pressure Vessels

    Science.gov (United States)

    Sanchez, Christopher M.

    2011-01-01

    NASA White Sands Test Facility (WSTF) is leading an evaluation effort in advanced destructive and nondestructive testing of composite pressure vessels and structures. WSTF is using progressive finite element analysis methods for test design and for confirmation of composite pressure vessel performance. Using composite finite element analysis models and failure theories tested in the World-Wide Failure Exercise, WSTF is able to estimate the static strength of composite pressure vessels. Additionally, test and evaluation on composites that have been impact damaged is in progress so that models can be developed to estimate damage tolerance and the degradation in static strength.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  13. Coupled thermomechanical response of a pressure vessel lower head to relocated corium

    International Nuclear Information System (INIS)

    Previous investigations of the pressure vessel lower head subjected to relocated corium, performed as part of the OECD-sponsored Three Mile Island Unit 2 Vessel Investigation Project (TMI-2 VIP), indicated the vessel would fail under the high pressures and extreme temperatures expected from corium relocation. The TMI-2 lower head, however, remained intact and experienced no finite deformation, despite the settling of some 19 tonnes of previously molten material on top of it. The inconsistency between model predictions and TMI-2 vessel behavior were ascribed to an incomplete understanding of corium interaction with the lower head as well as insufficient coupling between vessel deformation and thermal loading. This paper summarizes results obtained to date and efforts to couple the thermal and mechanical response of the corium, overlaying water pool and vessel lower head

  14. Nonlinear finite element analysis of mechanical characteristics on CFRP composite pressure vessels

    International Nuclear Information System (INIS)

    CFRP(Carbon Fibre Reinforced Plastic) composite pressure vessel was calculated using finite element program of ANSYS for their mechanical characteristics in this paper. The elastic-plastic model and elements of Solid95 were selected for aluminium alloys of gas cylinder. Also liner-elastic model and layer elements of Shell99 were adopted for carbon fibre/epoxy resin. The stress state of CFRP composite pressure vessel was calculated under different internal pressures include pre-stressing pressures, working pressures, test hydraulic pressures, minimum destructive pressures etcetera to determine the size of gas cylinder and layer parameter of carbon fibre. The mechanical characteristics CFRP composite vessel could were using to design and test of gas cylinder. Numerical results showed that finite element model and calculating method were efficient for study of CFRP gas cylinder and useful for engineering design.

  15. 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.

  16. 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.

  17. Probabilistic assessment of a reactor pressure vessel subjected to pressurized thermal shocks by using crack distributions

    International Nuclear Information System (INIS)

    Highlights: • Probabilistic methods are used to analyze a reactor pressure vessel. • Crack distribution data from the decommissioned plants, Shoreham and PVRUF is used. • Weld type, size and its manufacturing process are also considered. • Embedded and surface short cracks result in the highest probability for failure. - Abstract: Probabilistic methods are used to analyze a reactor pressure vessel (RPV) subjected to pressurized thermal shocks (PTSs) initiated by a small loss-of-coolant accident (SLOCA) and a medium loss-of-coolant accident (MLOCA). The FAVOR code is applied to calculate the probabilities for crack initiation and failure by considering crack distributions based on cracks observed in the Shoreham and PVRUF RPVs in the U.S. The crack parameters, i.e. crack density, depth, aspect ratio, orientation and location are assumed as random variables following different distributions. The Vflaw code is used to generate FAVOR input files for the crack distribution data from the decommissioned plants. Weld type, size and its manufacturing process are also considered in the calculation. In this paper it is shown that the calculated failure probability of the RPV subjected to the SLOCA is higher than that subjected to the MLOCA due to different loading. The failure probabilities are compared with those based on a different crack assumption. Among the analyzed cracks, the embedded crack with a depth of 6.83 mm and surface crack with a depth of 5.13 mm result in the highest probability for failure. Maximum stress intensity factors of the simulated cracks range from 36 MPa m0.5 to 91 MPa m0.5 for the MLOCA and from 30 to 41 MPa m0.5 for the SLOCA, respectively. We conclude that considering the observed crack distribution in probabilistic PTS analyses may lead to higher failure probabilities than by assuming cracks of specific size

  18. Tecnique for probabilistic calculation of brittle fracture of power plant pressure vessels

    International Nuclear Information System (INIS)

    Technique for probabilistic calculation of brittle fracture of power plant pressure vessels is presented. Effect of static spread in data on mechanical material properties, defect sizes and errors of nondestructive test means on the accuracy of brittle fracture time prediction is taken account of. Example of probabilistic calculation of nuclear reactor vessel fracture during its operation is given

  19. Gas-cooled HTR reactor installed in a pressure vessel cavern

    International Nuclear Information System (INIS)

    A pebble-bed reactor in a pressure vessel cavern is described which has a reflector which in case of accidents with pressure equalisation between cold gas and hot gas transfers the resulting loads to a lateral thermal shield constructed in the form of a pressure-tight metal cylinder. (TK)

  20. Multilayer Pressure Vessel Materials Testing and Analysis Phase 2

    Science.gov (United States)

    Popelar, Carl F.; Cardinal, Joseph W.

    2014-01-01

    To provide NASA with a suite of materials strength, fracture toughness and crack growth rate test results for use in remaining life calculations for the vessels described above, Southwest Research Institute® (SwRI®) was contracted in two phases to obtain relevant material property data from a representative vessel. An initial characterization of the strength, fracture and fatigue crack growth properties was performed in Phase 1. Based on the results and recommendations of Phase 1, a more extensive material property characterization effort was developed in this Phase 2 effort. This Phase 2 characterization included additional strength, fracture and fatigue crack growth of the multilayer vessel and head materials. In addition, some more limited characterization of the welds and heat affected zones (HAZs) were performed. This report

  1. Neutron flux and fluence determination for BWR reactors

    International Nuclear Information System (INIS)

    Measurements of gamma emission rates from Fe and Cu dosimeters extracted from a BWR type reactor vessel were carried out in order to determine their total activity. The dosimeter's activity is related to the neutron flux there by taking into account the reactor material's embrittlement caused by neutron bombardment. The dosimeters were taken out after the first reactor operation cycle. From gamma radioactivity measurements of these dosimeters, neutron flux and fluence were calculated. These parameters are used in the determination of shift and adjusted reference temperature values needed for the development of pressure-temperature curves used during reactor operation

  2. VVER reactor pressure vessels dosimetry for NPP lifetime management

    International Nuclear Information System (INIS)

    Assessment of the current state and lifetime of the Kozloduy NPP VVER-type reactors has been carried out by monitoring the irradiation exposure of reactor vessels and metal surveillance specimens. The neutron fluence determination is based on three-dimensional neutron transport calculation by the TORT code using the problem-oriented neutron cross-section library BGL, and by the MCNP code using the continuous energy dependence library DLC200. The fluence data are validated by measurements of the induced activity of threshold foil detectors. Reactor core loading patterns are determined to achieve minimal reactor vessel neutron exposure. (author)

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

    International Nuclear Information System (INIS)

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

  4. The long-term properties of concrete used in prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Data are presented on the late-life properties of concretes used in prestressed concrete pressure vessels (PCPV) and containment structures. The effects of ageing under simulated PCPV conditions are discussed. (author)

  5. Research progress and recommendations on reactor pressure vessel integrity under hypothetical core melt down accident

    International Nuclear Information System (INIS)

    Background: It is very important to ensure the integrity of the reactor pressure vessel under core melt down accident. The high-temperature creep failure is the main failure mode of the reactor pressure vessel under core melt down accident. Purpose: This paper is to present an overview of research status and progress on high-temperature creep behavior of reactor pressure vessel considering the hypothetical core melt down scenario. Methods: Emphasis is placed on accomplished achievements in creep tests, scale model experiments and numerical simulation, and the domestic newly research productions on high-temperature creep behavior of reactor pressure vessel structure integrity. Conclusions: This paper also discusses the limitations of existing researches and indicates future research directions, such as multi-axis tensile tests, analysis of three-dimensional coupling temperature field, scaled model tests, and so on. (authors)

  6. Nozzle protection device in a pressure vessel of nuclear power plant

    International Nuclear Information System (INIS)

    Purpose: To improve the security of a pressure vessel by reducing the thermal stress in a nozzle. Constitution: Water with a small temperature difference to the coolant in a pressure vessel is charged into the annular portion between a nozzle for injecting a cold water in the pressure vessel and a thermal sleeve in the nozzle. An injection pipe for extracting a part of water for a reactor core from the discharging pipe of a recycle pump in the nuclear power plant and injecting the same into the annular portion is provided. The nozzle corner can thus be cooled by cold water supplied from the safe end passing through the inside of the thermal sleeve, whereby no thermal stress is resulted therein to improve the security of the pressure vessel. (Seki, T.)

  7. On the domestically-made heavy forging for reactor pressure vessels of PWR nuclear power plant

    International Nuclear Information System (INIS)

    The present situation of the foreign heavy forgings for nuclear reactor pressure vessels and the heavy forgings condition which is used for the Qinshan 300MWe nuclear power plant are described. Some opinions of domestic products is proposed

  8. Selected bibliography on pressure vessels for light-water-cooled power reactors (LWRs)

    International Nuclear Information System (INIS)

    Abstracts on LWR pressure vessels are arranged in the following categories: general, design, materials technology, fabrication techniques, inspection and testing, and failures. Author, keyword, and KWIC (keyword-in-content) indices are provided. (U.S.)

