Program to develop acoustic emission-flaw relationship for inservice monitoring of nuclear pressure vessels. Annual report, July 1, 1976 - October 1, 1977

Energy Technology Data Exchange (ETDEWEB)

Hutton, P.H.; Kurtz, R.J.; Schwenk, E.B.; Pavloff, C.

1978-06-01

Laboratory mechanical tests were conducted to evaluate AE during uniaxial tensile, fracture and fatigue crack growth in A533B pressure vessel steel. The A533B steel included two heats of Class 1, one heat of Class 2 and a weldment made for the Heavy Section Steel Technology (HSST) Program. Specimen types included uniaxial tensile specimens, size 2 compact tension specimens for fatigue crack growth and fracture tests, and a single-edge notch specimen also for fatigue crack growth through material that was uniformly strained 3% prior to fatigue testing. In addition, AE monitoring was conducted on the HSST V-7B 6-inch thick pressure vessel test. AE data were partitioned into four ranges of signal amplitude and rise time. All the AE data were analyzed, with respect to mechanical behavior of A533B steel. Linear elastic fracture mechanics analysis methods were used to relate AE parameters to fracture and fatigue crack growth parameters. AE data from the V-7B vessel test were correlated with stress intensity factor and crack opening displacement. AE data from the fatigue crack growth tests were investigated using models based on fatigue crack growth rate, fatigue crack area and theoretical crack tip plastic zone size.

Program to develop acoustic emission-flaw relationship for inservice monitoring of nuclear pressure vessels. Annual report, July 1, 1976--October 1, 1977

International Nuclear Information System (INIS)

Hutton, P.H.; Kurtz, R.J.; Schwenk, E.B.; Pavloff, C.

1978-03-01

Laboratory mechanical tests were conducted to evaluate AE during uniaxial tensile, fracture and fatigue crack growth in A533B pressure vessel steel. The A533B steel included two heats of Class 1, one heat of Class 2 and a weldment made for the Heavy Section Steel Technology (HSST) Program. Specimen types included uniaxial tensile specimens, size 2 compact tension specimens for fatigue crack growth and fracture tests, and a single-edge notch specimen also for fatigue crack growth through material that was uniformly strained 3% prior to fatigue testing. In addition, AE monitoring was conducted on the HSST V-7B 6-inch thick pressure vessel test. AE data were partitioned into four ranges of signal amplitude and rise time. All the AE data were analyzed, with respect to mechanical behavior of A533B steel. Linear elastic fracture mechanics analysis methods were used to relate AE parameters to fracture and fatigue crack growth parameters. AE data from the V-7B vessel test were correlated with stress intensity factor and crack opening displacement. AE data from the fatigue crack growth tests were investigated using models based on fatigue crack growth rate, fatigue crack area and theoretical crack tip plastic zone size

Study of radiation damage of steels for light water pressure vessels at UJV

International Nuclear Information System (INIS)

Vacek, N.; Stoces, B.

1980-01-01

Preoperational determination of radiation resistance of pressure vessel steels is performed at accelerated neutron exposure in a test or materials research reactor. The results obtained at accelerated and operating exposure are not fully identical and surveillance bodies are therefore used manufactured from the pressure vessel material. Currently, the following steels are used for the manufacture of light water reactor pressure vessels: Mn-Mo-Ni (ASTM-A533-B, ASTM-A508), Cr-Mo-V (15Kh2M1FA). At UJV Rez, for irradiation Chanca-M probes imported from France are used featuring electric temperature control. Almost identical radiation embrittlement was measured for all three steels after irradiation with a neutron fluence of 3x10 23 n.m -2 at a temperature of 290 degC. (H.S.)

Application of electron beam welding to large size pressure vessels made of thick low alloy steel

International Nuclear Information System (INIS)

Kuri, S.; Yamamoto, M.; Aoki, S.; Kimura, M.; Nayama, M.; Takano, G.

1993-01-01

The authors describe the results of studies for application of the electron beam welding to the large size pressure vessels made of thick low alloy steel (ASME A533 Gr.B cl.2 and A533 Gr.A cl.1). Two major problems for applying the EBW, the poor toughness of weld metal and the equipment to weld huge pressure vessels are focused on. For the first problem, the effects of Ni content of weld metal, welding conditions and post weld heat treatment are investigated. For the second problem, an applicability of the local vacuum EBW to a large size pressure vessel made of thick plate is qualified by the construction of a 120 mm thick, 2350 mm outside diameter cylindrical model. The model was electron beam welded using local vacuum chamber and the performance of the weld joint is investigated. Based on these results, the electron beam welding has been applied to the production of a steam generator for a PWR. (author). 3 refs., 10 figs., 4 tabs

H.B. Robinson-2 pressure vessel benchmark

Energy Technology Data Exchange (ETDEWEB)

Remec, I.; Kam, F.B.K.

1998-02-01

The H. B. Robinson Unit 2 Pressure Vessel Benchmark (HBR-2 benchmark) is described and analyzed in this report. Analysis of the HBR-2 benchmark can be used as partial fulfillment of the requirements for the qualification of the methodology for calculating neutron fluence in pressure vessels, as required by the U.S. Nuclear Regulatory Commission Regulatory Guide DG-1053, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. Section 1 of this report describes the HBR-2 benchmark and provides all the dimensions, material compositions, and neutron source data necessary for the analysis. The measured quantities, to be compared with the calculated values, are the specific activities at the end of fuel cycle 9. The characteristic feature of the HBR-2 benchmark is that it provides measurements on both sides of the pressure vessel: in the surveillance capsule attached to the thermal shield and in the reactor cavity. In section 2, the analysis of the HBR-2 benchmark is described. Calculations with the computer code DORT, based on the discrete-ordinates method, were performed with three multigroup libraries based on ENDF/B-VI: BUGLE-93, SAILOR-95 and BUGLE-96. The average ratio of the calculated-to-measured specific activities (C/M) for the six dosimeters in the surveillance capsule was 0.90 {+-} 0.04 for all three libraries. The average C/Ms for the cavity dosimeters (without neptunium dosimeter) were 0.89 {+-} 0.10, 0.91 {+-} 0.10, and 0.90 {+-} 0.09 for the BUGLE-93, SAILOR-95 and BUGLE-96 libraries, respectively. It is expected that the agreement of the calculations with the measurements, similar to the agreement obtained in this research, should typically be observed when the discrete-ordinates method and ENDF/B-VI libraries are used for the HBR-2 benchmark analysis.

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

International Nuclear Information System (INIS)

Canonico, D.A.; Stelzman, W.J.

1976-01-01

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, 649 0 C can result in an increase in the NDT and a decrease in the upper shelf energy

Radiation embrittlement of WWER 440 pressure vessel steel and of some improved steels by western producers

International Nuclear Information System (INIS)

Koutsky, J.; Vacek, M.; Stoces, B.; Pav, T.; Otruba, J.; Novosad, P.; Brumovsky, M.

1982-01-01

The resistance was studied of Cr-Mo-V type steel 15Kh2MFA to radiation embrittlement at an irradiation temperature of around 288 degC. Studied was the steel used for the manufacture of the pressure vessel of the Paks nuclear reactor in Hungary. The obtained results of radiation embrittlement and hardening of steel 15Kh2MFA were compared with similar values of Mn-Ni-Mo type steels A 533-B and A 508 manufactured by leading western manufacturers within the international research programme coordinated by the IAEA. It was found that the resistance of steel 15Kh2MFA to radiation embrittlement is comparable with steels A 533-B and A 508 by western manufacturers. (author)

J-R Fracture Resistance of SA533 Gr.B-Cl.1 Steel for Reactor Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Yoon, Ji-Hyun; Hong, Seokmin; Lee, Bong-Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

A rolled plate might show different mechanical behaviors from a forging, even though they contain same chemical compositions. Furthermore, it is known that the fracture behavior of a rolled plate is very sensitive to material orientation comparing to a forging. In this study, the J-R fracture resistances of SA533 Gr.B-Cl.1 plate were measured at reactor operating temperature and the material orientation sensitivity was discussed. The decrease of fracture resistance of this kind of low alloy steel at an elevated temperature is known as the effect of dynamic strain aging (DSA). It was attributed to that the carbides and grains elongated to primary rolling direction, so that the aspect ratio of carbides and grains in the specimen with T-L orientation is larger. Generally, the hard second phase could take a roll of trigger point of unstable fracture. It is needed that the fracture surfaces of the tested specimens to be examined profoundly.

Study of Irradiation Effects on the Fracture Properties of A533-Series Ferritic Steels

International Nuclear Information System (INIS)

Lee, Yong Bok; Lee, Gyeong Geun; Kwon, Jun Hyun

2011-01-01

Since the Kori nuclear power plant unit 3 (Kori-3) was founded in 1986, the surveillance tests have been conducted five times. One of the primary objectives of the surveillance test is to determine the effects of irradiation on reactor pressure vessel (RPV) steel embrittlement. The RPV is made out of ferritic steels such as SA533 type B class 1, which were used for early nuclear power plants industry including Kori-2, 3, 4 and Yonggwang-1, 2 units in Korea. The Westinghouse supplied Kori-3 with the RPV steels ASTM A533 grade B class 1, which is equivalent to SA533 type B class 1. The irradiation effects on tensile properties in ASTM A533 grade B class 1 steel had been studied by Steichen and Williams. They experimentally determined the effect of strain rate and temperature on the tensile properties of unirradiated and irradiated A533 grade B steel 1. The effects of neutron irradiation on ferritic steels could be determined from tensile properties, as well as the fracture strength and toughness measurements. Hunter and Williams have reported that the strength and ductility for unirradiated material at a low strain rate increase with decreasing test temperature. Also, neutron irradiation increases strength and decreases ductility. Crosley and Ripling revealed that the yield strength of unirradiated material rapidly increases with the strain rate. Therefore, yield strength for unirradiated and irradiated materials should be determined by test parameters along with strain rate and temperature. In this study we compare ASTM A533 grad B class 1 steel obtained from several papers with SA533 type B class 1 steel taken from the surveillance data of Kori-3 unit, whose mechanical property of unirradiated and irradiated materials was correlated with the rate-temperature parameter

Fabrication history and mechanical properties for ASTM A-533 Grade-B Class-2 steel weld for fully welded nuclear pressure vessels

International Nuclear Information System (INIS)

Pachur, D.

1979-01-01

Till now pressure vessels for light water reactors were made from rolled plates and forgings connected with each other by welding. The optimal quality of plates and forgings are limited in principle by the foundry technology. It is well known that in this process decomposition and segregation zones occur. Besides the heat affected zone created by the welding process is a weak link. The heat affected zone is heterogeneous and can be harbinger of risks leading to cracks. The production of a pressure vessel through shape welding is an alternative. The cylindrical container is produced by the application of one layer of welding after the other in a preshaped form. During the welding process the previously applied layers are simultaneously being tempered. The undesirable chemical residual elements are evenly distributed and segregation zones do not occur. Since we have only welding material the disadvantages of a heat affected zone are avoided. Furthermore the mechanical properties are independent of location and orientation. This shape welding process proved to be highly economical already during the experimental stay. Besides this process is applicable for vessel of any desired dimension

Thermal Aging Effect on Corrosion Resistance in Fusion Boundary of A533 Gr. B and Alloy 152

Energy Technology Data Exchange (ETDEWEB)

Choi, Kyoung Joon; Yoo, Seung Chang; Kim, Taeho; Ham, Junhyuk; Kim, Ji Hyun [UNIST, Ulsan (Korea, Republic of)

2016-10-15

Dissimilar metal weldment (DMW) is frequently used for joining low-alloy steel pressure vessel nozzles and steam generator nozzles to nickel-based wrought alloy or austenitic stainless steel components in high energy systems. This feature also significantly hinders C diffusion from the ferrite base metal to the weld metal. Until now, stress corrosion cracking has not occurred in DMWs where a High-Cr weld metal (such as Alloy 152 or Alloy 690), which is Ni-base weld metal including relative high Cr, is used as the weld metal in the weld between the nickel-based alloy and low-alloy steel. To understand the microstructure and corrosion evolution on fusion boundary between low-alloy steel and Ni-base weld metal, microstructural analysis and polarization test were performed with A533 Gr. B/Alloy 152/Alloy 690. Remarkable changes were observed in corrosion resistance and hardness at fusion boundary between low-alloy steel and Ni-base weld metal. The precipitate, which has different potential with peripheral region, can cause galvanic corrosion or pitting corrosion and is the one of hardening methods by disturbing movement of the dislocation. At initial step of heat treatment, the number of precipitates was increased. In fusion boundary between A533 Gr. B and Alloy 152, the corrosion resistance was decreased, and the hardness was increased. Next, at further step, the number of precipitates.

Investigation of irradiation embrittlement and annealing behaviour of JRQ pressure vessel steel by instrumented impact tests

Energy Technology Data Exchange (ETDEWEB)

Valo, M; Rintamaa, R; Nevalainen, M; Wallin, K; Torronen, K [Technical Research Centre of Finland, Espoo (Finland); Tipping, P [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1994-12-31

Seven series of A533-B type pressure vessel steel specimens irradiated as well as irradiated - annealed - re-irradiated to different fast neutron fluences (up to 5.10{sup 19}/cm{sup 2}) have been tested with a new type of instrumented impact test machine. The radiation embrittlement and the effect of the intermediate annealing was assessed by using the ductile and cleavage fracture initiation toughness. Although the ductile fracture initiation toughness exhibited scatter, the transition temperature shift corresponding to the dynamic cleavage fracture initiation agreed well with the 41 J Charpy-V shift. The results indicate that annealing is beneficial in restoring mechanical properties in an irradiated nuclear pressure vessel steel. (authors). 8 refs., 11 figs., 1 tab.

The behavior of shallow flaws in reactor pressure vessels

International Nuclear Information System (INIS)

Rolfe, S.T.

1991-11-01

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

Microstructural evolution in reactor pressure vessel steel under neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Ohno, Katsumi; Fukuya, Koji [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

2000-09-01

Understanding microstructural changes in reactor pressure vessel steels is important in order to evaluate radiation-induced embrittlement, one of the major aging phenomena affecting the extension of plant life. In this study, actual surveillance test specimens and samples of rector vessel low-alloy steel (A533B steel) irradiated in a research reactor were examined using state-of-the-art techniques to clarify the neutron flux effect on the microstructural changes. These techniques included small angle neutron scattering and atom probes. Microstructural changes which are considered to be the main factors affecting embrittlement, including the production of copper-rich precipitates and the segregation of impurity elements, were confirmed by the results of the study. In addition, the mechanical properties were predicted based on the obtained quantitative data such as the diameters of precipitates. Consequently, the hardening due to irradiation was almost simulated. (author)

Fracture toughness determination of the pressure vessel steel A508 Cl 2 between 100 and 350 degree C

International Nuclear Information System (INIS)

Rao, S.

1980-09-01

The fracture toughness of the pressure vessel steel A508 was determined in the temperature range 100 - 350 degree C. The J-integral method with crack growth resistance curves, the so-called R-curves, was used. The results show that the steel does not have an 'upper-shelf' and the fracture toughness, K sub (JC), decreases with increasing temperature to a minimum around 300 degree C and an increase above it. These results are compared to those obtained previously on an other pressure vessel steel A533B which has essentially the same temperature dependence. The results were also analysed using the Tearing modulus, T. The conclusion iw that the crack growth resistance and the crack initiation resistance (K sub (JC)) show a significant decrease around the operating temperatures as compared to 100 degree C. (author)

Changes of the mechanical properties of ASTM A 533 type B class 1 (JRQ) steel used in pressure vessels of nuclear power plants

International Nuclear Information System (INIS)

Balderrama, Juan J.; Iorio, Antonio F.

1999-01-01

The steels used in pressure vessels generally present a non-homogenous microstructure across the thickness of their walls due to their manufacturing process. Average thickness being between 200-250 mm also makes the problem more serious. These facts lead us to think that the variation affects not only microstructure, but also mechanical properties. For this reason the methodology for the evaluation of materials should be standardized for their use before and after radiation by means of a surveillance program which allows us to verify the conditions of the steel of the pressure vessel by using Charpy-v, tensile and fracto-mechanics specimens inside the reactor to obtain information about the condition of the pressure vessel material. In order to analyze these changes, tests were carried out using Charpy-v specimens with different orientation inside the block representing the wall thickness and the corresponding ductile-to-brittle transition curves were made for each direction. The orientations to be considered will be four in all and will be those called TL, LT, ST and LS by ASTM E 399 (1993). The conclusions reached arise from a comparative analysis of the results obtained for each orientation under study and confirm the recommendation by Standards regarding the selection of the TL orientation as the most conservative. (author)

Temperature dependence of the fracture toughness and the cleavage fracture strength of a pressure vessel steel

International Nuclear Information System (INIS)

Kotilainen, H.