  9. Mass optimization of a small pressure vessel using metal/FRP (fiber reinforced polymers) hybrid structures

    International Nuclear Information System (INIS)

    In hybrid pressure vessels, composite (Fiber) is wound over a metallic liner (Steel/Aluminum) in hoop direction. In this concept of hybrid pressure vessel structure, metallic liner takes all the axial loads and fiber reinforced polymers (FRP/sub s/) takes load in circumferential (Hoop) direction. Hybrid structures combine the relatively high shear stiffness and ductility of metal alloy with high specific stiffness, strength and fatigue properties of FRP/sub s/. The relatively simple methods for producing hybrid structures circumvent the need for the complex and expensive equipment that is used for advanced composites processing. This paper presents an efficient way of designing a hybrid pressure vessel where prime concern is weight reduction over an equivalent aluminum structure and investigates various methodologies regarding combinations of metals and FRP/sub s/ for optimization of a given pressure vessel. For this purpose we adopted two different methods of simulation one is computer simulation using ANSYS and other is experimental verification by hydrostatic testing of manufactured pressure vessel. Two different pressure vessels one with aluminum liner and other with steel liner were fabricated. Kevlar 49/epoxy was wrapped around the liners in hoop direction. Both the pressure vessels were put into hydrostatic test. Strains were measured during the test and then converted into corresponding stresses. Results of hydrostatic test were quite in favor of the ANSYS results. In this way we have successfully designed, manufactured and tested the Hybrid pressure vessel saving almost 40% weight in case of aluminum liner and 43.6% in case of steel liner. (author)

  10. Assessment of relative contributions from different mechanisms to radiation embrittlement of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Experimental data on radiation embrittlement in pressure vessel steels of both Russian and American grades, obtained by the authors and also taken from the literature, have been analyzed to assess the relative contributions from the following mechanisms: radiation-induced hardening, inter- and intragranular segregation of impurities at precipitate/matrix interfaces. It is demonstrated that radiation-induced intragranular segregation of impurities frequently provides a significant contribution to radiation embrittlement of pressure vessel steels. (orig.)

  11. Validation of flux and spectrum predictions for a pressure vessel wall environment

    International Nuclear Information System (INIS)

    Aging light water reactor pressure vessels change their fracture toughness properties due to neutron fluence. In order to improve and standardize surveillance procedures for reactor pressure vessels the US Nuclear Regulatory Commission has established the LWR-PV Surveillance Dosimetry Improvement Program. This program is being conducted in close cooperation with several US and European Laboratories. Irradiations for this program are being performed at ORNL at the PCA-PV benchmark facility and the ORR-PV facility

  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. Sampling of reactor pressure vessel and core internals

    International Nuclear Information System (INIS)

    Decommissioning and dismantling of nuclear power plants is a growing business as a huge number of plants built in the 1970's have now reached their lifetime. It is well known that dismantling a nuclear power plant means an extraordinary expense for the owner respectively operator. Beside the dismantling works for itself, the disposal of activated components and other nuclear waste is very expensive. What comes next is the fact that final disposal facilities are not available yet in most countries meaning a need for interim storage on-site in specially built facilities. It can be concluded that a special attention is paid on producing a minimal radioactive waste volume. For this, optimized dismantling and packaging concepts have to be developed. AREVA is proud of versatile experience in successfully dismantling nuclear components like core internals and reactor pressure vessel (RPV). The basis of a well-founded and optimized dismantling and packaging concept must always be the detailed knowledge of the radiological condition of the component to be and in the best case a 3D activation- model. For keeping the necessary sampling effort as small as possible, but simultaneously as efficient as possible, representative sampling positions are defined in advance by theoretical radiological examinations. For this, a detailed 3D-CAD-model of the components to be dismantled has proven very helpful and effective. Under these aspects a sampling of RPV and its components is necessary to verify the theoretically calculated radiological data. The obtained results of activation and contamination are taken into account for the optimized dismantling and packaging strategy. The precise 3D-activation-model will reduce the necessary number and type of final disposal containers as security factors are minimized leading to a lower shielding effort, too. Besides, components or even parts of components may be subject of release measurement. In the end, costs can be reduced. In this context

  14. Design Improvement of Double Pressure Vessel in the In-pile Test Section

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jintae; Heo, Sung-Ho; Joung, Chang-Young; Kim, Ka-Hye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    To carry out an irradiation test of nuclear fuels, a nuclear fuel test rig should be fabricated and installed in the in-pile test section (IPS), which is installed in the reactor hall. While carrying out an irradiation test, sealing out coolant which passes through the test rig is one of the most important issues. In particular, although the double pressure vessel is assembled with the IPS head by two o-rings and six bolts, 15.5 MPa of highly pressurized coolant leaks through the gap between the vessel and IPS head. Because the temperature of the coolant in the test loop is 300 .deg. C , and the pool of HANARO is 40 .deg. C, the double pressure vessel is necessary to insulate them. Therefore, a new design to prevent the leakage of coolant needs to be developed. In this study, EB welding technique is considered to assemble the double pressure vessel and the IPS head, and their mechanical design is modified to enable the welding process. In this study, an improved design for sealing out the coolant at the pressure boundary between the double pressure vessel and the IPS head has been developed. An EB weld is applied to seal out the pressure boundary, and its sealing performance is verified by NDE, a cross section test, and a hydraulic pressure test. From the verification test results, the improved design can be used in fabricating the IPS for a nuclear fuel irradiation test.

  15. Test system carrier for the ultrasonic testing of the area of connecting nozzles in the case of pressure vessels, in particular reactor pressure vessels from nuclear power plants

    International Nuclear Information System (INIS)

    In the invention at hand a system carrier for the ultrasonic testing of a reactor pressure vessel is described which enables a test for nozzle welds, pipe fitting welds and nozzle edges to be conducted with a single telescope arm. (RW)

  16. 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

  17. The decompression of pressurized vessels filled with two-phase fluids through pipes

    International Nuclear Information System (INIS)

    A unique fluid behavioral model which is applicable to many homogeneous multiphase fluids is used as a generic behavioral model to examine the decompression characteristics of pressurized vessels filled with single component two-phase fluids due to the breakage of connecting pipes. A corresponding analytic solution for the quasi steady state flow of two-phase fluids through pipes with friction is used in conjunction with a conventional vessel decompression methodology to evaluate the vessel pressure history and the mass flow rate efflux

  18. Pressure vessel inspection criteria based on fitness-for-purpose assessment

    International Nuclear Information System (INIS)

    The paper on pressure vessel inspection investigates the methodology required to establish an inspection strategy consistent with fracture mechanics analysis, i.e. to define allowable flaw sizes based on location within the vessel. The methodology is demonstrated using a sample problem for a typical pressurised water reactor pressure vessel, and shows the impact of certain assumptions on the inspection strategy. The results indicate that the flaw size varies with the shape of the assumed residual stress field and the through-thickness location. Also in general, the fracture mechanics evaluation allows flaws much larger than are allowed by the inspection acceptance criteria. (UK)

  19. Predicting Structural Behavior of Filament Wound Composite Pressure Vessel Using Three Dimensional Shell Analysis

    Science.gov (United States)

    Madhavi, M.; Venkat, R.

    2014-01-01

    Fiber reinforced polymer composite materials with their higher specific strength, moduli and tailorability characteristics will result in reduction of weight of the structure. The composite pressure vessels with integrated end domes develop hoop stresses that are twice longitudinal stresses and when isotropic materials like metals are used for development of the hardware and the material is not fully utilized in the longitudinal/meridional direction resulting in over weight components. The determination of a proper winding angles and thickness is very important to decrease manufacturing difficulties and to increase structural efficiency. In the present study a methodology is developed to understand structural characteristics of filament wound pressure vessels with integrated end domes. Progressive ply wise failure analysis of composite pressure vessel with geodesic end domes is carried out to determine matrix crack failure, burst pressure values at various positions of the shell. A three dimensional finite element analysis is computed to predict the deformations and stresses in the composite pressure vessel. The proposed method could save the time to design filament wound structures, to check whether the ply design is safe for the given input conditions and also can be adapted to non-geodesic structures. The results can be utilized to understand structural characteristics of filament wound pressure vessels with integrated end domes. This approach can be adopted for various applications like solid rocket motor casings, automobile fuel storage tanks and chemical storage tanks. Based on the predictions a composite pressure vessel is designed and developed. Hydraulic test is performed on the composite pressure vessel till the burst pressure.

  20. Technical report on material selection and processing guidelines for BWR [boiling water reactor] coolant pressure boundary piping: Final report

    International Nuclear Information System (INIS)

    This report provides the technical bases for the NRC staff's revised recommended methods to control the intergranular stress corrosion cracking susceptibility of BWR piping. For piping that does not fully comply with the material selection, testing, and processing guideline combinations of this document, varying degrees of augmented inservice inspection are recommended. This revision also includes guidance and NRC staff recommendations (not requirements) regarding crack evaluation and weld overlay repair methods for long-term operation or for continuing interim operation of plants until a more permanent solution is implemented

  1. Integrated surveillance specimen program for WWER-1000/V-320 reactor pressure vessels

    International Nuclear Information System (INIS)

    Surveillance specimen programs play an important role in reactor pressure vessel lifetime assessment as they should monitor changes in pressure vessel materials, mainly their irradiation embrittlement. Standard surveillance programs in WWER-1000/V-320 reactor pressure vessels have some deficiencies resulting from their design-nonuniformity of neutron field and even within individual specimen sets, large gradient in neutron flux between specimens and containers, lack of neutron monitors in most of containers and no suitable temperature monitors. Moreover, location of surveillance specimens does not assure similar conditions as the beltline region of reactor pressure vessels. Thus, Modified surveillance program for WWER-1000/V-320C type reactors was designed and realized in two units of NPP Temelin, Czech Republic. In this program, large flat type containers are located on inner wall of reactor pressure vessel in the beltline region that assures their practically identical irradiation conditions with critical vessel materials. These containers with inner dimensions of 210 x 300 mm have two layers of specimens; using inserts (10 x 10 x 14 mm) instead of fully Charpy size specimens allows irradiation of materials from several pressure vessels at once in one container. This design advantage has been used for the creation of the Integrated Surveillance Program for several WWER-1000 units-Temelin 1 + 2, Belene (Bulgaria), Rovno 3 + 4, Khmelnick 2, Zaporozhie 6 (Ukraine) and Kalinin 3 (Russia). Irradiation of these archive materials together with the IAEA reference steel JRQ (of ASTM A 533-B type) and reference steel VVER-1000 will allow to compare irradiation embrittlement of these materials and to obtain more reliable and objective results as no reliable predictive formulae exist up to no due to a higher content of nickel in welds. Irradiation of specimens from cladding region will help in the evaluation of resistance of pressure vessels against PTS regimes. (authors)

  2. Nuclear power plant with pressure vessel boiling water reactor VK-300 for district heating and electricity supply

    International Nuclear Information System (INIS)

    The viability for Russia of the Boiling Water Reactor (BWR) concept has been shown by a number of feasibility studies fulfilled for perspective sites with increased energy demands. Russia has long (31 year) successful experience in operation of NPPs with the vessel-type boiling reactor VK-50 which is located in the city of Dimitrovgrad. Taking into account the large Russian district heating market, it is expedient to apply this concept (BWR) not only for electricity supply, but also for district heating. This is a way to increase of nuclear power plant competitiveness along with good safety performance. The safety and protection of nuclear heat customer is guaranteed by reliable technical means which are well checked at Russian nuclear sites. (author)

  3. Testing of VVER reactor pressure vessels by TOFD method

    International Nuclear Information System (INIS)

    The Time of Flight Diffraction Method (TOFD) - one of the new testing methods capable to obtain the real dimensions of flaws - is presented in the paper.The laboratory experiments on samples with artificial flaws and samples with artificially prepared cracks confirmed the high accuracy of flaw through wall extent sizing by TOFD. This accuracy was confirmed by qualification of methods and systems used by Skoda JS for the in-service inspections of WWER 440 vessel circumferential weld. The qualification also confirmed the ability of TOFD to detect reliably flaws, which can are not reliably detected by standard pulse echo testing. Based on the result of experiments and qualification, the TOFD method shall be used routinely by Skoda JS for the inspection of vessel circumferential welds root area and for sizing of flaws exceeding the acceptance level

  4. Transmitted Ultrasound Pressure Variation in Micro Blood Vessel Phantoms

    OpenAIRE

    Qin, Shengping; Kruse, Dustin E; Ferrara, Katherine W.

    2008-01-01

    Silica, cellulose, and polymethylmethacrylate tubes with inner diameters of ten to a few hundred microns are commonly used as blood vessel phantoms in in vitro studies of microbubble or nanodroplet behavior during insonation. However, a detailed investigation of the ultrasonic fields within these micro-tubes has not yet been performed. This technical note provides a theoretical analysis of the ultrasonic fields within micro-tubes. Numerical results show that for the same tube material, the in...

  5. Numerical Simulation of Debris Cloud Propagation inside Gas-Filled Pressure Vessels under Hypervelocity Impact

    Science.gov (United States)

    Gai, F. F.; Pang, B. J.; Guan, G. S.

    2009-03-01

    In the paper SPH methods in AUTODYN-2D is used to investigate the characteristics of debris clouds propagation inside the gas-filled pressure vessels for hypervelocity impact on the pressure vessels. The effect of equation of state on debris cloud has been investigated. The numerical simulation performed to analyze the effect of the gas pressure and the impact condition on the propagation of the debris clouds. The result shows that the increase of gas pressure can reduce the damage of the debris clouds' impact on the back wall of vessels when the pressure value is in a certain range. The smaller projectile lead the axial velocity of the debris cloud to stronger deceleration and the debris cloud deceleration is increasing with increased impact velocity. The time of venting begins to occur is related to the "vacuum column" at the direction of impact-axial. The paper studied the effect of impact velocities on gas shock wave.

  6. Forging technology for large nuclear pressure vessel parts

    International Nuclear Information System (INIS)

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

  7. 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)

  8. Stress analysis and evaluation of a rectangular pressure vessel. [For equipment for sampling Hanford tank radwaste

    Energy Technology Data Exchange (ETDEWEB)

    Rezvani, M.A.; Ziada, H.H. (Westinghouse Hanford Co., Richland, WA (United States)); Shurrab, M.S. (Westinghouse Savannah River Co., Aiken, SC (United States))

    1992-10-01

    This study addresses structural analysis and evaluation of an abnormal rectangular pressure vessel, designed to house equipment for drilling and collecting samples from Hanford radioactive waste storage tanks. It had to be qualified according to ASME boiler and pressure vessel code, Section VIII; however, it had the cover plate bolted along the long face, a configuration not addressed by the code. Finite element method was used to calculate stresses resulting from internal pressure; these stresses were then used to evaluate and qualify the vessel. Fatigue is not a concern; thus, it can be built according to Section VIII, Division I instead of Division 2. Stress analysis was checked against the code. A stayed plate was added to stiffen the long side of the vessel.

  9. 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

  10. Comparison between CFD and acoustic methods in calculation of flow-induced loads in a reactor vessel at a simulated steam line failure in a BWR

    International Nuclear Information System (INIS)

    Det Norske Veritas has evaluated a method to analyse the pressure transient in Boiling Water Reactors after postulated main steam line break. The pipe break is postulated to occur in the vicinity of the reactor pressure vessel. The work was initiated by a pilot study in 1999, which was reported earlier in year 2000 by two SAQ reports. The Swedish Nuclear Power Inspectorate financed the work. At the first stage a validation of two calculation methods for flow induced dynamic loads performed. A method based on non-stationary potential flow (linear approach) was validated against Computational Fluid Dynamics (STAR-CD). The calculations were performed for a simplified geometry, the steam was considered as a perfect gas and the flow as isentropic. This report contains the description of the models, geometry, initial and boundary conditions and the medium. The theoretical background of the linear approach is presented. Calculated by the two methods oscillating pressures close to the steam dryer surface and dynamic, flow induced forces, acting on the steam dryer wall are presented and compared. Good agreement between the two methods was found concerning the pressure signal and the time dependent force acting on the steam dryer wall. The linear approach has a number of advantages comparing to the CFD-computations. Using CFD-technique requires significantly more computer resources and in addition a large amount of data needs to be transferred to the structural code. Furthermore this large amount of data makes practically impossible to use CFD-technique for calculation of non-elementary problems considering fluid-structure interaction (FSI). On the other hand it can be shown that the linear approach is connected to the acoustic pressure formulation used in commercial structural FEM-codes. This makes it possible to take FSI into consideration and reach a new, higher level of quality in calculations of the structural integrity of components and substructures in the RPV. The

  11. Brief account of the effect of overcooling accidents on the integrity of PWR pressure vessels

    International Nuclear Information System (INIS)

    The occurrence in recent years of several (PWR) accident initiating events that could lead to severe thermal shock to the reactor pressure vessel, and the growing awareness that copper and nickel in the vessel material significantly enhance radiation damage in the vessel, have resulted in a reevaluation of pressure-vessel integrity during postulated overcooling accidents. Analyses indicate that the accidents of concern are those involving both thermal shock and pressure loadings, and that an accident similar to that at Rancho Seco in 1978 could, under some circumstances and at a time late in the normal life of the vessel, result in propagation of preexistent flaws in the vessel wall to the extent that they might completely penetrate the wall. More severe accidents have been postulated that would result in even shorter permissible lifetimes. However, the state-of-the-art fracture-mechanics analysis may contain excessive conservatism, and this possibility is being investigated. Furthermore, there are several remedial measures, such as fuel shuffling, to reduce the damage rate, and vessel annealing, to restore favorable material properties, that may be practical and used if necessary. 5 figures

  12. Ex vessel steam explosion loads in pressurized water reactors

    International Nuclear Information System (INIS)

    In light water reactor core melt accidents, the molten fuel can be brought into contact with coolant water in the course of the melt relocation in vessel and ex vessel as well as in an accident mitigation action of water addition. The potential risk of explosive molten fuel coolant interactions (FCI, steam explosion) has drawn substantial attention in the safety analysis of reactor severe accidents. The steam explosion intensity is largely dependent upon the degree of volumetric fractions of melt droplets and steam in the fuel coolant mixture. The rate of melt jet breakup and droplet sizes are, therefore, the key physical parameters in the analysis of FCIs. In a recent OECD/NEA international program SERENA, FCI has been studied, in particular, on the status of code capabilities to predict FCI induced dynamic loading of the reactor structures, and identifying area where additional research may be needed to reduce the level of uncertainties in the code predictions. The first phase of SERENA project showed that the codes still cannot calculate all attributes with equal degree of precision. The predicted void fractions in the mixture are generally much higher than the data and are up to the level at which an energetic explosion is not likely, but the codes predict energetic explosion under such highly voided mixture. Currently the SERENA project is in the second phase with experimental as well as analytical work. In this paper, ex vessel steam explosion loads in PWRs calculated by the TRACER II code, a four field numerical model of fuel coolant mixing and explosion propagation, are presented with an emphasis on the jet breakup modeling

  13. Experiences from AE-monitoring during destructive and nondestructive hydrotests of large pressure vessels

    International Nuclear Information System (INIS)

    In Nordic countries, a four-year research programme in the area of elastic-plastic fracture mechanics was initiated in 1985. The main task in this programme is to assess the leak-before-break (LBB) criteria for pressure vessels and piping. The major experimental objective is pressurization until rupture of a large pressure vessel of a length of 16.3 m, a diameter of 2.9 and a wall thickness of 152 mm. Before pressurization an artificial flaw was made in the inner wall of the vessel. The versatile instrumentation used in the test included a multichannel acoustic emission system. AE instrumentation covered the entire vessel. A special sensor group was used in the area of the artificial flaw to monitor it more closely. Data analysis of the AE-system was based on the localization of the AE-events. The results of the AE-test are reported and compared with those for other instrumentation and fracture behaviour. The same real time AE-instrumentation was used in the years 1984 and 1986 to monitor the core areas of the reactor pressure vessels of the nuclear power plants Loviisa I and Loviisa II during the hydrostatic pressure test. The only way to mount the transducers on the area of the pressure vessel was to use the mast of the ultrasonic inspection equipment operating from the outside of the vessel. The hydrotest was performed using a 30 % overpressure. Four minor AE sources were found in Loviisa I while none in Loviisa II. Experiences and results of the AE-test are discussed briefly. (author). 13 figs., 1 ref

  14. 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.)