1980-01-01

A new model for the temperature dependence of the fracture toughness has been sought. It is based on the yielding processes at the crack tip, which are thought to be competitive with fracture. Using this method a good correlation between measured and calculated values of fracture toughness has been found for a Cr-Mo-V pressure vessel steel as well as for A533B. It has been thought that the application of this method can reduce the number of surveillance specimens in nuclear reactors. A method for the determination of the cleavage fracture strength has been proposed. 28 refs

ZZ SAIL, Albedo Scattering Data Library for 3-D Monte-Carlo Radiation Transport in LWR Pressure Vessel

International Nuclear Information System (INIS)

1982-01-01

1 - Description of problem or function: Format: SAIL format; Number of groups: 23 neutron / 17 gamma-ray; Nuclides: Type 04 Concrete and Low Carbon Steel (A533B). Origin: Science Applications, Inc (SAI); Weighting spectrum: yes. SAIL is a library of albedo scattering data to be used in three-dimensional Monte Carlo codes to solve radiation transport problems specific to the reactor pressure vessel cavity region of a LWR. The library contains data for Type 04 Concrete and Low Carbon Steel (A533B). 2 - Method of solution: The calculation of the albedo data was perform- ed with a version of the discrete ordinates transport code DOT which treats the transport of neutrons, secondary gamma-rays and gamma- rays in one dimension, while maintaining the complete two-dimension- al treatment of the angular dependence

Slow strain rate stress corrosion tests on A508-III and A533B steel in de-ionized and PWR water at 563K

International Nuclear Information System (INIS)

Hurst, P.; Appleton, D.A.; Banks, P.; Raffel, A.S.

1985-01-01

An experimental programme is being undertaken to assess the extent to which PWR pressure vessel steels, including weldments, may be susceptible to stress corrosion cracking under relevant water chemistry and flow rate conditions. Initial results from slow strain rate tests on parent A533B and A508-III steels together with a weldment are described. No susceptibility to stress corrosion was observed for either steel, when tested in the rolling (L) direction, at a potential characteristic of normal quality PWR water. For cracking to occur on such specimens the potential must be displaced by 400 to 500 mV in the positive direction, requiring (in the case of high flow de-ionized water) the presence of ca 200 ppb oxygen. Some cracking was observed on transverse (S) direction specimens in water containing 2 , indicating that there may be micro-structural features which can cause cracking even in low oxygen water. This orientation is not directly relevant to pressure vessels and the cracking may only arise as a consequence of the deformation encountered in the slow strain rate test. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-01

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

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

International Nuclear Information System (INIS)

Bass, B.R.; Williams, P.T.; McAfee, W.J.; Pugh, C.E.

2001-01-01

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

Program to develop acoustic emission: flaw relationship for inservice monitoring of nuclear pressure vessels. Progress report No. 1, July 1, 1976--February 1, 1977

International Nuclear Information System (INIS)

Hutton, P.H.; Schwenk, E.B.

1977-03-01

This is a laboratory research program to characterize acoustic emission (AE) from flaw growth and noise from innocuous sources in A533B Class 1 pressure vessel steel. The objectives are: characterize AE from a limited range of defects and material property conditions of concern to reactor pressure vessel integrity; characterize AE from innocuous sources (including defects); develop criteria for distinguishing significant flaws from innocuous sources; and develop an AE flaw damage model to serve as a basis for relating in-service AE to pressure vessel integrity. The purpose of the program is to build an experimental evaluation of the feasibility of detecting and analyzing flaw growth in reactor pressure boundaries by continuously monitoring for AE. A detailed program plan in the form of an analysis-before-test document has been prepared and approved

Special enclosure for a pressure vessel

International Nuclear Information System (INIS)

Wedellsborg, B.W.; Wedellsborg, U.W.

1993-01-01

A pressure vessel enclosure is described comprising a primary pressure vessel, a first pressure vessel containment assembly adapted to enclose said primary pressure vessel and be spaced apart therefrom, a first upper pressure vessel jacket adapted to enclose the upper half of said first pressure vessel containment assembly and be spaced apart therefrom, said upper pressure vessel jacket having an upper rim and a lower rim, each of said rims connected in a slidable relationship to the outer surface of said first pressure vessel containment assembly, mean for connecting in a sealable relationship said upper rim of said first upper pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, means for connecting in a sealable relationship said lower rim of said first upper pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, a first lower pressure vessel jacket adapted to enclose the lower half of said first pressure vessel containment assembly and be spaced apart therefrom, said lower pressure vessel jacket having an upper rim connected in a slidable relationship to the outer surface of said first pressure vessel containment assembly, and means for connecting in a sealable relationship said upper rim of said first lower pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, a second upper pressure vessel jacket adapted to enclose said first upper pressure vessel jacket and be spaced apart therefrom, said second upper pressure vessel jacket having an upper rim and a lower rim, each of said rims adapted to slidably engage the outer surface of said first upper pressure vessel jacket, means for sealing said rims, a second lower pressure vessel jacket adapted to enclose said first lower pressure vessel jacket and be spaced apart therefrom

14 CFR 25.533 - Hull and main float bottom pressures.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Hull and main float bottom pressures. 25... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Structure Water Loads § 25.533 Hull and main float bottom pressures. (a) General. The hull and main float structure, including frames and bulkheads...

Strain ageing in welds of nuclear pressure vessels

International Nuclear Information System (INIS)

Otterberg, R.; Karlsson, C.

1979-01-01

Static and dynamic strain ageing have been investigated on submerged-arc welds and repair welds from plates of the pressure vessel steel A 533B. The results permit the determination of the worst strain ageing conditions existing in a nuclear pressure vessel. Static strain ageing was investigated by means of data from tension tests, hardness measurements and Charpy-V impact properties for prestrained and aged material for ageing temperatures from room temperature to 350 deg C and ageing times up to 1000h. Dynamic strain ageing was investigated by tensile tests up to 350 deg C at different strain rates. At the most static strain ageing was found to increase the impact transition temperature from -75 deg C in the as-received condition to -55 deg C after prestraining and ageing for the plate material, from -35 to -10 deg C for the submerged arc weld and from -90 to -40 deg C for the repair weld. Approximately 10 deg C of the deleterious effect is due to the effect of ageing for the two former materials whereas the corresponding figure for the repair weld amounts to 35 deg C. The dynamic strain ageing is strongest at very low strain rates at temperatures just below 300 deg C. The effect of strain ageing can be reduced by stress relief heat treatment or by other means decreasing the content of nitrogen in solution. (author)

Apparent embrittlement saturation and radiation mechanisms of reactor pressure vessel steels

International Nuclear Information System (INIS)

Pachur, D.

1981-01-01

The irradiation and annealing results of three different reactor pressure vessel steels are reported. Steel A, a basic material according to ASTM A-533 B having 0.15 percent vanadium; and Steel C contained 3.2 percent nickel. The steels were irradiated at 150, 300, and 400 degree C with neutron fluxes of 6 multiplied by 10 11 and 3 multiplied by 10 13 neutrons (n)/cm 2 /s. An apparent saturation-in-irradiation effect was found within certain neutron fluence ranges. During the annealing, various recovery processes occur in different temperature ranges. These are characterized by various activation energies. The individual processes were determined by the different time dependencies at various temperatures. Two causes for the apparent saturation were discovered from the behavior of the annealing curves

Multiple shell pressure vessel

International Nuclear Information System (INIS)

Wedellsborg, B.W.

1988-01-01

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

Reactor Pressure Vessel (RPV) Acquisition Strategy

Energy Technology Data Exchange (ETDEWEB)

Mizia, Ronald Eugene [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2008-04-01

The Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, TRISO-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. The purpose of this report is to address the acquisition strategy for the NGNP Reactor Pressure Vessel (RPV). This component will be larger than any nuclear reactor pressure vessel presently in service in the United States. The RPV will be taller, larger in diameter, thicker walled, heavier and most likely fabricated at the Idaho National Laboratory (INL) site of multiple subcomponent pieces. The pressure vessel steel can either be a conventional materials already used in the nuclear industry such as listed within ASME A508/A533 specifications or it will be fabricated from newer pressure vessel materials never before used for a nuclear reactor in the US. Each of these characteristics will present a

Validated automated ultrasonic inspections of the Sizewell 'B' reactor pressure vessel

International Nuclear Information System (INIS)

Dikstra, B.J.; Farley, J.M.

1992-01-01

Automated ultrasonic inspection was applied extensively during manufacture of the RPV for Sizewell 'B'. This was an important element of the safety case presented at the Sizewell 'B' public enquiry. This requirement reflected concern in the United Kingdom as to the effectiveness and reliability of ultrasonic inspections. By applying automated inspections in addition to the manual ultrasonic inspection carried out by the vessel manufacturer, the overall reliability of the inspection of the vessel would be considerably enhanced. The automated inspections carried out in the manufacturer's workshops were termed 'automated shop inspections' (ASIs). The ASIs were carried out in two contracts: the first to inspect the component forgings of the RPV, the second to inspect the pressure retaining welds. (author)

14 CFR 23.533 - Hull and main float bottom pressures.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Hull and main float bottom pressures. 23... Water Loads § 23.533 Hull and main float bottom pressures. (a) General. The hull and main float....00213; K2=hull station weighing factor, in accordance with figure 2 of appendix I of this part; VS1...

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

C.O.D. toughness testing of medium strength steel as a preliminary development for single specimen J integral toughness tests of SA533-B steel

International Nuclear Information System (INIS)

Dean, P.; Tait, R.B.; Garrett, G.G.

1981-10-01

The primary purpose of this project is to set up a test facility and to develop the necessary expertise to enable reliable elasto-plastic fracture toughness tests to be performed. Initially, tests are to be conducted on material similar to that used in the Koeberg pressure vessel walls, with the ultimate goal of performing single specimen J integral tests on the pressure vessel steel itself to determine through-thickness toughness variations. The project will comprise a number of stages, each one necessary for the development of the techniques used in J integral testing. These include: (i) development of an appropriate specimen design, of suitable size and shape that is applicable to both crack opening displacement (C.O.D.) and J integral tests; (ii) development, testing and calibration of the necessary associated mechanical and electrical equipment (e.g. clip gauge, amplifiers, interface unit, etc.), with (iii) an estimation of the probable errors and noise levels with a view to their elimantion, leading to (iv) perfection of the sensitivity and reproducibility of, firstly, the multiple specimen C.O.D. technique and, secondly, the multiple specimen J integral techniques. (v) Based on the above techniques, development of the single specimen J integral test method incorporating development of a computerised testing procedure. All the above procedure is to be conducted on similar, but non-Koeberg pressure vessel material ('ROQ Tough'). (vi) Finally, development and testing of both multiple specimen and single specimen J integral tests on actual SA533B material and an investigation of the through thickness toughness and fatigue crack propagation behaviour

Pressure vessel for nuclear reactor plant consisting of several pre-stressed cast pressure vessels

International Nuclear Information System (INIS)

Bodmann, E.

1984-01-01

Several cylindrical pressure vessel components made of pressure castings are arranged on a sector of a circle around the cylindrical cast pressure vessel for accommodating the helium cooled HTR. Each component pressure vessel is connected to the reactor vessel by a horizontal gas duct. The contact surfaces between reactor and component pressure vessel are in one plane. In the spaces between the individual component pressure vessels, there are supporting blocks made of cast iron, which are hollow and also have flat surfaces. With the reactor vessel and the component pressure vessels they form a disc-shaped connecting part below and above the gas ducts. (orig./PW)

Metallurgy of steels for PWR pressure vessels

International Nuclear Information System (INIS)

Kepka, M.; Mocek, J.; Barackova, L.

1980-01-01

A survey and the chemical compositions are presented of reactor pressure vessel steels. The metallurgy is described of steel making for pressure vessels in Japan and the USSR. Both acidic and alkaline open-hearth steel is used for the manufacture of ingots. The leading world manufacturers of forging ingots for pressure vessels, however, exclusively use electric steel. Vacuum casting techniques are exclusively used. Experience is shown gained with the introduction of the manufacture of forging ingots for pressure vessels at SKODA, Plzen. The metallurgical procedure was tested utilizing alkaline open hearths, electric arc furnaces and facilities for vacuum casting of steel. Pure charge raw materials should be used for securing high steel purity. Prior to forging pressure vessel rings, not only should sufficiently big bottoms and heads be removed but also the ingot middle part should be scrapped showing higher contents of impurities and nonhomogeneous structure. (B.S.)

Critical experiments, measurements, and analyses to establish a crack arrest methodology for nuclear pressure vessel steels

International Nuclear Information System (INIS)

Hahn, G.T.

1977-01-01

Substantial progress was made in three important areas: crack propagation and arrest theory, two-dimensional dynamic crack propagation analyses, and a laboratory test method for the material property data base. The major findings were as follows: Measurements of run-arrest events lent support to the dynamic, energy conservation theory of crack arrest. A two-dimensional, dynamic, finite-difference analysis, including inertia forces and thermal gradients, was developed. The analysis was successfully applied to run-arrest events in DCB (double-cantilever-beam) and SEN (single-edge notched) test pieces. A simplified procedure for measuring K/sub D/ and K/sub Im/ values with ordinary and duplex DCB specimens was demonstrated. The procedure employs a dynamic analysis of the crack length at arrest and requires no special instrumentation. The new method was applied to ''duplex'' specimens to measure the large K/sub D/ values displayed by A533B steel above the nil-ductility temperature. K/sub D/ crack velocity curves and K/sub Im/ values of two heats of A533B steel and the corresponding values for the plane strain fracture toughness associated with static initiation (K/sub Ic/), dynamic initiation (K/sub Id/), and the static stress intensity at crack arrest (K/sub Ia/) were measured. Possible relations among these toughness indices are identified. During the past year the principal investigators of the participating groups reached agreement on a crack arrest theory appropriate for the pressure vessel problem. 7 figures

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 ...

Dynamic fracture toughness of ASME SA508 Class 2a ASME SA533 grade A Class 2 base and heat affected zone material and applicable weld metals

International Nuclear Information System (INIS)

Logsdon, W.A.; Begley, J.A.; Gottshall, C.L.

1978-03-01

The ASME Boiler and Pressure Vessel Code, Section III, Article G-2000, requires that dynamic fracture toughness data be developed for materials with specified minimum yield strengths greater than 50 ksi to provide verification and utilization of the ASME specified minimum reference toughness K/sub IR/ curve. In order to qualify ASME SA508 Class 2a and ASME SA533 Grade A Class 2 pressure vessel steels (minimum yield strengths equal 65 kip/in. 2 and 70 kip/in. 2 , respectively) per this requirement, dynamic fracture toughness tests were performed on these materials. All dynamic fracture toughness values of SA508 Class 2a base and HAZ material, SA533 Grade A Class 2 base and HAZ material, and applicable weld metals exceeded the ASME specified minimum reference toughness K/sub IR/ curve

46 CFR 176.812 - Pressure vessels and boilers.

Science.gov (United States)

2010-10-01

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

Stress corrosion cracking studies on ferritic low alloy pressure vessel steel - water chemistry and modelling aspects

International Nuclear Information System (INIS)

Tipping, P.; Ineichen, U.; Cripps, R.

1994-01-01

The susceptibility of low alloy ferritic pressure vessel steels (A533-B type) to stress corrosion cracking (SCC) degradation has been examined using various BWR type coolant chemistries. Fatigue pre-cracked wedge-loaded double cantilever beams and also constantly loaded 25 mm thick compact tension specimens have shown classical SCC attack. The influence of parameters such as dissolved oxygen content, water impurity level and conductivity, material chemical composition (sulphur content) and stress intensity level are discussed. The relevance of SCC as a life-limiting degradation mechanism for low alloy ferritic nuclear power plant PV steel is examined. Some parameters, thought to be relevant for modelling SCC processes in low alloy steels in simulated BWR-type coolant, are discussed. 8 refs., 1 fig., 4 tabs

Welding of the A1 reactor pressure vessel

International Nuclear Information System (INIS)

Becka, J.

1975-01-01

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/cm 2 the welded joints in the reactor core are exposed to an integral dose of 3x10 18 n/cm 2 . 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.)

Burst pressure investigation of filament wound type IV composite pressure vessel

Science.gov (United States)

Farhood, Naseer H.; Karuppanan, Saravanan; Ya, H. H.; Baharom, Mohamad Ariff

2017-12-01

Currently, composite pressure vessels (PVs) are employed in many industries such as aerospace, transportations, medical etc. Basically, the use of PVs in automotive application as a compressed natural gas (CNG) storage cylinder has been growing rapidly. Burst failure due to the laminate failure is the most critical failure mechanism for composite pressure vessels. It is predominantly caused by excessive internal pressure due to an overfilling or an overheating. In order to reduce fabrication difficulties and increase the structural efficiency, researches and studies are conducted continuously towards the proper selection of vessel design parameters. Hence, this paper is focused on the prediction of first ply failure pressure for such vessels utilizing finite element simulation based on Tsai-Wu and maximum stress failure criterions. The effects of laminate stacking sequence and orientation angle on the burst pressure were investigated in this work for a constant layered thickness PV. Two types of winding design, A [90°2/∓θ16/90°2] and B [90°2/∓θ]ns with different orientations of helical winding reinforcement were analyzed for carbon/epoxy composite material. It was found that laminate A sustained a maximum burst pressure of 55 MPa for a sequence of [90°2/∓15°16/90°2] while the laminate B returned a maximum burst pressure of 45 MPa corresponding to a stacking sequence of [90°2/±15°/90°2/±15°/90°2/±15° ....] up to 20 layers for a constant vessel thickness. For verification, a comparison was done with the literature under similar conditions of analysis and good agreement was achieved with a maximum difference of 4% and 10% for symmetrical and unsymmetrical layout, respectively.

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 used...

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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

International Nuclear Information System (INIS)

McHenry, H.I.; Alers, G.A.

1998-01-01

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

How to replace a reactor pressure vessel

International Nuclear Information System (INIS)

Huber, R.

1996-01-01

A potential life extending procedure for a nuclear reactor after, say, 40 years of service life, might in some circumstances be the replacement of the reactor pressure vessel. Neutron induced degradation of the vessel might make replacement by one of a different material composition desirable, for example. Although the replacement of heavy components, such as steam generators, has been possible for many years, the pressure vessel presents a much more demanding task if only because it is highly irradiated. Some preliminary feasibility studies by Siemens are reported for the two removal strategies that might be considered. These are removal of the entire pressure vessel in one piece and dismantling it into sections. (UK)

Residual stresses in a weldment of pressure vessel steel

International Nuclear Information System (INIS)

Gott, K.E.

1978-01-01

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/mm 2 over a distance of 5mm. Such large variations have been found to coincide with the heat-affected zones of the individual weld runs. (author)

Acoustic emission during the elastic-plastic deformation of low alloy reactor pressure vessel steels. I

International Nuclear Information System (INIS)

Holt, J.; Goddard, D.J.

1980-01-01

Measurements of the acoustic emission behaviour of A533B and C-Mn low alloy reactor pressure vessel steels subjected to uniaxial tensile deformation are described. The effects on the emission activity of the rolling plane orientation and the carbide morphology were examined. Detailed discussions are given of the stress dependence of the emission activity below yield and of its recovery by annealing at the stress relief temperature. It is shown that the dominant emission source is the same in both steels and is associated with inclusions, such as MnS, elongated by the rolling process, the carbide morphology being relatively unimportant. A criterion for the occurrence of an emission is obtained which is directly analogous to the general criterion for yielding. It is also shown that a large fraction, at least, of the emission activity arises from a recoverable process such as localized yielding around inclusions or limited inclusion decohesion and not from inclusion fracture. Low activity in C-Mn steel taken from reactor pressure vessels, previously attributed to spheroidization of carbides, is shown to be due to the limited acoustic recovery of these relatively high sulphur content steels when annealed at the stress relief temperature. It is concluded that the limited amplitudes of these emissions during deformation severely restrict their potential application in practice. (Auth.)

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)

Fabry, A.; Chaouadi, T.; Puzzolante, J.L.; Van de Velde, J.; Biemiller, E.C.; Rossinski, S.T.; Carter, R.G.

1996-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Fabry, A.; Chaouadi, T.; Puzzolante, J.L.; Van de Velde, J. [Centre de l``Etude de l``Energie Nucleaire, Mol (Belgium); Biemiller, E.C. [Yankee Atomic Electric Company, Bolton (United States); Rossinski, S.T.; Carter, R.G. [Electric Power Research Institute, Charlotte (United States)

1996-07-01

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

Pressure vessel for nuclear reactors

International Nuclear Information System (INIS)

1975-01-01

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

A determination of the benefits of annealing irradiated pressure vessel weldments

International Nuclear Information System (INIS)

Lott, R.G.; Mager, T.R.

1988-01-01

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 450 0 C. The long-term benefit was less obvious at 400 0 C and no significant benefit was noted at 350 0 C. 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)

Heat treatment device for extending the life of a pressure vessel, particularly a reactor pressure vessel

International Nuclear Information System (INIS)

Krauss, P.; Mueller, E.; Poerner, H.; Weber, R.

1979-01-01

A support body in the form of an insulating cylinder is tightly sealed by connected surfaces at its outer circumference to the inner wall of the pressure vessel. It forms an annular heating space. The heat treatment or tempering of the pressure vessel takes place with the reactor space empty and screened from the outside by ceiling bolts. Heating gas or an induction winding can be used as the means of heating. (DG) [de

Pressure vessel for a BWR type reactor

International Nuclear Information System (INIS)

Shimamoto, Yoshiharu.

1980-01-01

Purpose: To prevent the retention of low temperature water and also prevent the thermal fatigue of the pressure vessel by making large the curvature radius of a pressure vessel of a feed water sparger fitting portion and accelerating the mixing of low-temperature water at the feed water sparger base and in-pile hot water. Constitution: The curvature radius of the corner of the feed water sparger fitting portion in a pressure vessel is formed largely. In-pile circulating water infiltrates up to the base portion of the feed water sparger to carry outside low-temperature water at the base part, which is mixed with in-pile hot water. Accordingly, low temperature water does not stay at the base portion of the feed water sparger and generation of thermal fatigue in the pressure vessel can be prevented and the safety of the BWR type reactor can be improved. (Yoshino, Y.)

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.

Application of small specimens to fracture mechanics characterization of irradiated pressure vessel steels

International Nuclear Information System (INIS)

Sokolov, M.A.; Wallin, K.; McCabe, D.E.

1996-01-01

In this study, precracked Charpy V-notch (PCVN) specimens were used to characterize the fracture toughness of unirradiated and irradiated reactor pressure vessel steels in the transition region by means of three-point static bending. Fracture toughness at cleavage instability was calculated in terms of elastic-plastic K Jc values. A statistical size correction based upon weakest-link theory was performed. The concept of a master curve was applied to analyze fracture toughness properties. Initially, size-corrected PCVN data from A 533 grade B steel, designated HSST Plate O2, were used to position the master curve and a 5% tolerance bound for K Jc data. By converting PCVN data to IT compact specimen equivalent K Jc data, the same master curve and 5% tolerance bound curve were plotted against the Electric Power Research Institute valid linear-elastic K Jc database and the ASME lower bound K Ic curve. Comparison shows that the master curve positioned by testing several PCVN specimens describes very well the massive fracture toughness database of large specimens. These results give strong support to the validity of K Jc with respect to K Ic in general and to the applicability of PCVN specimens to measure fracture toughness of reactor vessel steels in particular. Finally, irradiated PCVN specimens of other materials were tested, and the results are compared to compact specimen data. The current results show that PCVNs demonstrate very good capacity for fracture toughness characterization of reactor pressure vessel steels. It provides an opportunity for direct measurement of fracture toughness of irradiated materials by means of precracking and testing Charpy specimens from surveillance capsules. However, size limits based on constraint theory restrict the operational test temperature range for K Jc data from PCVN specimens. 13 refs., 8 figs., 1 tab

Fracture toughness behavior and its analysis on nuclear pressure vessel steels

Energy Technology Data Exchange (ETDEWEB)

Iwadate, Tadao; Tanaka, Yasuhiko; Ono, Shin-ichi; Tsukada, Hisashi [Japan Steel Works Ltd., Muroran, Hokkaido. Muroran Plant

1983-02-01

A drop weight J sub(Id) testing machine has been developed successfully, by which the multiple specimen J resistance curve test technique can be applied to measure the fracture toughness. In this study, the use of a small size round compact tension (RCT) specimen for measuring the fracture toughness J sub(Ic) or J sub(Id) of the nuclear pressure vessel steels is recommended and confirmed for the surveillance tests. The static and dynamic fracture toughness of ASTM A508 C 1.2, A508 C 1.3 and A533 Gr.B C 1.1 steels in the wide range of temperature including the upper shelf have been measured and their behavior has been analysed. The fracture toughness behavior under various strain rates and in a wide temperature range can be explained by the behavior of stretched zone formation preceding the crack initiation. The scatter of K sub(J) values in the transition range is caused by the amount of crack extension contained in the specimens. In this paper, the method to obtain the fracture toughness equivalent to the K sub(Ic) from the K sub(J) value is also presented.

Investigation of the Impact of ENDF/B-VI Cross Sections on the H.B. Robinson-2 Pressure-Vessel Flux Prediction

International Nuclear Information System (INIS)

Remec, I

1999-01-01

This report discusses the impact of the change from the SAILOR cross-section library, based on the ENDF/B-IV data, to the BUGLE-96 cross-section library, based on the ENDF/B-VI data, on the neutron flux prediction in the H. B. Robinson-2 pressure vessel, in the surveillance capsule, and in the cavity. The fast flux (E > 1 MeV) from the transport calculations with the BUGLE-96 library is approximately6% higher in the surveillance capsule and at the PV inner wall, and approximately25% higher in the reactor cavity than the flux from the transport calculations with the SAILOR library. These changes result from the combined effect of the changes in the cross sections, which cause significant increases in the calculated fluxes, and much smaller decreases in the fast fluxes due to the changes in the fission spectra. The increase in the calculated fast flux due to the changes in the cross sections only is approximately9% in the capsule and at the pressure vessel (PV) wall, and approximately30% in the cavity. The changes in the fission spectra lead to decreases in the order of approximately3-4% in calculated fast fluxes

Application of micromechanical models of ductile fracture initiation to reactor pressure vessel materials

International Nuclear Information System (INIS)

Chaouadi, R.; Walle, E. van; Fabry, A.; Velde, J. van de; Meester, P. de

1996-01-01

The aim of the current study is the application of local micromechanical models to predict crack initiation in ductile materials. Two reactor pressure vessel materials have been selected for this study: JRQ IAEA monitor base metal (A533B Cl.1) and Doel-IV weld material. Charpy impact tests have been performed in both un-irradiated and irradiated conditions. In addition to standard tensile tests, notched tensile specimens have been tested. The upper shelf energy of the weld material remains almost un-affected by irradiation, whereas a decrease of 20% is detected for the base metal. Accordingly, the tensile properties of the weld material do not reveal a clear irradiation effect on the yield and ultimate stresses, this in contrast to the base material flow properties. The tensile tests have been analyzed in terms of micromechanical models. A good correlation is found between the standard tests and the micromechanical models, that are able to predict the ductile damage evolution in these materials. Additional information on the ductility behavior of these materials is revealed by this micromechanical analysis

Welding in repair of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Pilous, V.; Kovarik, R.

1987-01-01

Specific welding conditions are described in repair of the pressure vessels of nuclear reactors in operation and the effect is pointed out to of neutrons on changes in steel properties. Some of the special regulations are discussed to be observed in welding jobs. The welding methods are briefly described; the half-bead method is most frequently used. It is stressed that the defect must first be identified using a nondestructive method and the stages must be defined of the welding repair of the pressure vessel. (J.B.). 4 figs., 1 tab., 16 refs

Pressure vessels and methods of sealing leaky tubes disposed in pressure vessels

International Nuclear Information System (INIS)

Larson, G.C.

1980-01-01

This invention relates to pressure vessels and to methods of sealing leaky tubes in them and is especially applicable to pressure vessels in the form of sheet-and-tube type heat exchangers constructed with a large number of relatively small diameter tubes grouped in a bundle. To seal off a leaky tube in such a heat exchanger an explosive activated plug in the form of a hollow metal body is used, inserted at each end of the tube to be sealed. Using the arrangement of pressure vessel and associated tube sheets and the explosive activated plug method of sealing a leaky tube as described in this invention it is claimed that distortion of the adjacent tubes and the tube sheets is reduced when the explosive activated plugs are detonated. (U.K.)

Analysis of the irradiation data for A302B and A533B correlation monitor materials

International Nuclear Information System (INIS)

Wang, J.A.

1996-04-01

The results of Charpy V-notch impact tests for A302B and A533B-1 Correlation Monitor Materials (CMM) listed in the surveillance power reactor data base (PR-EDB) and material test reactor data base (TR-EDB) are analyzed. The shift of the transition temperature at 30 ft-lb (T 30 ) is considered as the primary measure of radiation embrittlement in this report. The hyperbolic tangent fitting model and uncertainty of the fitting parameters for Charpy impact tests are presented in this report. For the surveillance CMM data, the transition temperature shifts at 30 ft-lb (ΔT 30 ) generally follow the predictions provided by Revision 2 of Regulatory Guide 1.99 (R.G. 1.99). Difference in capsule temperatures is a likely explanation for large deviations from R.G. 1.99 predictions. Deviations from the R.G. 1.99 predictions are correlated to similar deviations for the accompanying materials in the same capsules, but large random fluctuations prevent precise quantitative determination. Significant scatter is noted in the surveillance data, some of which may be attributed to variations from one specimen set to another, or inherent in Charpy V-notch testing. The major contributions to the uncertainty of the R.G. 1.99 prediction model, and the overall data scatter are from mechanical test results, chemical analysis, irradiation environments, fluence evaluation, and inhomogeneous material properties. Thus in order to improve the prediction model, control of the above-mentioned error sources needs to be improved. In general the embrittlement behavior of both the A302B and A533B-1 plate materials is similar. There is evidence for a fluence-rate effect in the CMM data irradiated in test reactors; thus its implication on power reactor surveillance programs deserves special attention

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

International Nuclear Information System (INIS)

Baum, M.R.

1989-04-01

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

Fatigue crack growth characteristics of a533 brade b glass i plate in an environment of high-temperature primary grade nuclear reactor water

International Nuclear Information System (INIS)

Mager, T.R.; Moon, D.M.; Landes, J.D.

1976-01-01

To characterize the effect of environment on crack growth rate properties of reactor pressure vessel materials, a program was initiated as part of the Heavy Section Steel Technology Program (HSST) to evaluate the effect of Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) primary grade coolant environments. The experiments included such variables as frequency, temperature and R ratio. This paper describes the investigation and presents the results of a fracture mechanics evaluation of the fatigue crack growth rate tests of A533 Grade B Class 1 steel plate material in an environment of primary reactor grade water at 550 0 F (288 0 C). A compliance crack growth monitoring technique was utilized to measure the crack growth. The compliance crack length monitor uses a linear variable differential transformer (LVDT) to measure the specimen front face displacement which is converted to crack length by the appropriate compliance calibration curve. The crack growth rate tests were conducted on constant load universal fatigue machines, under sinusoidal tension to tension loading conditions. Tests showed an increase in growth rates at a frequency of 1 cpm over previous results obtained at frequencies of 60 cpm and higher. This increase, the general character of the crack growth rate versus the $DELTA$K curve, and the results from fractographic studies, all indicated that stress corrosion cracking might have occurred for this material and environment. However, a specimen loaded statically in a PWR environment showed no static load crack growth. 13 refs

Nuclear reactor installation with outer shell enclosing a primary pressure vessel

International Nuclear Information System (INIS)

1975-01-01

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

Seals for sealing a pressure vessel such as a nuclear reactor vessel or the like

International Nuclear Information System (INIS)

Bruns, H.J.; Huelsermann, K.H.

1975-01-01

A description is given of seals for sealing a pressure vessel such as a nuclear reactor vessel, steam boiler vessel, or any other vessel which is desirably sealed against pressure of the type including a housing and a housing closure that present opposed vertical sealing surfaces which define the sides of a channel. The seals of the present invention comprise at least one sealing member disposed in the channel, having at least one stop face, a base portion and two shank portions extending from the base portion to form a groove-like recess. The shank portions are provided with sealing surfaces arranged to mate with the opposed vertical pressure vessel sealing surfaces. A shank-spreading wedge element also disposed in the channel has at least one stop face and is engaged in the groove-like recess with the sealing member and wedge element stop face adjacent to each other

Code boiler and pressure vessel life assessment

International Nuclear Information System (INIS)

Farr, J.R.

1992-01-01

In the United States of America and in Canada, laws and controls for determining life assessment for continued operation of equipment exist only for those pressure vessels built to Section III and evaluated according to Section XI. In this presentation, some of those considerations which are made in the USA and Canada for deciding on life or condition assessment of boilers and pressure vessels designed and constructed to other sections of the ASME Boiler and Pressure Vessel Code are reviewed. Life assessment or condition assesssment is essential in determining what is necessary for continued operation. With no ASME rules being adopted by laws or regulations, other than OSHA in the USA and similar environmental controls in Canada, to control life assessment for continued operation, the equipment owner must decide if assessment is to be done and how much to do. Some of those considerations are reviewed along with methods and procedures to make an assessment along with a discussion of where the ASME B and PV Code currently stands regarding continued operation. (orig.)

Pressure vessel integrity 1991

International Nuclear Information System (INIS)

Bhandari, S.; Doney, R.O.; McDonald, M.S.; Jones, D.P.; Wilson, W.K.; Pennell, W.E.

1991-01-01

This volume contains papers relating to the structural integrity assessment of pressure vessels and piping, with special emphasis on nuclear industry applications. The papers were prepared for technical sessions developed under the sponsorship of the ASME Pressure Vessels and Piping Division Committees for Codes and Standards, Computer Technology, Design and Analysis, and Materials Fabrication. They were presented at the 1991 Pressure Vessels and Piping Division Conference in San Diego, California, June 23-27. The primary objective of the sponsoring organization is to provide a forum for the dissemination and discussion of information on development and application of technology for the structural integrity assessment of pressure vessels and piping. This publication includes contributions from authors from Australia, France, Japan, Sweden, Switzerland, the United Kingdom, and the United States. The papers here are organized in six sections, each with a particular emphasis as indicated in the following section titles: Fracture Technology Status and Application Experience; Crack Initiation, Propagation and Arrest; Ductile Tearing; Constraint, Stress State, and Local-Brittle-Zones Effects; Computational Techniques for Fracture and Corrosion Fatigue; and Codes and Standards for Fatigue, Fracture and Erosion/Corrosion

Flexible Composite-Material Pressure Vessel

Science.gov (United States)

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

2003-01-01

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

Tribology aspects of a pressure vessel closure subjected to pressure cycling

International Nuclear Information System (INIS)

George, A.F.; Williams, M.E.