  15. Fatigue propagation through welded joints of pressure vessels

    International Nuclear Information System (INIS)

    An assessment was made of the behaviour under cyclic load (crack propagation rate and low cycle fatigue) of the metal deposited and of the area affected by the heat of the welded joints assembling the components of the primary circuit. In order to be sure that the results are as representative as possible, the tests were made on metal from joints carried out with the same parameters as actual joints. The studies described here concerned the deposited metal of the welded joints of vessels and the deposited metal in Ni-Cr-Fe alloy of INCONEL 600 type joining the plates at the bottom of steam generators

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Kori Unit 3 Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    This report describes a neutron fluence assessment performed for the Kori Unit 3 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 16 of reactor operation, 2nd Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 3 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 17

  4. Final Report of the 3rd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Kori Unit 3 Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    This report describes a neutron fluence assessment performed for the Kori Unit 3 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 17 of reactor operation, 3rd Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 3 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 18

  5. Final Report of the 3rd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Kori Unit 4 Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    This report describes a neutron fluence assessment performed for the Kori Unit 4 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 18 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 4 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 18

  6. Final Report of the 1st Ex-Vessel Neutron Dosimetry Installation and Evaluations for Ulchin Unit 5 Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    This report describes a neutron fluence assessment performed for the Ulchin Unit 5 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 5 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Ulchin Unit 5 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 5

  7. Final Report of the 3rd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Kori Unit 2 Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    This report describes a neutron fluence assessment performed for the Kori 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 22 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Kori 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 22

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

    International Nuclear Information System (INIS)

    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

  9. Final Report of the 3rd Ex-vessel Neutron Dosimetry Installation And Evaluations for Yonggwang Unit 1 Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    This report describes a neutron fluence assessment performed for the Yonggwang 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. During Cycle 17 of reactor operation, 3rd Ex-Vessel Neutron Dosimetry Program was instituted at Yonggwang 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 17

  10. Final Report of the 4th Ex-Vessel Neutron Dosimetry Installations and Evaluations for Yonggwang Unit 1 Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    This report describes a neutron fluence assessment performed for the Yonggwang 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. During Cycle 18,19 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Yonggwang 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 18,19

  11. Final report for the 1st ex-vessel neutron dosimetry installation and evaluations for Kori unit 4 reactor pressure vessel

    International Nuclear Information System (INIS)

    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

  12. Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Ulchin Unit 1 Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    This report describes a neutron fluence assessment performed for the Ulchin 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. During Cycle 17 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Ulchin 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 1

  13. Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Kori Unit 4 Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    This report describes a neutron fluence assessment performed for the Kori Unit 4 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 17 of reactor operation. an Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 4 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 17

  14. Final Report of the 1st Ex-Vessel Neutron Dosimetry Installation And Evaluations for Ulchin Unit 4 Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    This report describes a neutron fluence assessment performed for the Ulchin Unit 4 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 9 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Ulchin Unit 4 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 9

  15. Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Yonggwang Unit 1 Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    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

  16. Final report of the 1st ex-vessel neutron dosimetry installation and evaluation for Kori unit 1 reactor pressure vessel

    International Nuclear Information System (INIS)

    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. During cycle 22 of reactor operation, an 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 22

  17. Final report of the 1st ex-vessel neutron dosimetry installation and evaluation for Kori unit 3 reactor pressure vessel

    International Nuclear Information System (INIS)

    This report describes a neutron fluence assessment performed for the Kori unit 3 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 3 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

  18. 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.

  19. HFIR Vessel Maximum Permissible Pressures for Operating Period 26 to 50 EFPY (100 MW)

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.; Inger, J.R.

    1999-01-01

    Extending the life of the HFIR pressure vessel from 26 to 50 EFPY (100 MW) requires an updated calculation of the maximum permissible pressure for a range in vessel operating temperatures (40-120 F). The maximum permissible pressure is calculated using the equal-potential method, which takes advantage of knowledge gained from periodic hydrostatic proof tests and uses the test conditions (pressure, temperature, and frequency) as input. The maximum permissible pressure decreases with increasing time between hydro tests but is increased each time a test is conducted. The minimum values that occur just prior to a test either increase or decrease with time, depending on the vessel temperature. The minimum value of these minimums is presently specified as the maximum permissible pressure. For three vessel temperatures of particular interest (80, 88, and 110 F) and a nominal time of 3.0 EFPY(100 MVV)between hydro tests, these pressures are 677, 753, and 850 psi. For the lowest temperature of interest (40 F), the maximum permissible pressure is 295 psi.

  20. HFIR Vessel Maximum Permissible Pressures for Operating Period 26 to 50 EFPY (100 MW)

    International Nuclear Information System (INIS)

    Extending the life of the HFIR pressure vessel from 26 to 50 EFPY (100 MW) requires an updated calculation of the maximum permissible pressure for a range in vessel operating temperatures (40-120 F). The maximum permissible pressure is calculated using the equal-potential method, which takes advantage of knowledge gained from periodic hydrostatic proof tests and uses the test conditions (pressure, temperature, and frequency) as input. The maximum permissible pressure decreases with increasing time between hydro tests but is increased each time a test is conducted. The minimum values that occur just prior to a test either increase or decrease with time, depending on the vessel temperature. The minimum value of these minimums is presently specified as the maximum permissible pressure. For three vessel temperatures of particular interest (80, 88, and 110 F) and a nominal time of 3.0 EFPY(100 MVV)between hydro tests, these pressures are 677, 753, and 850 psi. For the lowest temperature of interest (40 F), the maximum permissible pressure is 295 psi

  1. Evaluation of the integrity of SEP reactor vessels

    International Nuclear Information System (INIS)

    A documented review is presented of the integrity of the 11 reactor pressure vessels covered in the Systematic Evaluation Program. This review deals primarily with the design specifications and quality assurance programs used in the vessel construction and the status of material surveillance programs, pressure-temperature operating limits, and inservice inspection programs of the applicable plants. Several generic items such as PWR overpressurization protection and BWR nozzle and safe-end cracking also are evaluated. The 11 vessels evaluated include Dresden Units 1 and 2, Big Rock Point, Haddam Neck, Yankee Rowe, Oyster Creek, San Onofre 1, LaCrosse, Ginna, Millstone 1, and Palisades

  2. Ultrasonic stress evaluation through thickness of a stainless steel pressure vessel

    International Nuclear Information System (INIS)

    This paper investigates ultrasonic method in stress measurement through thickness of a pressure vessel. Longitudinal critically refracted (LCR) waves are employed to measure the welding residual stresses in a vessel constructed from austenitic stainless steel 304L. The acoustoelastic constant is measured through a hydro test to keep the pressure vessel intact. Hoop and axial residual stresses are evaluated by using different frequency range of ultrasonic transducers. The welding processes of vessel shell and caps are simulated by a 3D finite element (FE) model which is validated by hole-drilling method. The residual stresses calculated by FE simulation are then compared with those obtained from the ultrasonic measurement while a good agreement is observed. It is demonstrated that the residual stresses through thickness of the stainless steel pressure vessel can be evaluated by combining FE and LCR method (known as FELCR method). - Highlights: • The main goal is ultrasonic evaluation of through thickness stresses. • Welding processes of a stainless steel pressure vessel are modelled by FE. • The hole-drilling method is used to validate the FE results. • Residual stresses are measured by four different series of ultrasonic transducers. • The comparison between ultrasonic and FE results show an acceptable agreement

  3. Neutron irradiation effects in reactor pressure vessel steels and weldments. Working document

    International Nuclear Information System (INIS)

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. A separate abstract was prepared for the introduction and for each of the eleven chapters, which are: 1. Reactor Pressure Vessel Design, 2. Reactor Pressure Materials, 3. WWER Pressure Vessels, 4. Determination of Mechanical Properties, 5. Neutron Exposure, 6. Methodology of Irradiation Experiments, 7. Effect of Irradiation on Mechanical Properties, 8. Mechanisms of Irradiation Embrittlement, 9. Modelling of Irradiation Damage, 10. Annealing of Irradiation Damage, 11. Safety Assessment using Surveillance Programmes and Data Bases

  4. Investigation of the safety testing of extraordinary thick steel material for pressure vessels

    International Nuclear Information System (INIS)

    With increasing size of nuclear reactors it was necessary to increase the wall thickness of the reactor pressure vessels. So, for example the wall thickness of the pressure vessel of the LWR pressurized water reactor type with 1.200.000 kW performance is 250 mm. The fabrication of these extraordinary thick steel plates is accurately carried out, nevertheless it is very important to control the characteristics of strength. For this reason extraordinarily thick steel plates of forged manganese-molybdenum-nickel steel produced in Japan and used for the production of reactor pressure vessels was utilized. The aim of this investigation is to know if and how the K sub(IR)- values correspond to the actual regulation for the prevention of brittle fracture and to determine the characteristics. (orig./RW)