1988-04-01

A repair method being considered for a steel pressure vessel is to cut away the faulty part leaving an unreinforced circular hole in the curved wall and cover it with a sealed plate placed inside. In order to investigate the structural properties of such a repair a large model vessel (6m by 2m) was tested under pressure (about 2.5 MPa) and pressure cycling. This cycling caused relative movements at the loaded interface between the lid and the vessel. A tribological examination of the rubbing surfaces was carried out. The tribological examination is described and a small supporting programme of laboratory scaling tests. It gives the results and attempts to interpret them with particular attention given to wear, fretting fatigue and scaling to plant conditions. (author)

Design optimization of a thin walled pressure vessel

International Nuclear Information System (INIS)

Sadiq, S.

2001-01-01

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

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

International Nuclear Information System (INIS)

Froehling, W.; Boettcher, A.; Bounin, D.; Steinwarz, W.; Geiss, M.; Trauth, M.

2000-01-01

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.) [de

Personal computer based system to evaluate J-integral by a single specimen unloading compliance method: Pt. 2. Results on A533B Class I steel and CORONA-5

Energy Technology Data Exchange (ETDEWEB)

Narendrnath, K R; Margolin, H; Jung, Y H; Godavarti, P S; Murty, K L

1988-01-01

A personal computer based single specimen unloading technique has been demonstrated using a nuclear grade pressure vessel material, A533B Class I steel, and an ..cap alpha..-..beta.. titanium alloy, CORONA-5. Experimental parameters investigated in the case of steel include the effect of test temperature, loading rate and neutron irradiation. Results revealed the effect of dynamic strain aging (DSA) as a drop in the critical crack initiation fracture toughness (Jsub(q)); the minimum in Jsub(q) shifts to higher temperatures at higher test speeds as predicted from DSA models. In CORONA-5, the effects of microstructure and heat treatment for two different sizes of equiaxed ..cap alpha.. morphology in a ..beta..-matrix heat treated to different yield strengths were investigated. It is found that the observed higher values of Jsub(q) and tearing modulus result from a greater tortuosity of the crack path and ease of crack blunting. The largest increase in Jsub(q) occurred in fine equiaxed ..cap alpha.. microstructure aged to a nominal yield strength of 896 MPa. Results reveal potential application of the low cost technique for determining fracture toughness of engineering materials.

Fatigue crack growth behavior of pressure vessel steels and submerged arc weldments in a high-temperature pressurized water environment

International Nuclear Information System (INIS)

Liaw, P.K.; Logsdon, W.A.; Begley, J.A.

1989-01-01

The fatigue crack growth rate (FCGR) properties of SA508 Cl 2a and SA533 Gr A Cl 2 pressure vessel steels and the corresponding automatic submerged arc weldments were developed in a high-temperature pressurized water (HPW) environment at 288 degrees C (550 degrees F) and 7.2 MPa (1044 psi) at load ratios of 0.20 and 0.50. The properties were generally conservative compared to American Society of Mechanical Engineers Section XI water environment reference curve. The growth rate of fatigue cracks in the base materials, however, was faster in the HPW environment than in a 288 degrees C (550 degrees F) base line air environment. The growth rate of fatigue cracks in the two submerged arc weldments was also accelerated in the HPW environment but to a lesser degree than that demonstrated by the base materials. In the air environment, fatigue striations were observed, independent of material and load ratio, while in the HPW environment, some intergranular facets were present. The greater environmental effect on crack growth rates displayed by the base materials compared the weldments attributed to a different sulfide composition and morphology

Appropriate welding conditions of temper bead weld repair for SQV2A pressure vessel steel

International Nuclear Information System (INIS)

Mizuno, R.; Matsuda, F.; Brziak, P.; Lomozik, M.

2004-01-01

Temper bead welding technique is one of the most important repair welding methods for large structures for which it is difficult to perform the specified post weld heat treatment. In this study, appropriate temper bead welding conditions to improve the characteristics of heat affected zone (HAZ) are studied using pressure vessel steel SQV2A corresponding to ASTM A533 Type B Class 1. Thermal/mechanical simulator is employed to give specimens welding thermal cycles from single to quadruple cycle. Charpy absorbed energy and hardness of simulated CGHAZ by first cycle were degraded as compared with base metal. Improvability of these degradations by subsequent cycles is discussed and appropriate temper bead thermal cycles are clarified. When the peak temperature lower than Ac1 and near Ac1 in the second thermal cycle is applied to CGAHZ by first thermal cycle, the characteristics of CGHAZ improve enough. When the other peak temperatures (that is, higher than Ac1) in the second thermal cycle are applied to the CGHAZ, third or more thermal cycle temper bead process should be applied to improve the properties. Appropriate weld condition ranges are selected based on the above results. The validity of the selected ranges is verified by the temper bead welding test. (orig.)

Nuclear power plant pressure vessels. Control of piping

International Nuclear Information System (INIS)

2000-01-01

The guide presents requirements for the pipework of nuclear facilities in Finland. According to the section 117 of the Finnish Nuclear Energy Degree (161/88), the Radiation and Nuclear Safety Authority of Finland (STUK) controls the pressure vessels of nuclear facilities in accordance with the Nuclear Energy Act (990/87) and, to the extent applicable in accordance with the Act of Pressure Vessels (98/73) and the rules and regulations issued by the virtue of these. In addition STUK is an inspecting authority of pressure vessels of nuclear facilities in accordance with the Pressure Vessel Degree (549/1973). According to the section of the Pressure Vessel Degree, a pressure vessel is a steam boiler, pressure container, pipework of other such appliance in which the pressure is above or may come to exceed the atmospheric pressure. Guide YVL 3.0 describes in general terms how STUK controls pressure vessels. STUK controls Safety Class 1, 2 and 3 piping as well as Class EYT (non-nuclear) and their support structures in accordance with this guide and applies the provisions of the Decision of the Ministry of Trade and Industry on piping (71/1975) issued by virtue of the Pressure Vessel Decree

Stress analysis and evaluation of a rectangular pressure vessel

International Nuclear Information System (INIS)

Rezvani, M.A.; Ziada, H.H.; Shurrab, M.S.

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

High-performance fiber/epoxy composite pressure vessels

Science.gov (United States)

Chiao, T. T.; Hamstad, M. A.; Jessop, E. S.; Toland, R. H.

1978-01-01

Activities described include: (1) determining the applicability of an ultrahigh-strength graphite fiber to composite pressure vessels; (2) defining the fatigue performance of thin-titanium-lined, high-strength graphite/epoxy pressure vessel; (3) selecting epoxy resin systems suitable for filament winding; (4) studying the fatigue life potential of Kevlar 49/epoxy pressure vessels; and (5) developing polymer liners for composite pressure vessels. Kevlar 49/epoxy and graphite fiber/epoxy pressure vessels, 10.2 cm in diameter, some with aluminum liners and some with alternation layers of rubber and polymer were fabricated. To determine liner performance, vessels were subjected to gas permeation tests, fatigue cycling, and burst tests, measuring composite performance, fatigue life, and leak rates. Both the metal and the rubber/polymer liner performed well. Proportionately larger pressure vessels (20.3 and 38 cm in diameter) were made and subjected to the same tests. In these larger vessels, line leakage problems with both liners developed the causes of the leaks were identified and some solutions to such liner problems are recommended.

Model tests for prestressed concrete pressure vessels

International Nuclear Information System (INIS)

Stoever, R.

1975-01-01

Investigations with models of reactor pressure vessels are used to check results of three dimensional calculation methods and to predict the behaviour of the prototype. Model tests with 1:50 elastic pressure vessel models and with a 1:5 prestressed concrete pressure vessel are described and experimental results are presented. (orig.) [de

Nuclear power plant pressure vessels. Inservice inspections

International Nuclear Information System (INIS)

1995-01-01

The requirements for the planning and reporting of inservice inspections of nuclear power plant pressure vessels are presented. The guide specifically applies to inservice inspections of Safety class 1 and 2 nuclear power plant pressure vessels, piping, pumps and valves plus their supports and reactor pressure vessel internals by non- destructive examination methods (NDE). Inservice inspections according to the Pressure Vessel Degree (549/73) are discussed separately in the guide YVL 3.0. (4 refs.)

Reactor pressure vessel design

International Nuclear Information System (INIS)

Foehl, J.

1998-01-01

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 2, the general principles of reactor pressure vessel design are elaborated. Crack and fracture initiation and propagation are treated in some detail

Crack arrest toughness measurements with A533B steel

International Nuclear Information System (INIS)

Salonen, Seppo.

1979-11-01

This work covers crack arrest toughness measurements on A533B steel done at the Technical Research Centre of Finland. These measurements are one part of a multinational effort, involving 30 laboratories. The aim of the cooperative test program is to examine two test procedures for measuring the crack arrest toughness, to give information about their reproducibility, and to identify the factors affecting the interpretation. The principles given for the testing were easy to apply in general and the results were satisfactory. Some factors in the test runs and in the specimen's behaviour are indicated which can cause error in the results or make implementation of the test more difficult. By comparing the results from our laboratory with average values from the test program a good agreement can be seen. Crack arrest toughness values derived from the compared procedures with a static analysis agree closely, but values calculated using a dynamic analysis differ considerably. (author)

Pressure test method for reactor pressure vessel in construction field

International Nuclear Information System (INIS)

Takeda, Masakado; Ushiroda, Koichi; Miyahara, Ryohei; Takano, Hiroshi; Matsuura, Tadashi; Sato, Keiya.

1998-01-01

Plant constitutional parts as targets of both of a primary pressure test and a secondary pressure test are disposed in communication with a reactor pressure vessel, and a pressure of the primary pressure test is applied to the targets of both tests, so that the primary pressure test and the second pressure test are conducted together. Since the number of pressure tests can be reduced to promote construction, and the number of workers can also be reduced. A pressure exceeding the maximum pressure upon use is applied to the pressure vessel after disposing the incore structures, to continuously conduct the primary pressure test and the secondary pressure test joined together and an incore flowing test while closing the upper lid of the pressure vessel as it is in the construction field. The number of opening/closing of the upper lid upon conducting every test can be reduced, and since the pressure resistance test is conducted after arranging circumference conditions for the incore flowing test, the tests can be conducted collectively also in view of time. (N.H.)

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.

Safety of light-water reactor pressure vessels against brittle fracture

International Nuclear Information System (INIS)

Brumovsky, M.

1979-01-01

The results are surveyed of research by SKODA Trust into brittle failure resistance of materials for WWER type reactor pressure vessels and into pressure vessel operating safety. Conditions are discussed in detail decisive for initiation, propagation and arrest of brittle fracture. The tests on the Cr-Mo-V type steel showed high resistance of the steel to the formation and the propagation of brittle fracture. They also confirmed the high operating reliability and the required service life of the steel. (B.S.)

Metallurgy of steels for PWR pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Kepka, M; Mocek, J; Barackova, L [Skoda, Plzen (Czechoslovakia)

1980-09-01

A survey and the chemical compositions are presented of reactor pressure vessel steels. The metallurgy is described of steel making for pressure vessels in Japan and the USSR. Both acidic and alkaline open-hearth steel is used for the manufacture of ingots. The leading world manufacturers of forging ingots for pressure vessels, however, exclusively use electric steel. Vacuum casting techniques are exclusively used. Experience is shown gained with the introduction of the manufacture of forging ingots for pressure vessels at SKODA, Plzen. The metallurgical procedure was tested utilizing alkaline open hearths, electric arc furnaces and facilities for vacuum casting of steel. Pure charge raw materials should be used for securing high steel purity. Prior to forging pressure vessel rings, not only should sufficiently big bottoms and heads be removed but also the ingot middle part should be scrapped showing higher contents of impurities and nonhomogeneous structure.

Hardening and microstructural evolution of A533b steels irradiated with Fe ions and electrons

Energy Technology Data Exchange (ETDEWEB)

Watanabe, H., E-mail: watanabe@riam.kyushu-u.ac.jp [Research Institute for Applied Mechanics, Kyushu University, 6-1, Kasuga-kouenn, Kasugashi, Fukuoka, 816-8580 (Japan); Arase, S. [Interdisciplinary Graduate School of Kyushu University, 6-1, Kasuga-kouenn, Kasugashi, Fukuoka, 816-8580 (Japan); Yamamoto, T.; Wells, P. [Dept. Chemical Engineering, UCSB Engineering II, RM3357, Santa Barbara, CA, 93106-5080 (United States); Onishi, T. [Interdisciplinary Graduate School of Kyushu University, 6-1, Kasuga-kouenn, Kasugashi, Fukuoka, 816-8580 (Japan); Odette, G.R. [Dept. Chemical Engineering, UCSB Engineering II, RM3357, Santa Barbara, CA, 93106-5080 (United States)

2016-04-01

Radiation hardening and embrittlement of A533B steels is heavily dependent on the Cu content. In this study, to investigate the effect of copper on the microstructural evolution of these materials, A533B steels with different Cu levels were irradiated with 2.4 MeV Fe ions and 1.0 MeV electrons. Ion irradiation was performed from room temperature (RT) to 350 °C with doses up to 1 dpa. At RT and 290 °C, low dose (<0.1 dpa) hardening trend corresponded with ΔH ∝ (dpa){sup n}, with n initially approximately 0.5 and consistent with a barrier hardening mechanism, but saturating at ≈0.1 dpa. At higher dose levels, the radiation-induced hardening exhibited a strong Cu content dependence at 290 °C, but not at 350 °C. Electron irradiation using high-voltage electron microscopy revealed the growth of interstitial-type dislocation loops and enrichment of Ni, Mn, and Si in the vicinities of pre-existing dislocations at doses for which the radiation-induced hardness due to ion irradiation was prominent.

Manipulator for testing a top-opened reactor pressure vessel

International Nuclear Information System (INIS)

Bauer, R.; Kastl, H.

1991-01-01

The design is described of a manipulator to be inserted into the inside of reactor pressure vessels opened at the top. The main components of the manipulator include a fixed column protruding into the pressure vessel and a support which is slidable on the column and carries the bearing component for the measuring, testing, inspection and repair instruments. The device includes a driving equipment for the support as well as the power supply for the sets accommodated on the support, with the aim to reduce the failure rate of the manipulator as a whole, shorten the time necessary for its assembling and thus the time of staying in the reactor pressure vessel and, at the same time, make its maintenance and operation easier. (Z.S.). 13 figs

Conformable pressure vessel for high pressure gas storage

Science.gov (United States)

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.

Relaxation response of A533B steel from 25 to 600/degree/C

International Nuclear Information System (INIS)

Swindeman, R.W.; Bolling, E.

1989-01-01

Relaxation tests were performed on A533B steel over the range 25 to 600/degree/C in order to examine the general features of time- dependent deformation. It was found that the relaxation strength increased with the flow stress at low temperatures and was relatively independent of history at high temperatures. In the temperature range 400 to 600/degree/C the inelastic strain rates calculated from the relaxation rates followed stress dependencies that were consistent with expectations based on a model proposed by Hart and coworkers for matrix deformation. 21 refs., 10 figs

Power reactor pressure vessel benchmarks

International Nuclear Information System (INIS)

Rahn, F.J.

1978-01-01

A review is given of the current status of experimental and calculational benchmarks for use in understanding the radiation embrittlement effects in the pressure vessels of operating light water power reactors. The requirements of such benchmarks for application to pressure vessel dosimetry are stated. Recent developments in active and passive neutron detectors sensitive in the ranges of importance to embrittlement studies are summarized and recommendations for improvements in the benchmark are made. (author)

Reactor vessel pressure transient protection for pressurized water reactors

International Nuclear Information System (INIS)

Zech, G.

1978-09-01

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

46 CFR 115.812 - Pressure vessels and boilers.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Pressure vessels and boilers. 115.812 Section 115.812... CERTIFICATION Material Inspections § 115.812 Pressure vessels and boilers. (a) Pressure vessels must be tested... testing requirements for boilers are contained in § 61.05 in subchapter F of this chapter. [CGD 85-080, 61...

Head spray nozzle in reactor pressure vessel

International Nuclear Information System (INIS)

Hatano, Shun-ichi.

1990-01-01

In a reactor pressure vessel of a BWR type reactor, a head spray nozzle is used for cooling the head of the pressure vessel and, in view of the thermal stresses, it is desirable that cooling is applied as uniformly as possible. A conventional head spray is constituted by combining full cone type nozzles. Since the sprayed water is flown down upon water spraying and the sprayed water in the vertical direction is overlapped, the flow rate distribution has a high sharpness to form a shape as having a maximum value near the center and it is difficult to obtain a uniform flow rate distribution in the circumferential direction. Then, in the present invention, flat nozzles each having a spray water cross section of laterally long shape, having less sharpness in the circumferential distribution upon spraying water to the inner wall of the pressure vessel and having a wide angle of water spray are combined, to make the flow rate distribution of spray water uniform in the inner wall of the pressure vessel. Accordingly, the pressure vessel can be cooled uniformly and thermal stresses upon cooling can be decreased. (N.H.)

The pressure vessel for the NSF tandem

International Nuclear Information System (INIS)

Jones, C.W.

1979-04-01

The pressure vessel is a major component of the 30 MV tandem Van de Graaff electrostatic accelerator to be used in nuclear structure research at Daresbury Laboratory. The accelerator will be capable of accelerating the full range of ions in the form of a beam. Acceleration takes place in a vertical evacuated tube (beam tube) by means of a high potential on a terminal at the central position, the terminal and beam tube assembly being supported by an insulated stack structure within the pressure vessel. Under operating conditions the vessel is filled with sulphur hexafluoride gas (SF 6 ) at high pressure which acts as an insulating medium between the centre terminal and the vessel wall. The vessel is situated inside a concrete tower which besides supporting the injector room above the vessel also acts as radiation shielding around the accelerator. The report covers: functional requirements; fundamental considerations with regard to the design and procurement; detail design; materials; manufacture; acceptance test; surface treatment; final leak test. (U.K.)