  5. Reaction forces due to the decompression of pressurized vessels filled with two-phase fluids

    International Nuclear Information System (INIS)

    A unique fluid behavioral model which is applicable to many homogeneous multiphase fluids is used as a generic model to examine the reaction forces developed by the decompression of pressurized vessels filled with single component two-phase fluids. The decompression is the result of the sudden creation of an outflow area such as by the accidental puncture of the vessel shell or the rupture of a pipe connection at the vessel. The study includes the treatment of two basic initial conditions: the case of a vessel completely filled with a saturated liquid, and the case of a vessel partially filled with a saturated liquid and covered with its saturated vapor, that is, the two region stratified condition. The principal variables include an expansion parameter which characterizes the fluid behavior in the two-phase domain, the vapor volume fraction, and the location of the outflow area for the stratified case

  6. Effects of fatigue in the elastic regime on the mechanical properties of nuclear pressure vessel steels

    International Nuclear Information System (INIS)

    The mechanical properties behavior of one low carbon steel, SAE 1018, and two nuclear pressure vessel steels, A533B Class I and A533B Class II, all with similar compositions, is examined after fatigue in the elastic regime. Cyclic stresses were applied at stress amplitudes of +.207 MPa (+.30 ksi), the design limit for the pressure vessel steels, and well below the upper and lower yield strengths for all of the steels. The fatigue cycling was carried out at room temperature in all cases, and at 3000C for the pressure vessel steels, to 104 and 106 cycles. Specimens which had undergone the fatigue treatment were then tensile and notch tensile tested to determine mechanical properties changes. In addition, electron microscopy techniques were used to characterize the microstructure developed during fatigue. The results show that mechanical properties are affected by elastic fatigue. In general, the materials are more ductile after fatigue than in the as-received condition. The microstructural examination indicates that this change is due to the dislocation substructure which is developed during fatigue. These findings may have a profound impact on predicting behavior of steels in nuclear pressure vessels where alternating stresses are present during operation. To date, the tendency for pressure vessel steel embrittlement under irradiation has been determined from unstressed specimens

  7. A fracture mechanics and reliability based method to assess non-destructive testings for pressure vessels

    International Nuclear Information System (INIS)

    Quantitative evaluation has not been made on the effects of carrying out preservice and in-service nondestructive tests for securing the soundness, safety and maintainability of pressure vessels, spending large expenses and labor. Especially the problems concerning the time and interval of in-service inspections lack the reasonable, quantitative evaluation method. In this paper, the problems of pressure vessels are treated by having developed the analysis method based on reliability technology and probability theory. The growth of surface cracks in pressure vessels was estimated, using the results of previous studies. The effects of nondestructive inspection on the defects in pressure vessels were evaluated, and the influences of many factors, such as plate thickness, stress, the accuracy of inspection and so on, on the effects of inspection, and the method of evaluating the inspections at unequal intervals were investigated. The analysis of reliability taking in-service inspection into consideration, the evaluation of in-service inspection and other affecting factors through the typical examples of analysis, and the review concerning the time of inspection are described. The method of analyzing the reliability of pressure vessels, considering the growth of defects and preservice and in-service nondestructive tests, was able to be systematized so as to be practically usable. (Kako, I.)

  8. Modelling the interaction between corium and pressure vessel steel in the context of melt retention scenarios

    International Nuclear Information System (INIS)

    In case of a reactor accident with core meltdown into the lower plenum, the reactor pressure vessel is the final safety barrier before the containment will be affected. For a detailed investigation of the processes and phenomena involved in this scenario, EU- and OECD-funded experiments were carried out on pressure vessel failure as a result of creep fracture. Models were developed at FZR that comprised both temperature field calculations and the viscoplastic mechanics of the reactor pressure vessel and were capable of identifying the failure mode and failure time of the pressure vessel. The models were validated by precalculations and recalculations of the experiments. A key aspect in the modelling of prototypical LWR scenarios was the consideration of the thermochemical interacation of the corium melt with the pressure vessel wall. In the METCOR experiments carried out by Alexandrov Research Institute (NITI) at Sosnovy Bor, the melt-metal interaction was investigated on a small scale. The results show that steel ablation well start significantly below steel melting temperature as a result of eutectics formation and depending on the composition of the melt. Meltdown models have therefore been integrated in FE models for simulation of melt retention scenarios. The application of these models resulted in significantly reduced residual wall thicknesses. (orig.)

  9. Environmental factor approach to account for water effects in pressure vessel and piping fatigue evaluations

    International Nuclear Information System (INIS)

    This paper summarizes past and current studies of the environmental fatigue effects in light water reactor (LWR) applications. Current Argonne and Japanese research efforts are reviewed and an approach to calculate an environmental correction factor is described. A description of how the proposed approach can be implemented in section III, NB-3600 and NB-3200-type fatigue evaluations is presented along with examples of applying the approach to piping (NB-3600) and safe end fatigue evaluations. These procedures were applied to several BWR and pressurized water reactor (PWR) example cases. The results of these case studies indicated that there is a modest increase in calculated fatigue usage, which is considerably less than the results obtained when the NUREG/CR-5999 curves are applied directly. (orig.)

  10. High-temperature strain measurements on the closure studs of reactor pressure vessels

    International Nuclear Information System (INIS)

    High-temperature strain measurements have been carried out in different operating phases on two closure studs each of several reactor pressure vessels. The strains and stresses during instationary states (e.g. start-up) were of special interest. On the basis of the strains measured, it was to be proved that the calculation methods and boundary conditions on which the analytical investigation of the vessel is based are sufficiently accurate. (orig./RW)

  11. The application of acoustic emission measurements on laboratory testpieces to large scale pressure vessel monitoring

    International Nuclear Information System (INIS)

    A test pressure vessel containing 4 artificial defects was monitored for emission whilst pressure cycling to failure. Testpieces cut from both the failed vessel and from as-rolled plate material were tested in the laboratory. A marked difference in emission characteristics was observed between plate and vessel testpieces. Activity from vessel material was virtually constant after general yield and emission amplitudes were low. Plate testpieces showed maximum activity at general yield and more frequent high amplitude emissions. An attempt has been made to compare the system sensitivities between the pressure vessel test and laboratory tests. In the absence of an absolute calibration device, system sensitivities were estimated using dummy signals generated by the excitation of an emission sensor. The measurements have shown an overall difference in sensitivity between vessel and laboratory tests of approximately 25db. The reduced sensitivity in the vessel test is attributed to a combination of differences in sensors, acoustic couplant, attenuation, and dispersion relative to laboratory tests and the relative significance of these factors is discussed. Signal amplitude analysis of the emissions monitored from laboratory testpieces showed that, whith losses of the order of 25 to 30db, few emissions would be detected from the pressure vessel test. It is concluded that no reliable prediction of acoustic behaviour of a structure may be made from laboratory test unless testpieces of the actual structural material are used. A considerable improvement in detection sensitivity, is also required for reliable detection of defects in low strength ductile materials and an absolute method of system calibration is required between tests

  12. Integrity assessment of TAPS reactor pressure vessel at extended EOL using surveillance test results

    International Nuclear Information System (INIS)

    Integrity assessment of pressure vessels of nuclear reactors (RPV) primarily concentrates on the prevention of brittle failure and conditions are defined under which brittle failure can be excluded. Accordingly, two approaches based on Transition Temperature Concept and Fracture Mechanics Concept were adopted using the impact test results of three credible surveillance data sets obtained from the surveillance specimens of Tarapur Atomic Power Station. RTNDT data towards end of life (EOL) were estimated from the impact test results in accordance with the procedures of USNRC Regulatory Guide 1.99, Rev. 2 and were used as primary input for assessment of the vessel integrity. SA302B (nickel modified) steel cladded with stainless steel is used as the pressure vessel material for the two 210 MWe boiling water reactors of the Tarapur Atomic Power Station (TAPS). The reactors were commissioned during the year 1969. The chemical compositions of SA302B (modified) steel used in fabricating the vessel and the specified tensile property and the Charpy impact property requirements of the steel broadly meet ASME specified requirements. Therefore, the pressure temperature limit curves prescribed by General Electric (G.E.) were compared with those as obtained using procedures of ASME Section XII, Appendix G. The tensile and the Charpy impact properties at 60 EFPY of vessel operation as derived from the surveillance specimens even fulfilled the specified requirements for the virgin material of ASME. Integrity assessment carried out using the two approaches indicated the safety of the vessel for continued operation up to 60 EFPY. (author)

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

    International Nuclear Information System (INIS)

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

  14. Residual stress analysis of autofrettaged thick-walled spherical pressure vessel

    International Nuclear Information System (INIS)

    In this study, residual stress distributions in autofrettaged homogenous spherical pressure vessels subjected to different autofrettage pressures are evaluated. Results are obtained by developing an extension of variable material properties (VMP) method. The modification makes VMP method applicable for analyses of spherical vessels based on actual material behavior both in loading and unloading and considering variable Bauschinger effect. The residual stresses determined by employing finite element method are compared with VMP results and it is demonstrated that the using of simplified material models can cause significant error in estimation of hoop residual stress, especially near the inner surface of the vessel. By performing a parametric study, the optimum autofrettage pressure and corresponding autofrettage percent for creating desirable residual stress state are introduced and determined.

  15. Residual stress analysis of autofrettaged thick-walled spherical pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Maleki, M., E-mail: milad.maleki@epfl.c [School of Mechanical Engineering, Sharif University of Technology, Tehran (Iran, Islamic Republic of); Farrahi, G.H. [School of Mechanical Engineering, Sharif University of Technology, Tehran (Iran, Islamic Republic of); Haghpanah Jahromi, B. [School of Mechanical Engineering, Sharif University of Technology, Tehran (Iran, Islamic Republic of); Department of Mechanical and Industrial Engineering, Northeastern University, Boston (United States); Hosseinian, E. [School of Mechanical Engineering, Sharif University of Technology, Tehran (Iran, Islamic Republic of)

    2010-07-15

    In this study, residual stress distributions in autofrettaged homogenous spherical pressure vessels subjected to different autofrettage pressures are evaluated. Results are obtained by developing an extension of variable material properties (VMP) method. The modification makes VMP method applicable for analyses of spherical vessels based on actual material behavior both in loading and unloading and considering variable Bauschinger effect. The residual stresses determined by employing finite element method are compared with VMP results and it is demonstrated that the using of simplified material models can cause significant error in estimation of hoop residual stress, especially near the inner surface of the vessel. By performing a parametric study, the optimum autofrettage pressure and corresponding autofrettage percent for creating desirable residual stress state are introduced and determined.

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

    International Nuclear Information System (INIS)

    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

  17. A determination of the benefits of annealing irradiated pressure vessel weldments

    International Nuclear Information System (INIS)

    The long-term benefit of annealing an irradiated reactor pressure vessel steel may be described in terms of a benefit factor, B. The benefit factor compares the mechanical properties of an annealed and reirradiated specimen with an equivalent specimen having no intermediate anneal. The benefit factor was determined using a series of microhardness specimens prepared from nuclear pressure vessel surveillance program materials. These specimens were annealed and then reirradiated in a test reactor. There was an obvious long-term benefit in the specimens annealed at 4500C. The long-term benefit was less obvious at 4000C and no significant benefit was noted at 3500C. The benefit factor may also be used as the basis of a surveillance program for an annealed pressure vessel. A strategy for such a surveillance program is described. (author)

  18. The extension and application of a simplified model for PWR type reactor pressure vessel reliability

    International Nuclear Information System (INIS)

    This report deals with two important modifications to a simplified model for the reliability of P.W.R. reactor pressure vessels. Previously the fracture criterion in the model assumed linear elastic behaviour; this has been changed to allow an approximate treatment of elastic-plastic material response. The algorithm has also been changed to include a model for in-service inspection (ultrasonic crack detection). The altered fracture criterion gave failure rates little different from those calculated previously using linear elastic assumptions. A study of in-service inspection has produced new results concerning the performance of P.W.R. pressure vessels. The report includes a discussion of some approximations in the reliability algorithm and how these relate to present physical information. An illustration is given of how the algorithm could be used, to quantify the necessary standards of design, construction and operation. A limited comparison with other pressure vessel reliability algorithms is also presented. (author)

  19. Prediction of thermoplastic failure of a reactor pressure vessel under a postulated core melt accident

    International Nuclear Information System (INIS)

    This paper presents the lower head failure calculations performed for a postulated accident scenario in a commercial nuclear power plant. A postulated one inch break in the primary coolant circuit leads to dryout and subsequent meltdown of the core. The reference plant is a pressurized water reactor without penetrations in the reactor vessel lower head. The molten core material accumulates in the lower head, eventually causing failure of the vessel. The analysis investigates flow conditions in the melt pool, temperature evolution in the reactor vessel wall, and structure mechanical evaluation of the vessel under strong thermal loads and a range of internal pressures. The calculations were performed using the ADINA finite element codes. The analysis focusses on the failure processes, time and mode of failure. The most likely mode of failure at low pressure is global rupture due to gradual accumulation of creep strain over a large part of the heated area. In contrast, thermoplasticity becomes important at high pressure or following a pressure spike and can lead to earlier local failure. In situations in which part of the heat load is concentrated over a small area, resulting in a hot spot, local failure occurs, but not until the temperatures are close to the melting point. At low pressure, in particular, the hot spot area remains intact until the structure is molten across more than half of the thickness. (author) 14 figs., 16 refs

  20. Reactor pressure vessel neutron fluence surveillance dosimetry using solid-state track recorders

    International Nuclear Information System (INIS)

    Many neutron dosimetry issues arose in the context of life assurance/extension programs for nuclear power plants. The most important of these relates to determining the extent of reactor pressure vessel steel embrittlement caused by fast neutron leakage from the reactor core, since a key factor in reactor plant life is the service lifetime of the reactor pressure vessel. In the past, neutron dosimetry for pressure vessel surveillance purposes was carried out mainly through analyses of dosimetry sets installed in surveillance capsules. Surveillance capsule neutron dosimetry programs are limited by restrictions on the position and number of locations available and by the frequency of removal of the surveillance capsules. In many cases, neutron surveillance data cannot be obtained near critical vessel locations, and dosimetry data are only obtained when the surveillance capsules are removed, so most results are averaged over several reactor operating cycles. During the past 5 years, pressure vessel surveillance dosimetry sets were deployed in the reactor cavity (annular gap) region to perform neutron dosimetry on a cycle-by-cycle basis. Solid-state track recorder (SSTR) neutron dosimeters are a key component of these reactor-cavity surveillance dosimetry sets. Measurements by SSTR neutron dosimetry have demonstrably produced high-quality results in a large number of cases, and the absolute accuracy of the technique was validated at the National Institute of Standards and Technology (NIST) through irradiations in standard neutron fields

  1. Device for the determination of the vibrations occurring at the internals of a reactor pressure vessel

    International Nuclear Information System (INIS)

    A device is described for the determination of the vibration occuring at the internals of a reactor pressure vessel by means of a deflection detector connected to the internals. The device consists of: a. a housing and a deflection detector arranged therein; b. a flexible rod attached to the housing; c. a guide tube that extends to the internals through the wall of the vessel which ends in a receptacle adapted to receive the housing and rigidly connected to the internals wherein the end of the guide tube facing away from the pressure vessel is cup-shaped and sealed by means of a cover; and d. the housing being mobile within the guide tube; e. a pipe penetrating the reactor pressure vessel and sealed from the outside having the guide tube connected through the pipe with the end of the guide tube in the pressure vessel facing away from the pipe connected to the receptacle; and f. a cable routed via the flexible rod and through the cover for transmitting the measuring results of the deflection detector to a plug-in-coupler

  2. Design of Semi-composite Pressure Vessel using Fuzzy and FEM

    Science.gov (United States)

    Sabour, Mohammad H.; Foghani, Mohammad F.

    2010-04-01

    The present study attempts to present a new method to design a semi-composite pressure vessel (known as hoop-wrapped composite cylinder) using fuzzy decision making and finite element method. A metal-composite vessel was designed based on ISO criteria and then the weight of the vessel was optimized for various fibers of carbon, glass and Kevlar in the cylindrical vessel. Failure criteria of von-Mises and Hoffman were respectively employed for the steel liner and the composite reinforcement to characterize the yielding/ buckling of the cylindrical pressure vessel. The fuzzy decision maker was used to estimate the thickness of the steel liner and the number of composite layers. The ratio of stresses on the composite fibers and the working pressure as well as the ratio of stresses on the composite fibers and the burst (failure) pressure were assessed. ANSYS nonlinear finite element solver was used to analyze the residual stress in the steel liner induced due to an auto-frettage process. Result of analysis verified that carbon fibers are the most suitable reinforcement to increase strength of cylinder while the weight stayed appreciably low.

  3. The deterministic structural integrity assessment of reactor pressure vessels under pressurized thermal shock loading

    International Nuclear Information System (INIS)

    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 (RTNDT) 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 KIC or fracture arrest toughness KIa) 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 determining the

  4. VISA: a computer code for predicting the probability of reactor pressure-vessel failure

    International Nuclear Information System (INIS)

    The VISA (Vessel Integrity Simulation Analysis) code was developed as part of the NRC staff evaluation of pressurized thermal shock. VISA uses Monte Carlo simulation to evaluate the failure probability of a pressurized water reactor (PWR) pressure vessel subjected to a pressure and thermal transient specified by the user. Linear elastic fracture mechanics are used to model crack initiation and propagation. parameters for initial crack size, copper content, initial RT/sub NDT/, fluence, crack-initiation fracture toughness, and arrest fracture toughness are treated as random variables. This report documents the version of VISA used in the NRC staff report (Policy Issue from J.W. Dircks to NRC Commissioners, Enclosure A: NRC Staff Evaluation of Pressurized Thermal Shock, November 1982, SECY-82-465) and includes a user's guide for the code

  5. Service experience and design in pressure vessels and piping (including high pressure technology). PVP-Volume 335

    Energy Technology Data Exchange (ETDEWEB)

    Bamford, W.H.; Cohn, M.J.; Cipolla, R.C.; Swindeman, R.W.; Nickel, H.; Burns, D.J. [eds.

    1996-12-01

    This volume is divided into the following four sessions: (1) Service Experience in Nuclear Plants; (2) Service Experience in Fossil Plants; (3) High Temperature Structural Materials; and (4) Design and Analysis of High Pressure Vessels. Separate abstracts were prepared for most of the papers in this volume.