Foundamental characteristics of layered pressure vessel

International Nuclear Information System (INIS)

Moriwaki, Yoshikazu; Fugino, Masayuki; Shimizu, Yasuhiro; Nakamura, Takeshi

1978-01-01

Pressure vessels become larger and the working pressure become higher with the remarkable development of petroleum, chemical, thermal power generation and atomic energy industries. Multi-layered pressure vessels can be manufactured cheaply without large installations, and large wall thickness can be made, therefore they are suitable for large pressure vessels. The stress and deformation behaviors of such vessels are very complex because of the effect of frictional force working between layers. In this study, the phenomena arising in multiple layers and the difference as compared with single wall were studied fundamentally as one step for analyzing multi-layered pressure vessels as a whole. Finite element technique was employed as the analyzing method, and the behavior of multiple layers was analyzed, regarding it as multiple contact problem. The behavior of multiple layers seems to appear conspicuously in case of bending load, therefore the basic characteristics regarding bending were examined. The evaluation of interfacial stiffness was carried out by experiment. The computer program for analyzing multiple contact problem was developed. In order to examine the validity of the program, comparison with the analytical solution heretofore and the result of calculation by finite element technique was carried out. Moreover, the experimental proof with multi-layered models was made. The frictional force between layers hardly contributes to the stiffness. (Kako, I.)

Integrity of PWR pressure vessels during overcooling accidents

International Nuclear Information System (INIS)

Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

1982-01-01

The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents, vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. A state-of-the-art fracture-mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure in a few years if subjected to a Rancho Seco-type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation

Nuclear reactor installation with outer shell enclosing a primary pressure vessel

International Nuclear Information System (INIS)

1975-01-01

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

Leak detection device for nuclear reactor pressure vessel

International Nuclear Information System (INIS)

Ikeda, Jun.

1988-01-01

Purpose: To test the leakage of a nuclear reactor pressure vessel during stopping for a short period of time with no change to the pressure vessel itself. Constitution: The device of the present invention comprises two O-rings disposed on the flange surface that connects a pressure vessel main body and an upper cover, a leak-off pipeway derived from the gap of the O-rings at the flange surface to the outside of the pressure vessel, a pressure detection means connected to the end of the pipeway, a humidity detection means disposed to the lead-off pipeway, a humidity detection means disposed to the lead-off pipeway, and gas supply means and gas suction means disposed each by way of a check valve to a side pipe branched from the pipeway. After stopping the operation of the nuclear reactor and pressurizing the pressure vessel by filling water, gases supplied to the gap between the O-rings at the flange surface by opening the check valve. In a case where water in the pressure vessel should leak to the flange surface, when gas suction is applied by properly opening the check valve, increase in the humidity due to the steams of leaked water diffused into the gas is detected to recognize the occurrence of leakage. (Kamimura, M.)

Integrity of PWR pressure vessels during overcooling accidents

International Nuclear Information System (INIS)

Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

1982-01-01

The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. For the purpose of evaluating this problem a state-of-the-art fracture mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure today if subjected to a Rancho Seco (1978) or TMI-2 (1979) type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation

Guidelines for pressure vessel safety assessment

Science.gov (United States)

Yukawa, S.

1990-04-01

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

Computerized reactor pressure vessel materials information system

International Nuclear Information System (INIS)

Strosnider, J.; Monserrate, C.; Kenworthy, L.D.; Tether, C.D.

1980-10-01

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

In-service supervision of a prestressed concrete pressure vessel

International Nuclear Information System (INIS)

Zemann, H.; Mayer, N.; Amberg, C.

1985-01-01

On-line measurements of the physical state of a prestressed concrete pressure vessel and a comparison of the distribution of temperature, strain and stress within the concrete member to the optimized statical predictions and the criterions of layout yield to an efficient and economical method of operating the vessel with a high potential of safety. The requirements of instrumentation and the comparison with static calculations are discussed on the prototype vessel at Seibersdorf Research Center during the phase of construction and prestressing, the phase of the first thermal treatment (stabilization), the pressure tests and under the operating conditions of a high temperature reactor (150 0 C/50 bar). (Author)

Some aspects of reactor pressure vessel integrity

International Nuclear Information System (INIS)

Korosec, D.; Vojvodic, G.J.

1996-01-01

Reactor pressure vessel of the pressurized water reactor nuclear power plant is the subject of extreme interest due to the fact that presents the pressure boundary of the reactor coolant system, which is under extreme thermal, mechanical and irradiation effects. Reactor pressure vessel by itself prevents the release of fission products to the environment. Design, construction and in-service inspection of such component is governed by strict ASME rules and other forms of administrative control. The reactor pressure vessel in nuclear power plant Kriko is designed and constructed in accordance with related ASME rules. The in-service inspection program includes all requests presented in ASME Code section XI. In the present article all major requests for the periodic inspections of reactor pressure vessel and fracture mechanics analysis are discussed. Detailed and strict fulfillment of all prescribed provisions guarantee the appropriate level of nuclear safety. (author)

In-service supervision of a prestressed concrete pressure vessel

International Nuclear Information System (INIS)

Zemann, H.; Weissbacher, L.; Mayer, N.; Amberge, C.

1985-01-01

On-line measurements of the physical state of a prestressed concrete pressure vessel, and comparison with the design predictions of the distribution of temperature, strain and stress within the concrete member and the criteria of layout, provide an efficient and economical method of operating the vessel with a high potential of safety. The requirements of instrumentation and the comparison with static calculations are discussed with reference to the prototype vessel at Seibersdorf Research Centre during the phase of construction and prestressing, the phase of the first thermal treatment (stabilization), the pressure tests and under the operating conditions of a high temperature reactor (150 0 C, 50 bar). (author)

Reactor pressure vessel status report

International Nuclear Information System (INIS)

Strosnider, J.; Wichman, K.; Elliot, B.

1994-12-01

This report gives a brief description of the reactor pressure vessel (RPV), followed by a discussion of the radiation embrittlement of RPV beltline materials and the two indicators for measuring embrittlement, the end-of-license (EOL) reference temperature and the EOL upper-shelf energy. It also summarizes the GL 92-01 effort and presents, for all 37 boiling water reactor plants and 74 pressurized water reactor plants in the United States, the current status of compliance with regulatory requirements related to ensuring RPV integrity. The staff has evaluated the material data needed to predict neutron embrittlement of the reactor vessel beltline materials. These data will be stored in a computer database entitled the reactor vessel integrity database (RVID). This database will be updated annually to reflect the changes made by the licensees in future submittals and will be used by the NRC staff to assess the issues related to vessel structural integrity

Results of reactor pressure vessels ISI

International Nuclear Information System (INIS)

Cepcek, S.

1994-01-01

To find out the possible influence of the annealing process to reactor pressure vessel integrity, a large in-service inspection programme has been implemented as an associated activity to reactor pressure vessel annealing. In this paper the approach to the RPV in-service inspection is shown. Also, the main results and conclusions following in-service inspection are presented. (author). 3 refs, 1 fig

46 CFR 197.462 - Pressure vessels and pressure piping.

Science.gov (United States)

2010-10-01

... that each pressure vessel, including each volume tank, cylinder and PVHO, and each pressure piping... tests conducted in accordance with this section shall be either hydrostatic tests or pneumatic tests. (1... times the maximum allowable working pressure. (2) When a pneumatic test is conducted on a pressure...

Pressure vessel integrity and weld inspection procedure

International Nuclear Information System (INIS)

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

1975-01-01

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

Reactor pressure vessel support

International Nuclear Information System (INIS)

Butti, J.P.

1977-01-01

A link and pin support system provides the primary vertical and lateral support for a nuclear reactor pressure vessel without restricting thermally induced radial and vertical expansion and contraction. (Auth.)

Crack propagation on spherical pressure vessels

International Nuclear Information System (INIS)

Lebey, J.; Roche, R.

1975-01-01

The risk presented by a crack on a pressure vessel built with a ductile steel cannot be well evaluated by simple application of the rules of Linear Elastic Fracture Mechanics, which only apply to brittle materials. Tests were carried out on spherical vessels of three different scales built with the same steel. Cracks of different length were machined through the vessel wall. From the results obtained, crack initiation stress (beginning of stable propagation) and instable propagation stress may be plotted against the lengths of these cracks. For small and medium size, subject to ductile fracture, the resulting curves are identical, and may be used for ductile fracture prediction. Brittle rupture was observed on larger vessels and crack propagation occurred at lower stress level. Preceedings curves are not usable for fracture analysis. Ultimate pressure can be computed with a good accuracy by using equivalent energy toughness, Ksub(1cd), characteristic of the metal plates. Satisfactory measurements have been obtained on thin samples. The risks of brittle fracture may then judged by comparing Ksub(1cd) with the calculated K 1 value, in which corrections for vessel shape are taken into account. It is thus possible to establish the bursting pressure of cracked spherical vessels, with the help of two rules, one for brittle fracture, the other for ductile instability. A practical method is proposed on the basis of the work reported here

Structural integrity of water reactor pressure boundary components

International Nuclear Information System (INIS)

Loss, F.J.

1977-01-01

The dynamic fracture toughness was determined as a function of temperature for three-point bend specimens of A533-B, A508-2, and A302-B steels. Crack propagation rates at 288 0 C in a water reactor environment were determined for A533-B and A508-2. Radiation-induced degradation of notch toughness of reactor steels and welds was explored. The ''warm prestress'' occurring in a flawed reactor vessel following a LOCA and operation of ECCS was studied. 25 figures

Nickel hydrogen common pressure vessel battery development

Science.gov (United States)

Jones, Kenneth R.; Zagrodnik, Jeffrey P.

1992-01-01

Our present design for a common pressure vessel (CPV) battery, a nickel hydrogen battery system to combine all of the cells into a common pressure vessel, uses an open disk which allows the cell to be set into a shallow cavity; subsequent cells are stacked on each other with the total number based on the battery voltage required. This approach not only eliminates the assembly error threat, but also more readily assures equal contact pressure to the heat fin between each cell, which further assures balanced heat transfer. These heat fin dishes with their appropriate cell stacks are held together with tie bars which in turn are connected to the pressure vessel weld rings at each end of the tube.

49 CFR Appendix A to Part 533 - Example of Calculating Compliance Under § 533.5 Paragraph (g)

Science.gov (United States)

2010-10-01

... 49 Transportation 6 2010-10-01 2010-10-01 false Example of Calculating Compliance Under § 533.5 Paragraph (g) A Appendix A to Part 533 Transportation Other Regulations Relating to Transportation... ECONOMY STANDARDS Pt. 533, App. A Appendix A to Part 533—Example of Calculating Compliance Under § 533.5...

Reactor pressure vessel failure probability following through-wall cracks due to pressurized thermal shock events

International Nuclear Information System (INIS)

Simonen, F.A.; Garnich, M.R.; Simonen, E.P.; Bian, S.H.; Nomura, K.K.; Anderson, W.E.; Pedersen, L.T.

1986-04-01

A fracture mechanics model was developed at the Pacific Northwest Laboratory (PNL) to predict the behavior of a reactor pressure vessel following a through-wall crack that occurs during a pressurized thermal shock (PTS) event. This study, which contributed to a US Nuclear Regulatory Commission (NRC) program to study PTS risk, was coordinated with the Integrated Pressurized Thermal Shock (IPTS) Program at Oak Ridge National Laboratory (ORNL). The PNL fracture mechanics model uses the critical transients and probabilities of through-wall cracks from the IPTS Program. The PNL model predicts the arrest, reinitiation, and direction of crack growth for a postulated through-wall crack and thereby predicts the mode of vessel failure. A Monte-Carlo type of computer code was written to predict the probabilities of the alternative failure modes. This code treats the fracture mechanics properties of the various welds and plates of a vessel as random variables. Plant-specific calculations were performed for the Oconee-1, Calvert Cliffs-1, and H.B. Robinson-2 reactor pressure vessels for the conditions of postulated transients. The model predicted that 50% or more of the through-wall axial cracks will turn to follow a circumferential weld. The predicted failure mode is a complete circumferential fracture of the vessel, which results in a potential vertically directed missile consisting of the upper head assembly. Missile arrest calculations for the three nuclear plants predict that such vertical missiles, as well as all potential horizontally directed fragmentation type missiles, will be confined to the vessel enclosre cavity. The PNL failure mode model is recommended for use in future evaluations of other plants, to determine the failure modes that are most probable for postulated PTS events

Light-water reactor pressure vessel surveillance standards

International Nuclear Information System (INIS)

Anon.

1981-01-01

The master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel steels throughout a pressure vessel's service life. Some of these are existing American Society for Testing and Materials (ASTM) standards, some are ASTM standards that have been modified, and some are newly proposed ASTM standards. The current (1) scope, (2) areas of application, (3) interrelationships, and (4) status and time table of development, improvement, validation, and calibration for a series of 16 ASTM standards are defined. The standard also includes a discussion of LWR pressure vessel surveillance - justification, requirements, and status of work

Pressurized wet digestion in open vessels (T11)

International Nuclear Information System (INIS)

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

2002-01-01

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

Towards a new pressure vessel standard in the European Union

International Nuclear Information System (INIS)

Osweiller, F.

1995-01-01

Since 1990 the European Commission has been preparing a new Directive which will regulate the Pressure Equipment sector in the countries of the European Union. CEN Standards devoted to pressure vessels, piping, boilers, are currently being drawn up to complete and implement this Directive. This paper focuses on the European Unfired Pressure Vessel Standard (EPVS) which is in course of development under the responsibility of CEN/TC54. The main aspects of the Standard are outlined: general structure, materials, design, fabrication, inspection and testing. The link with the European Directive is explained in connection with regulatory aspects: conformity assessment, essential safety requirements, classes of vessels, notified bodies, EC mark, status of the standard

Performance demonstration experience for reactor pressure vessel shell ultrasonic testing

International Nuclear Information System (INIS)

Zado, V.

1998-01-01

The most ultrasonic testing techniques used by many vendors for pressurized water reactor (PWR) examinations were based on American Society of Mechanical Engineers 'Boiler and Pressurized Vessel Code' (ASME B and PV Code) Sections XI and V. The Addenda of ASME B and PV Code Section XI, Edition 1989 introduced Appendix VIII - 'Performance Demonstration for Ultrasonic Examination Systems'. In an effort to increase confidence in performance of ultrasonic testing of the operating nuclear power plants in United States, the ultrasonic testing performance demonstration examination of reactor vessel welds is performed in accordance with Performance Demonstration Initiative (PDI) program which is based on ASME Code Section XI, Appendix VIII requirements. This article provides information regarding extensive qualification preparation works performed prior EPRI guided performance demonstration exam of reactor vessel shell welds accomplished in January 1997 for the scope of Appendix VIII, Supplements IV and VI. Additionally, an overview of the procedures based on requirements of ASME Code Section XI and V in comparison to procedure prepared for Appendix VIII examination is given and discussed. The samples of ultrasonic signals obtained from artificial flaws implanted in vessel material are presented and results of ultrasonic testing are compared to actual flaw sizes. (author)

Additional Stress And Fracture Mechanics Analyses Of Pressurized Water Reactor Pressure Vessel Nozzles

International Nuclear Information System (INIS)

Walter, Matthew; Yin, Shengjun; Stevens, Gary; Sommerville, Daniel; Palm, Nathan; Heinecke, Carol

2012-01-01

) Conferences. This work is also relevant to the ongoing efforts of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B and PV) Code, Section XI, Working Group on Operating Plant Criteria (WGOPC) efforts to incorporate nozzle fracture mechanics solutions into a revision to ASME B and PV Code, Section XI, Nonmandatory Appendix G.

Hot cell examination on the surveillance capsule of SA 533 cl. 1 reactor pressure vessel (1st test report)

Energy Technology Data Exchange (ETDEWEB)

Choo, Yong Sun; Jung, Y. H.; Yoo, B. O.; Baik, S. J.; Oh, W. H.; Soong, W. S.; Hong, K. P

2000-08-01

The post-irradiated examinations such as impact test, tensile test, composition analysis and etc. were conducted to monitor and to evaluate the radiation-induced changes, so called radiation embrittlement, in the mechanical properties of ferritic materials. Those data should be applied to confirm safety as well as reliability of reactor pressure vessel. The scopes and contents of hot cell examination on the surveillance capsule are as follows; - Capsule transportation, cutting, dismantling and classification - Shim block and Dosimeter cutting and dismantling - Impact test - Tensile test - Composition analysis by EPMA - SEM observation on the fractured surface - Hardness test - Radwaste treatment.

A prestressed concrete pressure vessel for helium high temperature reactor system

International Nuclear Information System (INIS)

Horner, R.M.W.; Hodzic, A.

1976-01-01

A novel prestressed concrete pressure vessel has been developed to provide the primary containment for a fully integrated system comprising a high temperature nuclear reactor, three horizontally mounted helium turbines, associated heat exchangers and inter-connecting ducts. The design and analysis of the pressure vessel is described. Factors affecting the final choice of layout are discussed, and earlier development work seeking to resolve the conflicting requirements of the structural, mechanical, and system engineers outlined. Proposals to increase the present output of about 1000 MW of electrical power to over 3000 MW, by incorporating four turbines in a single pressure vessel are presented. (author)

Ultrasonic stress evaluation through thickness of a stainless steel pressure vessel

International Nuclear Information System (INIS)

Javadi, Yashar; Pirzaman, Hamed Salimi; Raeisi, Mohammadreza Hadizadeh; Najafabadi, Mehdi Ahmadi

2014-01-01

This paper investigates ultrasonic method in stress measurement through thickness of a pressure vessel. Longitudinal critically refracted (L CR ) 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 L CR method (known as FEL CR 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

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

International Nuclear Information System (INIS)

Laemmer, H.; Ritter, B.

1994-01-01

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) [de

Nuclear reactor pressure vessel flaw distribution development

International Nuclear Information System (INIS)

Kennedy, E.L.; Foulds, J.R.; Basin, S.L.

1991-12-01

Previous attempts to develop flaw distributions for probabilistic fracture mechanics analyses of pressurized water reactor (PWR) vessels have aimed at the estimation of a ''generic'' distribution applicable to all PWR vessels. In contrast, this report describes (1) a new flaw distribution development analytic methodology that can be applied to the analysis of vessel-specific inservice inspection (ISI) data, and (2) results of the application of the methodology to the analysis of flaw data for each vessel case (ISI data on three PWR vessels and laboratory inspection data on sections of the Midland reactor vessel). Results of this study show significant variation among the flaw distributions derived from the various data sets analyzed, strongly suggesting than a vessel-specific flaw distribution (for vessel integrity prediction under pressurized thermal shock) is preferred over a ''generic'' distribution. In addition, quantitative inspection system flaw sizing accuracy requirements have been identified for developing a flaw distribution from vessel ISI data. The new flaw data analysis methodology also permits quantifying the reliability of the flaw distribution estimate. Included in the report are identified needs for further development of several aspects of ISI data acquisition and vessel integrity prediction practice

Holographic and acoustic emission evaluation of pressure vessels

International Nuclear Information System (INIS)

Boyd, D.M.

1980-01-01

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

Procurement of replacement pressure vessels for MURR

International Nuclear Information System (INIS)

Meyer, W.A. Jr.; Edwards, C.B. Jr.; McKibben, J.C.; Schoone, A.R.

1989-01-01

The University of Missouri Research Reactor Facility (MURR) located in Columbia, Missouri, is the highest powered, highest steady-state flux university research reactor in the United States. The reactor is a 10-MW pressurized loop, in-pool-type, light-water-moderated, beryllium-reflected, flux trap reactor. MURR has a compact core (0.033 m 3 ) composed of eight fuel elements of the materials test reactor type arranged as an annular right circular cylinder between the inner and outer aluminum pressure vessels. Conservative engineering judgment resulted in the decision in 1988 to purchase new inner and outer pressure vessels. This paper details the difficulties encountered in procuring replacements for aluminum pressure vessels built to standards that are no longer applicable in attempting to meet nuclear standards that are not applicable to nonferrous material

A quantitative methodology for reactor vessel pressurized thermal shock decision making

International Nuclear Information System (INIS)

Ackerson, D.S.; Balkey, K.R.; Meyer, T.A.; Ofstun, R.P.; Rupprecht, S.D.; Sharp, D.R.

1983-01-01

The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). Previous reactor vessel integrity concerns have led to changes in vessel and plant system design and to operating procedures, and increased attention to the PTS issue is causing consideration of further modifications. Events such as excess feedwater, loss of normal feedwater, and steam generator tube rupture have led to significant primary system cooldowns. Each of these cooldown transients occurred concurrently with a relatively high primary system pressure. Considerations of these and other postulated cooldown events has drawn attention to the impact of operator action and control system effects on reactor vessel PTS. A methodology, which couples event sequence analysis with probabilistic fracture mechanics analyses, was developed to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation. (orig./RW)

The elevated temperature and thermal shock fracture toughnesses of nuclear pressure vessel steel

International Nuclear Information System (INIS)

Hirano, Kazumi; Kobayashi, Hideo; Nakazawa, Hajime; Nara, Atsushi.

1979-01-01

Thermal shock experiments were conducted on nuclear pressure vessel steel A533 Grade B Class 1. Elastic-plastic fracture toughness tests were carried out within the same high temperature range of the thermal shock experiment and the relation between stretched zone width, SZW and J-integral was clarified. An elastic-plastic thermal shock fracture toughness value. J sub(tsc) was evaluated from a critical value of stretched zone width, SZW sub(tsc) at the initiation of thermal shock fracture by using the relation between SZW and J. The J sub(tsc) value was compared with elastic-plastic fracture toughness values, J sub( ic), and the difference between the J sub(tsc) and J sub( ic) values was discussed. The results obtained are summarized as follows; (1) The relation between SZW and J before the initiation of stable crack growth in fracture toughness test at a high temperature can be expressed by the following equation regardless of test temperature, SZW = 95(J/E), where E is Young's modulus. (2) Elevated temperature fracture toughness values ranging from room temperature to 400 0 C are nearly constant regardless of test temperature. It is confirmed that upper shelf fracture toughness exists. (3) Thermal shock fracture toughness is smaller than elevated temperature fracture toughness within the same high temperature range of thermal shock experiment. (author)

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

International Nuclear Information System (INIS)

Loeb, Andreas; Stanke, Dieter; Thoma, Markus; Eisenmann, Beata; Prechtl, Erwin; Dehnke, Burckhard

2008-01-01

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.

Pressurized water reactor with a reactor pressure vessel

International Nuclear Information System (INIS)

Werres, L.

1979-01-01

The core barrel is suspended from a flange by means of a grid. The coolant enters the barrel from below through the grid. In order to get a uniform flow over the reactor core there is provided for a guiding device below the grid. It consists of a cylindrical shell with borings uniformly distributed around the shell as well as fins on the inner surface of the shell and slots at the bottom facing the pressure vessel. (GL) [de

40 CFR 35.533 - Programs eligible for inclusion.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 1 2010-07-01 2010-07-01 false Programs eligible for inclusion. 35.533... § 35.533 Programs eligible for inclusion. (a) Eligible programs. Except as provided in paragraph (b) of this section, the environmental programs eligible for inclusion in a Performance Partnership Grant are...

Transportable, small high-pressure preservation vessel for cells

International Nuclear Information System (INIS)

Kamimura, N; Sotome, S; Shimizu, A; Nakajima, K; Yoshimura, Y

2010-01-01

We have previously reported that the survival rate of astrocytes increases under high-pressure conditions at 4 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 0 C.

Probabilistic approach to the analysis of reactor pressure vessel integrity during a pressurized thermal shock

International Nuclear Information System (INIS)

Adamec, P.

2000-12-01

Following a general summary of the issue, an overview of international experience (USA; Belgium, France, Germany, Russia, Spain, Sweden, The Netherlands, and the UK; and probabilistic PTS assessment for the reactor pressure vessel at Loviisa-1, Finland) is presented, and the applicable computer codes (VISA-II, OCA-P, FAVOR, ZERBERUS) are highlighted and their applicability to VVER type reactor pressure vessels is outlined. (P.A.)

Neutron fluence determination for light water reactor pressure vessels

International Nuclear Information System (INIS)

Gold, R.

1994-01-01

A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced

Thermal annealing of an embrittled reactor pressure vessel

International Nuclear Information System (INIS)

Mager, T.R.; Dragunov, Y.G.; Leitz, C.

1998-01-01

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 11 deals with thermal annealing of an embrittled reactor pressure vessel. Anneal procedures for vessels from both the US and the former USSR are mentioned schematically, wet anneals at lower temperature and dry anneals above RPV design temperatures are investigated. It is shown that heat treatment is a means of recovering mechanical properties which were degraded by neutron radiation exposure, thus assuring reactor pressure vessel compliance with regulatory requirements

Apparatus for carrying out ultrasonic inspection of pressure vessels

International Nuclear Information System (INIS)

Dent, K.H.; Greenhalgh, F.G.

1975-01-01

An apparatus is described for moving an ultrasonic scanning mechanism over the interior surface of a pressure vessel and comprising a mast for supporting the scanning mechanism inside the vessel and a carriage for traversing the mast within the vessel, the mast being pivotably secured to the carriage so that when the ultrasonic scanning mechanism contacts the interior surface of the pressure vessel the mast is caused to pivot. (auth)

Pressure Tube and Pressure Vessel Reactors; certain comparisons

Energy Technology Data Exchange (ETDEWEB)

Margen, P H; Ahlstroem, P E; Pershagen, B

1961-04-15

In a comparison between pressure tube and pressure vessel type reactors for pressurized D{sub 2}O 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 D{sub 2}O 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.

Pressure Tube and Pressure Vessel Reactors; certain comparisons

International Nuclear Information System (INIS)

Margen, P.H.; Ahlstroem, P.E.; Pershagen, B.

1961-04-01

In a comparison between pressure tube and pressure vessel type reactors for pressurized D 2 O 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 D 2 O 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

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.

A mathematical model for pressure-based organs behaving as biological pressure vessels.

Science.gov (United States)

Casha, Aaron R; Camilleri, Liberato; Gauci, Marilyn; Gatt, Ruben; Sladden, David; Chetcuti, Stanley; Grima, Joseph N

2018-04-26

We introduce a mathematical model that describes the allometry of physical characteristics of hollow organs behaving as pressure vessels based on the physics of ideal pressure vessels. The model was validated by studying parameters such as body and organ mass, systolic and diastolic pressures, internal and external dimensions, pressurization energy and organ energy output measurements of pressure-based organs in a wide range of mammals and birds. Seven rules were derived that govern amongst others, lack of size efficiency on scaling to larger organ sizes, matching organ size in the same species, equal relative efficiency in pressurization energy across species and direct size matching between organ mass and mass of contents. The lung, heart and bladder follow these predicted theoretical relationships with a similar relative efficiency across various mammalian and avian species; an exception is cardiac output in mammals with a mass exceeding 10kg. This may limit massive body size in mammals, breaking Cope's rule that populations evolve to increase in body size over time. Such a limit was not found in large flightless birds exceeding 100kg, leading to speculation about unlimited dinosaur size should dinosaurs carry avian-like cardiac characteristics. Copyright © 2018. Published by Elsevier Ltd.

A prototype knowledge based system for pressure vessel design

Energy Technology Data Exchange (ETDEWEB)

Gunnarsson, L.

1991-11-22

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).

A prototype knowledge based system for pressure vessel design

Energy Technology Data Exchange (ETDEWEB)

Gunnarsson, L.

1991-11-22

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).

A prototype knowledge based system for pressure vessel design

International Nuclear Information System (INIS)

Gunnarsson, L.

1991-01-01

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)

Adjustable guide for a testing system for reactor pressure vessels

International Nuclear Information System (INIS)

Seifert, W.

1980-01-01

The device consisting of a guide rail and a manipulator is introduced into the gap between pressure vessel wall and biological shield by means of suspending wire drums and manipulator drums. For adjustment of the device an elbow telescope is used. The guide rail is fixed to the pressure vessel wall by means of electromagnets. The movements of the manipulator with respect to the guide rail are performed with the aid of a motor. (DG) [de

Dismantling method for reactor pressure vessel and system therefor

International Nuclear Information System (INIS)

Hayashi, Makoto; Enomoto, Kunio; Kurosawa, Koichi; Saito, Hideyo.

1994-01-01

Upon dismantling of a reactor pressure vessel, a containment building made of concretes is disposed underground and a spent pressure vessel is contained therein, and incore structures are contained in the spent pressure vessel. Further, a plasma-welder and a pressing machine are disposed to a pool for provisionally placing reactor equipments in the reactor building for devoluming the incore structures by welding and compression. An overhead-running crane and rails therefor are disposed on the roof and the outer side of the reactor building for transporting the pressure vessel from the reactor building to the containment building. They may be contained in the containment building after incorporation of the incore structures into the pressure vessel at the outside of the reactor building. For the devoluming treatment, a combination of cutting, welding, pressing and the like are optically conducted. A nuclear power plant can be installed by using a newly manufactured nuclear reactor, with no requirement for a new site and it is unnecessary to provide a new radioactive waste containing facility. (N.H.)

Stress analysis of pressure vessels

International Nuclear Information System (INIS)

Kim, B.K.; Song, D.H.; Son, K.H.; Kim, K.S.; Park, K.B.; Song, H.K.; So, J.Y.

1979-01-01

This interim report contains the results of the effort to establish the stress report preparation capability under the research project ''Stress analysis of pressure vessels.'' 1978 was the first year in this effort to lay the foundation through the acquisition of SAP V structural analysis code and a graphic terminal system for improved efficiency of using such code. Software programming work was developed in pre- and post processing, such as graphic presentation of input FEM mesh geometry and output deformation or mode shope patterns, which was proven to be useful when using the FEM computer code. Also, a scheme to apply fracture mechanics concept was developed in fatigue analysis of pressure vessels. (author)

Proof pressure tests of the PCPVs at Hinkley Point B and Hunterston B

International Nuclear Information System (INIS)

Eadie, D.McD.; Bell, D.J.

1976-01-01

The two PCPVs at Hinkley Point B were pressure tested in August 1973 and April 1974. The first vessel at Hunterston B was tested in December, 1973, and the second early in 1975. The vessels were pressurised up to 709 psig and, at various stages of pressurisation, readings were taken of external deflections, internal corner deflections and concrete strains. Surveys were taken of external concrete cracks and of crack gauges embedded in the concrete near the re-entrant corners. The external vessel deflections were measured optically using telescopes, targets and invar tapes. In some cases a recent design of manometric equipment was used to monitor the vertical deflections of the top slab during pressurisation and at proof pressure when access to the vessel was not possible for safety reasons. Internal concrete strains were measured using vibrating wire strain gauges. A brief description is given of the various measuring devices used. Deflection readings were also taken of some penetration primary closures. Summaries of the various recorded readings are given and compared with the design analyses. (author)

Residual Stress Estimation and Fatigue Life Prediction of an Autofrettaged Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Song, Kyung Jin; Kim, Eun Kyum; Koh, Seung Kee [Kunsan Nat’l Univ., Kunsan (Korea, Republic of)

2017-09-15

Fatigue failure of an autofrettaged pressure vessel with a groove at the outside surface occurs owing to the fatigue crack initiation and propagation at the groove root. In order to predict the fatigue life of the autofrettaged pressure vessel, residual stresses in the autofrettaged pressure vessel were evaluated using the finite element method, and the fatigue properties of the pressure vessel steel were obtained from the fatigue tests. Fatigue life of a pressure vessel obtained through summation of the crack initiation and propagation lives was calculated to be 2,598 cycles for an 80% autofrettaged pressure vessel subjected to a pulsating internal pressure of 424 MPa.

A framework expert system for pressure vessels

International Nuclear Information System (INIS)

Wang, Y.C.; Qin, S.J.

1989-01-01

Expert systems, known as a powerful tool to those numerical problems accompanied with logical argumentation, are facing the era of extended application into the engineering fields beyond the classical scopes of diagnosis and consultation. With regard to pressure vessels design it seems that the most important task is to establish a general purpose frame based on a microcomputer skeleton system to meet the various requirements of different vessels. The authors have made an attempt to perform such a skeleton designated file, ESTOOL, in order to achieve the objectives of executing numerical calculation combined with logical reasoning, and attaining higher efficiency of rules searching process. It has been successfully patched to the design software package for jacketed vessel with stirring shaft. This paper presents the guiding concepts and basic structure of ESTOOL via knowledge acquisition subsystem and inference engine

Strength-toughness requirements for thick walled high pressure vessels

International Nuclear Information System (INIS)

Kapp, J.A.

1990-01-01

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

Expanded Fermilab pressure vessel directory program

Energy Technology Data Exchange (ETDEWEB)

Tanner, A.

1983-01-01

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

Expanded Fermilab pressure vessel directory program

International Nuclear Information System (INIS)

Tanner, A.

1983-01-01

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

Cylindrical prestressed concrete pressure vessel for a nuclear power plant

International Nuclear Information System (INIS)

Horner, M.; Hodzic, A.; Haferkamp, D.

1976-01-01

A prestressed concrete pressure vessel for a HTGR is proposed which encloses, in addition to the reactor core, not only the heat-exchanging facilities but also the turbine unit. The reinforcement of the cylindrical concrete body is to be carried out with special care, it is provided for horizontal tendons, the prestressed concrete pressure vessel has a wire-winding device, while the longitudinal reinforcement is achieved by tendous guided in parallel to the vesses axes through the interspaces between the pods. (UWI) [de

Ductile fracture toughness of modified A 302 grade B plate materials. Volume 2

International Nuclear Information System (INIS)

McCabe, D.E.; Manneschmidt, E.T.; Swain, R.L.