  6. Advanced Construction of Compact Containment BWR

    International Nuclear Information System (INIS)

    The reactor concept considered in this paper has a mid/small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. Compact Containment BWR (CCR) is being developed with matured BWR technologies together with innovative systems/components, will provide attractiveness for the energy market in the world due to its flexibility in energy demands as well as in site conditions, its high potential in reducing investment risk and its safety feature facilitating public acceptance. The flexibility is achieved by CCR's mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). The high investment potential is expected from CCR's simplification/innovation in design such as natural circulation core cooling with the bottom located short core, top mounted upper entry control rod drives (CRDs) with ring-type dryers and simplified safety system with high pressure resistible primary containment vessel (PCV) concept. The natural circulation core eliminates recirculation pumps as well as needs for maintenance of such pumps. The top mounted upper entry CRDs enable the bottom located short core in RPV. The safety feature mainly consists of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), high pressure resistible PCV and in-vessel retention (IVR) capability. The large inventory increases the system response time in case of design base accidents including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. CCR's specific self-standing steel high pressure resistible PCV is designed to contain minimum piping and valves inside with reactor pressure vessel (RPV), only 13 m in diameter and 24 m in height. This compact PCV makes it possible to

  7. Effect of stress relief parameters on the mechanical properties of pressure vessel steels and weldments

    International Nuclear Information System (INIS)

    Post weld heat treatments of thick-section A533B steel for nuclear pressure vessels are discussed with reference to the ASME code. The discussion is in the form of a lecture and summarized by noting that the ASME code, in particular Section III, Division 1, imposes a post weld heat treatment requirement on pressure vessels fabricated from low alloy high strength steels. The Code permits a holding temperature range, the high side of which could result in poorer toughness properties. Long times in excess of 100 hours and/or high temperatures, 6490C can result in an increase in the NDT and a decrease in the upper shelf energy

  8. Irradiation induced embrittlement of steels used as reactor pressure vessel and end shields

    International Nuclear Information System (INIS)

    A review of the various factors that influence the irradiation induced changes in mechanical properties of reactor pressure vessel and end shield steel materials have been brought out in this report followed by some typical results on 3 w/o nickel steels and A 302B type of pressure vessel steels. These are the type of steels used in operating reactors in India. In this report the irradiation induced mechanical property changes have been analysed from reported results on nil ductility transition temperature changes (NDTT) from impact tests. The report also brings out the importance of carrying out reactor surveillance programme. (author)

  9. Pressure vessels and piping. V.1.: codes, standards, design and analysis

    International Nuclear Information System (INIS)

    This volume is an effort to share the recent advances in design approaches, current issues in the design of critical equipments. Codes and standards form the backbone of pressure vessels and piping design, which undergo periodic necessary changes based on the design and operating experiences gained in the recent past. Critical examination of design codes, standards and their impending improvements along with material issues have been highlighted in several reviews. Recent advances in theoretical finite elements modelling along with testing and evaluation of critical components have been presented through several original contributions. The papers cover both fundamental as well as application aspects towards safe and economical design of pressure vessels and piping

  10. Solid state track recorder pressure vessel surveillance neutron dosimetry at commercial nuclear power reactors

    International Nuclear Information System (INIS)

    Solid State Track Recorder neutron dosimetry methods developed under the U.S. Nuclear Regulatory Commission Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program have been applied for pressure vessel surveillance dosimetry at commercial nuclear power reactors. More than 800 SSTR neutron dosimeters have been deployed at twelve different power reactors during twenty-two cycles of operation. More than 300 SSTR have been analyzed, and results with uncertainties in the 2-5% range have been generally obtained. Several new areas of application of SSTRs for radiation damage assessment for safe routine operation or extended life operation of power reactors are planned and these applications are discussed. (author). 14 refs

  11. Device for cleaning the lid seal area of a reactor pressure vessel

    International Nuclear Information System (INIS)

    The device is distinguished by a guide rail which can be positioned above the free seal area with the lid removed, whose path corresponds to the seal area and with a trolley guided by the guide rail. The device has the advantage that the lid seal area can be cleaned without operators being involved. The guide rail is circular so that the device can be inserted in the open pressure vessel without any components inside the pressure vessel having to be removed first. (orig./HP)

  12. Investigations on bolts for reactor pressure vessels and connections of core internals

    International Nuclear Information System (INIS)

    As a material for studs with 300 mm diameter, 3,5 %-Ni steels were investigated experimentally with respect to their mechanical and fracture mechanical properties. The strength of the thread of large reactor vessel closure studs was determined by photoelastic tests on nut and bolt joints in tension and compression, by pressure tests on a scaled-down replica of the pressure vessel flange and on washers of AISI 4340, as well as thread stripping and fatigue tests on studs screwed into nuts. The specimens were made of the material normally used for the studs. (orig.)

  13. 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).

  14. Numerical simulation for the experiment on thermal shock to pressure vessel

    International Nuclear Information System (INIS)

    The severe thermal shock generates large tensile stresses in the skin of the pressure vessel. The evaluation of damage potential of thermo-mechanical stresses during thermal shock requires prediction of temperature history of pressure vessel accurately. The present work describes the approach to handle thermal shock problem. The paper deals with the finite element modelling of component for thermal shock problem, analysis procedure, the evaluation of temperature transient history and comparison of computed results with that of published experimental results. (author). 3 refs., 9 figs., 2 tabs

  15. Reactor pressure vessel strength calculations - comparing the AD/TRD and the ASME code. Pt. 1

    International Nuclear Information System (INIS)

    The dimensioning criteria applied in the various technical rules are illustrated by the example of a reactor pressure vessel nozzle, especially with a view to the characteristic data of the materials used. Using a detailed finite element analysis of the main coolant nozzle permits an evaluation of the different calculation methods. The second part of the report discusses safety problems, e.g. fatigue analysis, the necessity of carrying out 3D-elastoplastic FE calculations, or the assessment of transient loads on the reactor pressure vessel by means of a fracture-mechanical analysis. (orig./HP)

  16. Design and Analysis of Filament Wound Composite Pressure Vessel with Integrated-end Domes

    Directory of Open Access Journals (Sweden)

    M. Madhavi

    2009-01-01

    Full Text Available Filament-wound composite pressure vessels are an important type of high-pressure container that is widely used in the commercial and aerospace industries. The pressure vessels with integrated end domes develop hoop stresses that are twice longitudinal stresses and when isotropic materials like metals are used for realizing the hardware, the material is not fully utilized in the longitudinal/meridonial direction resulting in over weight components. On the other hand FRP composite materials with their higher specific strength and moduli and tailoribility characteristics will result in reduction of weight of the structure. The determination of a proper winding angle and thickness is very important to decrease manufacturing difficulties and to increase structural efficiency. In this study, material characterization of FRP of carbon T300/Epoxy for various configurations as per ASTM standards is experimentally determined using filament winding and matched die mould technique. The mechanical and physical properties thus obtained are used in the design of the composite shell. The design of the composite shell is described in detail. Netting analysis is used for the calculation of hoop and helical thickness of the shell. A balanced symmetric ply sequence for carbon T300/epoxy is considered for the entire pressure vessel. Progressive failure analysis of composite pressure vessel with geodesic end domes is carried out. A software code SHELL Solver is developed using Classical Lamination-theory to determine matrix crack failure, burst pressure values at various positions of the shell. The results can be utilized to understand structural characteristics of filament wound pressure vessels with integrated end domes.Defence Science Journal, 2009, 59(1, pp.73-81, DOI:http://dx.doi.org/10.14429/dsj.59.1488

  17. On the statics of supports and the geometry of wrinkling of large spherical pressure vessels

    International Nuclear Information System (INIS)

    Thin-walled spherical pressure vessels, the bending and compressive stiffnesses of which are small in comparison with their tensile stiffness, are discussed using membrane theory. In the first part of the paper linear membrane theory is used to analyze the statics of supports for large spherical pressure vessels. The reactions from such supports which are tangential or almost tangential to the pressure vessel surface, require reinforcements so as to distribute the reactions into the wall without causing undue stress concentrations and/or wrinkling. The size and contour of such reinforcing elements depend, of course, on the magnitude of the reactions as well as the internal pressure. In the second part of the paper, nonlinear membrane theory is used to analyze the geometry of wrinkled domains in such membrane pressure vessels. Using an Eulerian formulation, the parameters of the first and second fundamental forms of the surface are treated as key variables and are determined from the analysis as functions of the curvilinear coordinates referred to the current deformed configuration. The solution technique is applied to a simple example

  18. Analysis of radionuclide behavior in a BWR Mark-II containment under severe accident management condition in low pressure sequence

    International Nuclear Information System (INIS)

    In the Level 2 PSA program at INS/NUPEC, MELCOR1.8.3 is extensively applied to analyze radionuclide behavior of dominant sequences. In addition, the revised source terms provided in the NUREG-1465 report have been also discussed to examine the potential of the radionuclides release to the environment in the conventional siting criteria. In the present study, characteristics of source terms to the environment were examined comparing with results by the Hypothetical Accident (LOCA), NUREG-1465 and MELCOR1.8.3. calculation for a typical BWR with a Mark-II containment in order to assure conservatives of the Hypothetical Accident in Japan. Release fractions of iodine to the environment for the Hypothetical Accident and NUREG-1465, which used engineering models for predicting radionuclide behaviors, were about 10-4 and 10-6 of core inventory, respectively, while the best estimate MELCOR1.8.3 code predicted 10-9 of iodine to the environment. The present study showed that the engineering models in the Hypothetical Accident or NUREG-1465 have large conservatives to estimate source term of iodine to the environment. (author)

  19. 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

  20. BWR 90: The ABB advanced BWR design

    International Nuclear Information System (INIS)

    ABB has two evolutionary advanced fight water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and the total power generation costs have been low. In the development of BWR 90 specific changes were introduced to the reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher dim that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Thus, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The work is scheduled for completion in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with the 'evolutionary' design BWR 90+. The primary design goal is to develop the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is

  1. TNT-blast-equivalence for bursting of pressurized-gas conventional vessels

    International Nuclear Information System (INIS)

    A simple procedure is given for roughly simulating the positive phase of a blast wave arising from the sudden rupture of a conventional pressurized-gas vessel. The procedure differs somewhat from the usual TNT-energy-equivalence method. It is based on equating a bursting-gas-vessel blast to that of a high explosive charge of energy higher than the stored gas energy, and detonating farther away than the vessel considered. In conjunction with blast damage curves or formulas known from military technology, the present method can be used in roughly and quickly assessing the damage hazards that are due to a hypothetical vessel burst, or in designing protective structures. It can also be used in performing model-scale tests on structures that are potentially subject to blast loads. (orig.)