1997-02-01

The objective of this work was to develop ductile fracture toughness data in the form of J-R curves for modified A 302 grade B plate materials typical of those used in fabricating reactor pressure vessels. A previous experimental study at Materials Engineering Associates (MEA) on one particular heat of A 302 grade B plate showed decreasing J-R curves with increased specimen thickness. This characteristic has not been observed in numerous tests made on the more recent production materials of A 533 grade B and A 508 class 2 pressure vessel steels. It was unknown if the departure from norm for the MEA material was a generic characteristic for all heats of A 302 grade B steels or just unique to that one particular plate. Seven heats of modified A 302 grade B steel and one heat of vintage A 533 grade B steel were provided to this project by the General Electric Company of San Jose, California. All plates were tested for chemical content, tensile properties, Charpy transition temperature curves, drop-weight nil-ductility transition (NDT) temperature, and J-R curves. Tensile tests were made in the three principal orientations and at four temperatures, ranging from room temperature to 550 degrees F (288 degrees C). Charpy V-notch transition temperature curves were obtained in longitudinal, transverse, and short transverse orientations. J-R curves were made using four specimen sizes (1/2T, IT, 2T, and 4T). None of the seven heats of modified A 302 grade showed size effects of any consequence on the J-R curve behavior. Crack orientation effects were present, but none were severe enough to be reported as atypical. A test temperature increase from 180 to 550 degrees F (82 to 288 degrees C) produced the usual loss in J-R curve fracture toughness. Generic J-R curves and mathematical curve fits to the same were generated to represent each heat of material. This volume is a compilation of all data developed

Individual Pressure Vessel (PV) and Common Pressure Vessel (CPV) Nickel-Hydrogen Battery Performance Under LEO Cycling Conditions

Science.gov (United States)

Miller, Thomas B.; Lewis, Harlan L.

2004-01-01

LEO life cycle testing of Individual Pressure Vessel (PV) and Common Pressure Vessel (CPV) nickel-hydrogen cell packs have been sponsored by the NASA Aerospace Flight Battery Program. The cell packs have cycled under both 35% and 60% depth-of- discharge and temperature conditions of -5 C and +lO C. The packs have been on test since as early as 1992 and have generated a substantial database. This report will provide insight into performance trends as a function of the specific cell configuration and manufacturer for eight separate nickel-hydrogen battery cell packs.

Method for the construction of a nuclear reactor with a prestressed concrete pressure vessel

International Nuclear Information System (INIS)

Schoening, J.; Schwiers, H.G.

1981-01-01

Method for the construction of nuclear reactors with prestressed concrete pressure vessel, providing during the initial stage of construction of the prestressed concrete pressure vessel a support structure around the liner. This enables an early mounting of core components in clean conditions as well as load reductions for final concreting in layers of the prestressed concrete pressure vessel. By applying the support structure, the overall assembly time of these nuclear power plant is considerably reduced without extra cost. (orig.) [de

46 CFR 97.30-1 - Repairs to boilers and pressure vessels.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Repairs to boilers and pressure vessels. 97.30-1 Section... VESSELS OPERATIONS Reports of Accidents, Repairs, and Unsafe Equipment § 97.30-1 Repairs to boilers and pressure vessels. (a) Before making any repairs to boilers or unfired pressure vessels, the chief engineer...

46 CFR 196.30-1 - Repairs to boilers and pressure vessels.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Repairs to boilers and pressure vessels. 196.30-1... VESSELS OPERATIONS Reports of Accidents, Repairs, and Unsafe Equipment § 196.30-1 Repairs to boilers and pressure vessels. (a) Before making any repairs to boilers or unfired pressure vessels, the Chief Engineer...

Ultrasound periodic inspections of reactor pressure vessels

International Nuclear Information System (INIS)

Haniger, L.

1980-01-01

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.)

Development of PWR pressure vessel steels

International Nuclear Information System (INIS)

Druce, S.; Edwards, B.

1982-01-01

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

Development of PWR pressure vessel steels

Energy Technology Data Exchange (ETDEWEB)

Druce, S.; Edwards, B.

1982-01-01

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

In-service ultrasonic inspection of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Prepechal, J.; Sulc, J.

1982-01-01

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.)

Prestressed concrete pressure vessels for nuclear reactors - 1973

Energy Technology Data Exchange (ETDEWEB)

1977-01-01

This standard deals with the design, construction, inspection and testing of prestressed concrete pressure vessels for nuclear reactors. Such pressure vessels serve the dual purpose of shielding and containing gas cooled nuclear reactors and are a form of civil engineering structure requiring particularly high integrity, and ensured leak tightness. (Metric)

Fracture dynamics of a propagating crack in a pressurized ductile cylinder

International Nuclear Information System (INIS)

Emery, A.F.; Love, W.J.; Kobayashi, A.S.

1977-01-01

A suddenly-introduced axial through-crack in the wall of a pipe pressurized by hot water is allowed to propagate according to Weiss' notch-strength theory of ductile static fracture. For this somewhat ductile material of A533B steel, Weiss' criterion was extended of dynamic fracture without modification. This dynamic-fracture criterion enabled a unique comparison to be obtained for the results of ductile-fracture with those of brittle-fracture in a fracturing A533B steel pipe. Since the pipe cross-sectional area is likely to increase with large flap motions under ductile tearing, a large deformation-shell-finite-difference-dynamic-code which includes rotary inertia was used in this analysis. The uniaxial-stress-strain curve of A533B steel was approximated by a bilinear stress-strain where Von-Misses yield criterion and associated flow rule were used in the elastic-plastic analysis. The fluid pressure was assumed constant and thus pipe flaps are only lightly loaded by pressure in this analysis. In previous publications, the authors have compared their preliminary results for the shell motion obtained through their model for a fracturing pipe with those of Kanninen, et al., and Freund, et al., to evaluate the effects of pressure loading on the crack flaps and the differences between small and large deflection results. In this paper, the differences in crack-propagation behavior of a fracturing pipe composed of the same A533B but subjected to a brittle or a ductile-fracture criterion are discussed. An important conclusion in fracture dynamics derived from analyses is that a smoothly-varying crack velocity will require a non-unique crack-velocity-versus-dynamic-fracture-parameter-relation while a unique and smoothly-varying crack-velocity-versus-dynamic-fracture-parameter-relation will demand an intermittently-propagating crack

PWR pressure vessel integrity during overcooling accidents

International Nuclear Information System (INIS)

Cheverton, R.D.

1981-01-01

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

Pressurized Vessel Slurry Pumping

International Nuclear Information System (INIS)

Pound, C.R.

2001-01-01

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

Acoustic emission monitoring during hydrotests of a thin wall pressure vessel

International Nuclear Information System (INIS)

Fontana, E.; Grugni, G.; Panzani, C.; Pirovano, B.; Possa, G.; Tonolini, F.

1975-01-01

The results are presented of an acoustic emission monitoring performed during hydrotests of a thin wall steel pressure vessel. The location of acoustic sources was based on longitudinal wave front detection. The careful calibration of the three sensors instrumentation system used for acoustic source location was found to be useful, and alllowed an accurate location error analysis. Acoustic emission in the hydrotests was found to be mainly due to stress release in weld seams. (Fontana, E.; Grugni, G.; Panzani, C.; Pirovano, B.; Possa, G.; Tonolini, F.)

Nuclear reactor pressure vessel-specific flaw distribution development

International Nuclear Information System (INIS)

Rosinski, S.T.

1992-01-01

Vessel integrity predictions performed through fracture mechanics analysis of a pressurized thermal shock event have been shown to be significantly sensitive to the overall flaw distribution input. It has also been shown that modem vessel in-service inspection (ISI) results can be used for development of vessel flaw distribution(s) that are more representative of US vessels. This paper describes the development and application of a methodology to analyze ISI data for the purpose of flaw distribution determination. The resultant methodology considers detection reliability, flaw sizing accuracy, and flaw detection threshold in its application. Application of the methodology was then demonstrated using four recently acquired US PWR vessel inspection data sets. Throughout the program, new insight was obtained into several key inspection performance and vessel integrity prediction practice issues that will impact future vessel integrity evaluation. For example, the potential application of a vessel-specific flaw distribution now provides at least one method by which a vessel-specific reference flaw size applicable to pressure-temperature limit curves determination can be estimated. This paper will discuss the development and application of the methodology and the impact to future vessel integrity analyses

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

International Nuclear Information System (INIS)

Kasmuri, N.H.; Md Som, A.

2006-01-01

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)

Common-Pressure-Vessel Nickel-Hydrogen Battery Development

OpenAIRE

Otzinger, Burton; Wheeler, James

1991-01-01

The dual-cell, common-pressure vessel, nickel-hydrogen configuration has recently emerged as an option for small satellite nickel-hydrogen battery application. An important incentive is that the dual-cell, CPV configured battery presents a 30 percent reduction in volume and nearly 50 percent reduction in mounting footprint, when compared with an equivalent battery of individual pressure- vessel (IPV) cells. In addition energy density and cost benefits are significant. Eagle-Picher Industries ...

Analysis code for pressure in reactor containment vessel of ATR. CONPOL

International Nuclear Information System (INIS)

1997-08-01

For the evaluation of the pressure and temperature in containment vessels in the events which are classified in the abnormal change of pressure, atmosphere and others in reactor containment vessels in accident among the safety evaluation events of the ATR, the analysis code for the pressure in reactor containment vessels CONPOL is used. In this report, the functions of the analysis code and the analysis model are shown. By using this analysis code, the rise of the pressure and temperature in a containment vessel is evaluated when loss of coolant accident occurs, and high temperature, high pressure coolant flows into it. This code possesses the functions of computing blow-down quantity and heat dissipation from reactor cooling facility, steam condensing heat transfer to containment vessel walls, and the cooling effect by containment vessel spray system. As for the analysis techniques, the models of reactor cooling system, containment vessel and steam discharge pool, and the computation models for the pressure and temperature in containment vessels, wall surface temperature, condensing heat transfer, spray condensation and blow-down are explained. The experimental analysis of the evaluation of the pressure and temperature in containment vessels at the time of loss of coolant accident is reported. (K.I.)

Guiding device for a manipulator mast for internal inspection of a reactor pressure vessel

International Nuclear Information System (INIS)

Seifert, W.; Schlueter, H.

1977-01-01

A remote-controlled supporting device centering a manipulator mast is described which is mounted and operated above a reactor pressure vessel under water in such a way that rotations and vertical movements necessary for the internal inspection of the pressure vessel remain possible. (RW) [de

GOLD PRESSURE VESSEL SEAL

Science.gov (United States)

Smith, A.E.

1963-11-26

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

Pressurized water reactor with reactor pressure vessel

International Nuclear Information System (INIS)

Werres, L.

1985-01-01

The pressure vessel has a cylindrical jacket with a domed floor. A guide is arranged on the domed floor to even out the flow in the core. It consists of a cylindrical jacket, whose lower end has slots and fins. These fins are welded to the domed floor. (orig./PW)

Pressurized water reactor with reactor pressure vessel

International Nuclear Information System (INIS)

Werres, L.

1980-01-01

The pressure vessel has a cylindrical jacket with a domed floor. A guide is arranged on the domed floor to even out the flow in the core. It consists of a cylindrical jacket, whose lower end has slots and fins. These fins are welded to the domed floor. (DG) [de

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

International Nuclear Information System (INIS)

Chou, Hsoung-Wei; Huang, Chin-Cheng

2016-01-01

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

Prestressed concrete pressure vessels for boiling water reactors

International Nuclear Information System (INIS)

Menon, S.

1979-12-01

Following a general description of the Scandinavian cooperative project on prestressed concrete pressure vessels for boiling water reactors, detailed discussion is given in four appendices of the following aspects: the verification programme of tests and studies, the development and testing of a liner venting system, a preliminary safety philosophy and comparative assessment of cold and hot liners. Vessel failure probability is briefly discussed and some figures presented. The pressure gradients in the vessel wall resulting from various stipulated linear cracks, with a liner venting system are presented graphically. (JIW)

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.

A model for structural analysis of nuclear reactor pressure vessel flanges

International Nuclear Information System (INIS)

Oliveira, C.A. de.

1987-01-01

Due to the recent Brazilian advances in the nuclear technology area, it has been necessary the development of design and analysis methods for pressurized water reactor components, also as other components of a nuclear plant. This work proposes a methodology for the structural analysis of large diameter nuclear reactor pressure vessel flanges. In the analysis the vessel is divided into shell-of-revolution elements, the flanges are represented by rigid rings, and the bolts are treated as beams. The flexibility method is used for solving the problem. A computer program is shown, and the given results (displacements and stresses) are compared with results obtained by the finite element method. Although developed for nuclear reactor pressure vessel calculations, the program is more general, being possible its use for the analysis of any structure composed by shells of revolution. (author)

Radiation embrittlement of Spanish nuclear reactor pressure vessel steels

International Nuclear Information System (INIS)

Bros, J.; Ballesteros, A.; Lopez, A.

1993-01-01

Commercial pressurized water reactor (PWR) and boiling water reactor (BWR) nuclear power plants contain a series of pressure vessel steel surveillance capsules as the principal means of monitoring radiation effects on the pressure vessel. Changes in fracture toughness are more severe in surveillance capsules than in reactor vessel materials because of their proximity of the reactor core. Therefore, it is possible to predict changes in fracture toughness of the reactor vessel materials. This paper describes the status of the reactor vessel surveillance program relating to Spanish nuclear power plants. To date, twelve capsules have been removed and analyzed from seven of the nine Spanish reactors in operation. The results obtained from the analysis of these capsules are compared with the predictions of the Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Rev. 2, by means of measured and expected increase of the nil-ductility transition reference temperature (RT NDT ). The comparison is made considering the different variables normally included in the studies of radiation response of reactor pressure vessel materials, such as copper content of steel, level of neutron fluence above 1 MeV, base metal or weld metal, and so forth. The surveillance data have been used for determining the adjusted reference temperatures and upper shelf energies at any time. The results have shown that the seven pressure vessels are in excellent condition to continue operating with safety against brittle fracture beyond the design life, without the need to recuperate the degraded properties of the materials by annealing of the vessel

Silencing of Atp2b1 increases blood pressure through vasoconstriction.

Science.gov (United States)

Shin, Young-Bin; Lim, Ji Eun; Ji, Su-Min; Lee, Hyeon-Ju; Park, So-Yon; Hong, Kyung-Won; Lim, Mihwa; McCarthy, Mark I; Lee, Young-Ho; Oh, Bermseok

2013-08-01

Recent genome-wide association studies (GWASs) have identified 30 genetic loci that regulate blood pressure, increasing our understanding of the cause of hypertension. However, it has been difficult to define the causative genes at these loci due to a lack of functional analyses. In this study, we aimed to validate the candidate gene ATP2B1 in 12q21, variants near which have the strongest association with blood pressure in Asians and Europeans. ATP2B1 functions as a calcium pump to fine-tune calcium concentrations - necessary for repolarization following muscular contractions. We silenced Atp2b1 using an siRNA complex, injected into mouse tail veins. In treated mice, blood pressure rose and the mesenteric arteries increased in wall : lumen ratio. Moreover, the arteries showed enhanced myogenic responses to pressure, and contractile responses to phenylephrine increased compared with the control, suggesting that blood pressure is regulated by ATP2B1 through the contraction and dilation of the vessel, likely by controlling calcium concentrations in the resting state. These results support that ATP2B1 is the causative gene in the blood pressure-associated 12q21 locus and demonstrate that ATP2B1 expression in the vessel influences blood pressure.

Examination of VVER-1000 Reactor Pressure Vessel

International Nuclear Information System (INIS)

Matokovic, A.; Picek, E.; Markulin, K.

2008-01-01

The increasing demand of a higher level of safety in the operation of the nuclear power plants requires the utilisation of more precise automated equipment to perform in-service inspections. That has been achieved by technological advances in computer technology, in robotics, in examination probe technology with the development of the advanced inspection technique and has also been due to the considerable and varied experience gained in the performance of such inspections. In-service inspection of reactor pressure vessel, especially Russian-designed WWER-1000 presents one of the most important and extensive examination of nuclear power plants primary circuit components. Such examination demand high standards of inspection technology, quality and continual innovation in the field of non-destructive testing advanced technology. A remote underwater contact ultrasonic technique is employed for the examination of the base metal of vessel and reactor welds, whence eddy current method is applied for clad surface examinations. Visual testing is used for examination of the vessel interior. The movement of inspection probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with inspection systems. The successful performance of reactor pressure vessel is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen non-destructive techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state-of-the-art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. This paper presents advanced approach in the reactor pressure vessel in-service inspections and it is especially developed for WWER-1000 nuclear power plants.(author)

Eddy current testing of composite pressure vessels

Science.gov (United States)

Casperson, R.; Pohl, R.; Munzke, D.; Becker, B.; Pelkner, M.

2018-04-01

The use of composite pressure vessels instead of conventional vessels made of steel or aluminum grew strongly over the last decade. The reason for this trend is the tremendous weight saving in the case of composite vessels. However, the long-time behavior is not fully understood for filling and discharging cycles and creep strength and their influence on the CFRP coating (carbon fiber reinforced plastics) and the internal liner (steel, aluminum, or plastics). The CFRP ensures the pressure resistance while the inner liner is used as a container for liquid or gas. To overcome the missing knowledge of aging, BAM started an internal project to investigate degradation of these material systems. Therefore, applicable testing methods like eddy current testing are needed. Normally, high-frequency eddy current testing (HF-ET, f > 10 MHz) is deployed for CFRP due to its low conductivity of the fiber, which is in the order of 0.01 MS/s, and the capacitive coupling between the fibers. Nevertheless, in some cases conventional ET can be applied. We show a concise summary of studies on the application of conventional ET of composite pressure vessels.

Mechanical Behavior of A Metal Composite Vessels Under Pressure At Cryogenic Temperatures

Science.gov (United States)

Tsaplin, A. I.; Bochkarev, S. V.