  2. Testing of a steel containment vessel model

    International Nuclear Information System (INIS)

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

  3. Embrittlement and annealing of reactor pressure vessel steels: comparison of BR3 surveillance and vessel plates to the surrogate plates representative of the Yankee Rowe vessel

    International Nuclear Information System (INIS)

    The sister pressure vessels at the BR3 and Rowe Yankee PWR plants were operated at a lower-than-usual temperature (260 degrees Celsius) and their plates were austenitized at higher-than-usual temperature (970 degrees Celsius). A heat tratemement leading to a coarser microstructure than typical for the fine grain plates that are considered in development of USNRC Regulatory guide 1.99. This material displayed outlier behaviour charackterized by a 41J CVN-shift significantly larger than predicted by Regulatory Guide 1.99. Because lower radiation temperature and nickell alloying are generally considered detrimental to irradiation sensitivity, there was a major concern that the nickel-modified lower Rowe plate and the nickel-modified BR3 plate may become too embrittled to satisfy the toughness requirements enbodied in the PTS screening criterion. This paper compares three complementary studies undertaken to clarify these uncertainties: 1) the accelerated irradiation and test program launched in 1990 by Yankee Atomic Electric Company using typical vessel plate materials containing 0.24% copper at two nickel levels: YA1, 0.63% (A533-B) and YA9, 0.19% (A302-B). These were heat-treated to produce the coarse and fine grain microstructures representative of the Yankee/BR3 and the Regulatory Guide plates, respectively, 2) the BR3 surveillance and vessel testing program: this vessel was wet-annealed in 1984, relicensed for operation till the plant shutdown in 1987, ANCL was trepanned in early 1995, 3) the accelerated irradiations in the Belgian BR2 test reactor of the Yankee coarse grain plates YA1 and YA9 together with BR3 vessel specimens extracted at nozzle elevation, a location with negligible radiation exposure. It is shown that the PTS screening criterion was never attained by the BR3 and Rowe plates, and that the BR3 vessel anneal was neither necessary nor sufficient. Finally, the sensitivity of embrittlement, annealing and post-annealing reembrittlement to irradiation

  4. 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.

  5. Fatigue and fracture mechanics in pressure vessels and piping. PVP-Volume 304

    International Nuclear Information System (INIS)

    Fracture mechanics and fatigue evaluations are an important part of the structural integrity analyses to assure safe operation of pressure vessels and piping components during their service life. The paper presented in this volume illustrate the application of fatigue and fracture mechanics techniques to assess the structural integrity of a wide variety of Pressure Vessels and Piping components. The papers are organized in six sections: (1) fatigue and fracture--vessels; (2) fatigue and fracture--piping; (3) fatigue and fracture--material property evaluations; (4) constraint effects in fracture mechanics; (5) probabilistic fracture mechanics analyses; and (6) user's experience with failure assessment diagrams. Separate abstracts were prepared for most of the papers in this book

  6. Stresses in the thread base of several studs of a reactor pressure vessel

    International Nuclear Information System (INIS)

    In the present study the deformation behavior during the pressure test in the thread base of studs (dimensions M21x8) from a reactor pressure vessel was tried to find out experimentally with a minimum of effort of instrumentation. It was intended to check the mathematical assumptions for a detail problem of bolt stresses by means of an experiment. In general the results known from other investigations and from the author's preleminary experiments have been confirmed. (orig./RW)

  7. Parametric study of reinforcement of pressure vessel head with offset nozzle

    International Nuclear Information System (INIS)

    In this research, stress analysis of reinforced nozzle connections in ellipsoidal heads of pressure vessels was carried out using shell theory and the finite element method. Various reinforcement configurations such as integral reinforcement, torus transition and protruding nozzle were considered. A parametric study of the effects of reinforcements on the maximum stresses in the head-nozzle intersections under internal pressure loading was performed. The effects of the geometric parameters of the reinforcements are discussed

  8. 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.

  9. Reactor pressure vessel steels ASME SA533B and SA508 C1.2

    International Nuclear Information System (INIS)

    The report presents the results of the microstructural studies of steels SA533B and SA508 C1.2 obtained in connection with a programme initiated to gather and create information needed for the assessment of structural integrity of the reactor pressure vessels. The as-quenched and variably tempered microstructures were studied with optical, scanning and transmission electron microscopes. (author)

  10. Stress analysis of a column supported hemispherical pressure vessel having a large nozzle

    International Nuclear Information System (INIS)

    Stress analysis of a hemispherical vessel with large nozzle attached to it is studied here when it is subjected to internal pressure and varying temperature across the wall. The stress distribution are plotted to indicate the zones of stress concentrations. It is further shown how the reinforcements near the hemisphere - nozzle intersection reduces the stress concentrations. (orig.)

  11. Remote controlled ultrasonic pre-service and in-service inspections of reactor pressure vessels

    International Nuclear Information System (INIS)

    Remote-controlled ultrasonic pre-service and in-service inspections of reactor pressure vessels are part of a lot of security controls of nuclear power stations. The necessary equipment needed as manipulators, probe-systems, ultrasonic electronic and data recording and evaluating devices is explained in detail. (orig.)

  12. 46 CFR 167.25-1 - Boilers, pressure vessels, piping and appurtenances.

    Science.gov (United States)

    2010-10-01

    ... SCHOOLS PUBLIC NAUTICAL SCHOOL SHIPS Marine Engineering § 167.25-1 Boilers, pressure vessels, piping and... or reconverted to use as nautical school ships on or after July 1, 1951, shall conform with one of... engineering regulations in parts 50 to 63, inclusive, of Subchapter F (Marine Engineering) of this chapter....

  13. Axisymmetric finite element analyses of the KKP-II containment and reactor pressure vessel structures

    International Nuclear Information System (INIS)

    Two refined axisymmetric finite element models were used for the dynamic seismic analyses of the KKP-II Containment and RPV structures, using a postulated ground motion time history. One model was established primarily for the response of the containment structure, whereas the other was used for the response of the reactor pressure vessel plus internals. (Auth.)

  14. Application of probabilistic methods for evaluation of thermal shock scenarios for reactor pressure vessel

    International Nuclear Information System (INIS)

    The paper describes analysis of PSA-1 models. The objectives of the analysis are to identify, group and the frequency of potential scenarios of brittle fracture of the reactor pressure vessel due to thermal shock and cold overpressure using Zaporizhzhya NPP -1 as an example

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

    Science.gov (United States)

    2010-10-01

    ...) Whenever doubt exists as to the safety of an existing boiler, pressure vessel, or piping system, the marine... reboilering work is contracted for. The existing steam piping shall be examined. Those portions which are in... work is contracted for may be continued in service. The steam piping replaced shall be in...

  16. Radiation embrittlement of Sizewell 'B' PWR pressure vessel during a 40 year lifetime

    International Nuclear Information System (INIS)

    Irradiation embrittlement data produced from PWR surveillance and materials test reactor accelerated experiments on low alloy Mn Mo Ni pressure vessel steels, which satisfy the Sizewell 'B' forging materials specification on copper (<= 0.09% wt.%) and nickel (<= 0.85 wt.%), have been examined. (author)

  17. Chemical methods for the use of niobium from pressure vessel cladding as a fast neutron dosimeter

    International Nuclear Information System (INIS)

    the steel samples from the cladding of a pressure vessel of an operating nuclear power reactor were obtained by scraping. The cladding material of the pressure vessel contained about 0.5 % niobium. It was desired to use the niobium as a dosimeter for estimating fast fluences at the pressure vessel. The weak radiation from the reaction product 93mNb cannot be measured in the presence of other elements and interfering activities. A method was developed to separate niobium from other metals present; the concentration and yield of niobium were determined spectrophotometrically. The irradiated niobium was electrodeposited from aqueous solutions on copper discs. The amount of the deposited niobium was determined by a radiochemical method which makes use of its own radioactivity - measured with a liquid scintillation counter - and the known starting mass of niobium. It was possible to determine the deposited niobium masses (5 to 200 microgram) with a desired degree of accuracy. The absolute emission rate of X-rays could then be measured without any self-absorption or interference from other activities. The mass of niobium on each preparate and its X-ray emission rate, later on, were used as basic experimental data for the estimation of last neutron doses at the pressure vessel

  18. Contribution to evaluate the safety of reactor pressure vessels made from steel and prestressed concrete

    International Nuclear Information System (INIS)

    The probabilistic approach was used to quantify certain hypothetical accidents which so far have been classified as hypothetical and therefore have not been considered in the licensing process. The safety features of the different types of reactor pressure vessels are presented in the report. The types of hypothetical failures are described and the corresponding probabilities are discussed

  19. The electrogas and electroslag multipass high speed welding of nuclear pressure vessel steels

    International Nuclear Information System (INIS)

    High-speed electroslag and electrogas welding of 15 Mn Ni63 steel plates to achieve high strength and toughness joints for reactor pressure vessels are described. Mechanical testing of overheating-resistant, brittle fracture resistant low alloy steels is discussed. (UK)

  20. Screwing and holding device for lock nuts, especially for screwed joints of reactor pressure vessels

    International Nuclear Information System (INIS)

    A screwing and holding device for lock nuts of reactor pressure vessels is described which can be remote-controlled and will apply the forces required to unscrew the nuts. In addition, it allows unscrewing, tranporting to and from the place and screwing on again of the nuts within shorter time then all similar devices known until now. (RW)