2016-01-01

Results of an experimental investigation into the deformation and destruction of a metal composite vessel with a cryogenic gas are presented. Its structure is based on basalt, carbon, and organic fibers. The vessel proved to be serviceable at cryogenic temperatures up to a burst pressure of 45 MPa, and its destruction was without fragmentation. A mathematical model adequately describing the rise of pressure in the cryogenic vessel due to the formation of a gaseous phase upon boiling of the liquefied natural gas during its storage without drainage at the initial stage is proposed.

Internal Friction of Pressure Vessel Steel Embrittlement

International Nuclear Information System (INIS)

Van Ouytsel, K.

2001-01-01

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

Investigation of a weld defect, reactor vessel head Ringhals 2

International Nuclear Information System (INIS)

Embring, G.; Pers-Anderson, E.B.

1994-01-01

During the summer-outage 1993 Ringhals unit 2 vessel head was inspected at weld-area of Alloy 182. One major defect was found Two plus two ''boat-samples'' were taken out from the zone between the weld and the stainless cladding. All samples were sent to Studsviks laboratories for detailed investigations. The metallographic and fractographic investigations revealed that the major weld-defect had been there from manufacturing. The defect was located between the Alloy 182-buttering and the pressure vessel steel SA 533 grB cl 1. No indications of PWSCC or IDSCC were found. An inspection programme was defined. Different types of reference blocks were provided by Ringhals in cooperation with ABB TRC. Reference reflectors of type flat bottom hole (FBH) and eroded notches (EDM), with different sizes and separation were manufactured. One weld sample with manufacturing defects -lack of fusion and slag was inclusions- was present. ABB TRC performed UT inspection in the gap between the penetration and the thermal sleeve. Inspection results like defect identification, defect separation and defect sizing accuracy were compared with result from the destructive inspection. No relevant additional defects were found. An analysing and repair program was performed. A special designed disc sealed off the defect area. (authors). 5 figs., 3 refs

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.

Reactor Structural Materials: Reactor Pressure Vessel Steels

International Nuclear Information System (INIS)

Chaouadi, R.

2000-01-01

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

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

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

Science.gov (United States)

2010-10-01

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

The influence of chemistry concentration on the fracture risk of a reactor pressure vessel subjected to pressurized thermal shocks

Energy Technology Data Exchange (ETDEWEB)

Huang, Pin-Chiun [Institute of Nuclear Engineering and Science, National Tsing-Hua University, Hsinchu 30013, Taiwan, ROC (China); Chou, Hsoung-Wei, E-mail: hwchou@iner.gov.tw [Institute of Nuclear Energy Research, Taoyuan 32546, Taiwan, ROC (China); Ferng, Yuh-Ming [Institute of Nuclear Engineering and Science, National Tsing-Hua University, Hsinchu 30013, Taiwan, ROC (China)

2016-02-15

Highlights: • Probabilistic fracture mechanics method was used to analyze a reactor pressure vessel. • Effects of copper and nickel contents on RPV fracture probability under PTS were investigated and discussed. • Representative PTS transients of Beaver Valley nuclear power plant were utilized. • The range of copper and nickel contents of the RPV materials were suggested. • With different embrittlement levels the dominated PTS category is different. - Abstract: The radiation embrittlement behavior of reactor pressure vessel shell is influenced by the chemistry concentration of metal materials. This paper aims to study the effects of copper and nickel content variations on the fracture risk of pressurized water reactor (PWR) pressure vessel subjected to pressurized thermal shock (PTS) transients. The probabilistic fracture mechanics (PFM) code, FAVOR, which was developed by the Oak Ridge National Laboratory in the United States, is employed to perform the analyses. A Taiwan domestic PWR pressure vessel assumed with varied copper and nickel contents of beltline region welds and plates is investigated in the study. Some PTS transients analyzed from Beaver Valley Unit 1 for establishing the U.S. NRC's new PTS rule are applied as the loading condition. It is found that the content variation of copper and nickel will significantly affect the radiation embrittlement and the fracture probability of PWR pressure vessels. The results can be regarded as the risk incremental factors for comparison with the safety regulation requirements on vessel degradation as well as a reference for the operation of PWR plants in Taiwan.

The need to pressure test prestressed concrete reactor vessels

International Nuclear Information System (INIS)

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

1983-01-01

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

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

International Nuclear Information System (INIS)

Nisar, J.A.; Abdullah, A.N.; Iqbal, N.

2004-01-01

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)

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

International Nuclear Information System (INIS)

Chatterjee, S.; Shah, Priti Kotak

2008-05-01

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. RT NDT 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)

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)

Moranchel y R, M.

2012-01-01

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

ITER cryostat main chamber and vacuum vessel pressure suppression system design

International Nuclear Information System (INIS)

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

1999-03-01

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

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)

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

International Nuclear Information System (INIS)

Frid, W.

1988-04-01

Discharge of the molten core debris from a pressurized reactor vessel has been recognized as an important accident scenario for pressurized water reactors. Recent high-pressure melt streaming experiments conducted at Sandia National Laboratories, designed to study cavity and containment events related to melt ejection, have resulted in two important observations: (1) Expansion and breakup of the ejected molten jet. (2) Significant aerosol generation during the ejection process. The expansion and breakup of the jet in the experiments are attributed to rapid evolution of the pressurizing gas (nitrogen or hydrogen) dissolved in the melt. It has been concluded that aerosol particles may be formed by condensation of melt vapor and mechanical breakup of the melt and generation. It was also shown that the above stated phenomena are likely to occur in reactor accidents. 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 that are addressed in this work

Steels and welding nuclear

International Nuclear Information System (INIS)

Sessa, M.; Milella, P.P.

1987-01-01

This ENEA Data-Base regards mechanical properties, chemical composition and heat treatments of nuclear pressure vessel materials: type A533-B, A302-B, A508 steel plates and forgings, submerged arc welds and HAZ before and after nuclear irradiation. Irradiation experiments were generally performed in high flux material test reactors. Data were collected from international available literature about water nuclear reactors pressure vessel materials embrittlement

Leak detector for reactor pressure vessel

International Nuclear Information System (INIS)

Morimoto, Mikio.

1991-01-01

A branched pipe is disposed to a leak off pipeline led from a flange surface which connects the main body and the upper lid of a reactor pressure vessel. An exhaust pump is disposed to the branched pipe and a moisture gage is disposed on the side of the exhaustion and a dry air supplier is connected to the branched pipe. Upon conducting a pressure-proof leak test for the reactor pressure vessel, the exhaust pump is operated and an electromagnet valve disposed at the upstream of the dry air supplier is opened and closed repeatedly. The humidity of air sucked by the exhaust pump is detected by the moisture gage. If leaks should be caused in the joining surface of the flange, leaked water is diffused as steams. Accordingly, occurrence of leak can be detected instantly based on the comparison with the moisture level of the dry air as a standard. In this way, a leak test can be conducted reliably in a short period of time with no change of for the reactor pressure container itself. (I.N.)

Concept of a Prestressed Cast Iron Pressure Vessel for a Modular High Temperature Reactor

International Nuclear Information System (INIS)

Steinwarz, Wolfgang; Bounin, Dieter

2014-01-01

High Temperature Reactors (HTR) are representing one of the most interesting solutions for the upcoming generation of nuclear technology, especially with view to their inherent safety characteristics. To complete the safety concept of such plants already in the first phase of the technical development, Prestressed Cast Iron Pressure Vessels (PCIV) instead of the established forged steel reactor pressure vessels have been considered under the aspect of safety against bursting. A longterm research and development work, mainly performed in Germany, showed the excellent features of this technical solution. Diverse prototypic vessels were tested and officially proven. Design studies confirmed the feasibility of such a vessel concept also for Light Water Reactor types, too. The main concept elements of such a burst-proof vessel are: Strength and tightness functions are structurally separated. The tensile forces are carried by the prestressing systems consisting of a large number of independent wires. Compressive forces are applied to the vessel walls and heads. These are segmented into blocks of ductile cast iron. All cast iron blocks are prestressed to high levels of compression. The sealing function is assigned to a steel liner fixed to the cast iron blocks. The prestressing system is designed for an ultimate pressure of 2.3 times the design pressure. The prestress of the lids is designed for gapping at a much smaller pressure. Therefore, a drop of pressure will always occur before loss of strength (“leakage before failure”). In addition to these safety features further technical as well as economic aspects generate favorable assessment criteria: high design flexibility, feasibility of large vessel diameters; advantageous conditions for transport, assembly and decommissioning due to the segmented construction; advantage of workshop manufacturing; high-level quality control of components. Nowadays, considering the globally newly standardized safety requirements

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

Energy Technology Data Exchange (ETDEWEB)

Grounes, M

1966-03-15

, the new steels mentioned above, and the U.S. steels ASTM A212B and ASTM A302B, are given. Results from a series of irradiations at temperatures between 60 deg C and 300 deg C to doses between 2x10{sup 18} and 3x10{sup 19} n/cm{sup 2} (> 1 MeV) show that a saturation effect is obtained at higher temperatures in some, but not all, steels. This means that, after some irradiation, further irradiation causes no further change in the properties. The experiments show that similar steels, and even different heats of the same steel, may behave in an entirely different manner as a result of irradiation. Experiments showing the effects of differences in neutron flux and neutron spectrum are described. Calculations show that the effect of differences in neutron spectra between materials testing reactors and pressure vessels may be considerable.

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

International Nuclear Information System (INIS)

Grounes, M.

1966-03-01

, the new steels mentioned above, and the U.S. steels ASTM A212B and ASTM A302B, are given. Results from a series of irradiations at temperatures between 60 deg C and 300 deg C to doses between 2x10 18 and 3x10 19 n/cm 2 (> 1 MeV) show that a saturation effect is obtained at higher temperatures in some, but not all, steels. This means that, after some irradiation, further irradiation causes no further change in the properties. The experiments show that similar steels, and even different heats of the same steel, may behave in an entirely different manner as a result of irradiation. Experiments showing the effects of differences in neutron flux and neutron spectrum are described. Calculations show that the effect of differences in neutron spectra between materials testing reactors and pressure vessels may be considerable

Factors affecting the integrity of PWR pressure vessels during overcooling accidents

International Nuclear Information System (INIS)

Cheverton, R.D.

1983-01-01

The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, if certain postulated accidents, referred to as overcooling accidents, were to occur, the pressure vessel could be subjected to severe thermal shock while the pressure is substantial. As a result, vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner-surface flaws prior to the vessel's normal end of life. A fracture-mechanics analysis for a typical postulated accident and also related thermal-shock experiments indicate that very shallow surface flaws that extend through the cladding into the base material could propagate. This is of particular concern because shallow flaws appear to be the most probable and presumably are the most difficult to detect

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

Science.gov (United States)

2010-10-01

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

Pressure vessel and method therefor

Science.gov (United States)

Saunders, Timothy

2017-09-05

A pressure vessel includes a pump having a passage that extends between an inlet and an outlet. A duct at the pump outlet includes at least one dimension that is adjustable to facilitate forming a dynamic seal that limits backflow of gas through the passage.

Evaluation of upper-shelf toughness requirements for reactor pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Gamble, R.M.; Zahoor, A. (NOVETECH Corp., Rockville, MD (USA)); Hiser, A. (Materials Engineering Associates, Inc., Lanham, MD (USA)); Ernst, H.A.; Pollitz, E.T. (Georgia Inst. of Tech., Atlanta, GA (USA))

1990-04-01

This work assesses and applies the criteria recommended by the ASME Subgroup on Evaluation Standards for the evaluation of reactor pressure vessel beltline materials having upper shelf Charpy energies less than 50 ft-lbs. The assessment included comparison of the upper shelf energies required by the criteria recommended for Service Level A and B conditions and criteria proposed for evaluation of postulated Service Level C and D events. The criteria recommended for Service Level A and B conditions was used to evaluate Linde 80 weld material. 9 refs., 4 figs.

Evaluation of upper-shelf toughness requirements for reactor pressure vessels

International Nuclear Information System (INIS)

Gamble, R.M.; Zahoor, A.; Hiser, A.; Ernst, H.A.; Pollitz, E.T.

1990-04-01

This work assesses and applies the criteria recommended by the ASME Subgroup on Evaluation Standards for the evaluation of reactor pressure vessel beltline materials having upper shelf Charpy energies less than 50 ft-lbs. The assessment included comparison of the upper shelf energies required by the criteria recommended for Service Level A and B conditions and criteria proposed for evaluation of postulated Service Level C and D events. The criteria recommended for Service Level A and B conditions was used to evaluate Linde 80 weld material. 9 refs., 4 figs

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 as...

Fabrication and mechanical test data for the four 6-inch-thick intermediate test vessels made from steel plate for the Heavy Section Steel Program

International Nuclear Information System (INIS)

Childress, C.E.

1976-01-01

The HSST Program has among its goals the objective of demonstrating the capability to predict safe behavior of thick-walled pressure vessels containing flaws of known dimensions under frangible, transitional, and tough loading regimes. To accomplish these objectives the program is conducting a series of tests involving 6-in.-thick pressure vessels which will serve as test specimens for assisting in the characterization of failure under these loading conditions. Among the vessels a number of parameters, such as weld type, weld location, flaw size and shape, and test temperature and pressure, will be selectively varied to show that a rationale exists for dealing with the varied stress and metallurgical states which normally exist in commercial nuclear reactor vessels. Each vessel will serve as a go, no-go determination of critical flaw size for a specific set of test parameters. Item 4 of the previous issues in this series covers the fabrication details of the first six 6-in.-thick test vessels, which were fabricated from ASTM A-508 Cl 2 forging materials. This report covers the fabrication details of four additional 6-in.-thick intermediate test vessels having shell courses fabricated from ASTM A-533 Gr B Cl 1 plate. The remaining components were made from forgings. Essentially this report is a continuation of ORNL-TM-4351; it describes the manufacturing details of the individual parts and their ultimate assembly into finished vessels. Details concerning chemical composition and mechanical and nondestructive test data are presented

A study on probabilistic fracture mechanics for nuclear pressure vessels and piping

International Nuclear Information System (INIS)

Yagawa, Genki; Yoshimura, Shinobu

1997-01-01

This paper describes some recent research activities on probabilistic fracture mechanics (PFM) for nuclear pressure vessels and piping (PV and P) performed by the RC111 research committee of the Japan Society of Mechanical Engineers (JSME) under a subcontract of the Japan Atomic Energy Research Institute (JAERI). To establish standard procedures for evaluating failure probabilities of nuclear PV and P, we have set up the following three kinds of PFM round-robin problems on: (a) primary piping under normal operating conditions, (b) aged reactor pressure vessel (RPV) under normal and upset operating conditions, and (c) aged RPV under pressurised thermal shock (PTS) events. The basic problems of the last one are chosen from some US benchmark problems such as EPRI (Electric Power Research Institute) and US NRC (Nuclear Regulatory Commission) joint PTS benchmark problems. This paper summarizes some sensitivity studies on the three kinds of problems mainly varying material properties such as flow stress, fracture toughness, fatigue crack growth rate, Cu content. Employed in this study are the PFM computer codes developed in Japan and USA. Failure probabilities of nuclear PV and P are quantitatively discussed in detail. (author)

A powerful methodology for reactor vessel pressurized thermal shock analysis

International Nuclear Information System (INIS)

Boucau, J.; Mager, T.

1994-01-01

The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). More specifically, the review of the old WWER-type of reactors (WWER 440/230) has indicated a sensitive behaviour to neutron embrittlement. This led already to some remedial actions including safety injection water preheating or vessel annealing. Such measures are usually taken based on the analysis of a selected number of conservative PTS events. Consideration of all postulated cooldown events would draw attention to the impact of operator action and control system effects on reactor vessel PTS. Westinghouse has developed a methodology which couples event sequence analysis with probabilistic fracture mechanics analyses, to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation. Once the event sequences of concern are identified, detailed deterministic thermal-hydraulic and structural evaluations can be performed to determine the conditions required to minimize the extension of postulated flaws or enhance flaw arrest in the reactor vessel. The results of these analyses can then be used to better define further modifications in vessel and plant system design and to operating procedures. The purpose of the present paper will be to describe this methodology and to show its benefits for decision making. (author). 1 ref., 3 figs

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

International Nuclear Information System (INIS)

Frid, W.

1987-01-01

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 10 8 per m 3 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)

Increase of cyclic durability of pressure vessels

International Nuclear Information System (INIS)

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

1980-01-01

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

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

Science.gov (United States)

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

2017-01-01

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

Fracture dynamics of a propagating crack in a pressurized ductile cylinder

International Nuclear Information System (INIS)

Emery, A.F.; Love, W.J.; Kobayashi, A.S.

1977-01-01

A suddenly-introduced axial through-crack in the wall of a pipe pressurized by hot water is allowed to propagate according to Weiss' notch-strength theory of ductile static fracture. The dynamic-fracture criterion used enabled the authors to obtain a unique comparison of the results of ductile-fracture with those of brittle-fracture in a fracturing A533B steel pipe. Since the pipe cross-sectional area is likely to increase with large flap motions under ductile tearing, a large deformation shell-finite-difference-dynamic-code which includes rotary inertia was used in this analysis. The uniaxial-stress-strain curve of A533B steel was approximated by a bilinear-stress-strain where Von-Mises yield criterion and associated flow rule were used in the elastic-plastic analysis. The fluid pressure was assumed constant and thus pipe flaps are only lightly loaded by pressure in this analysis. (Auth.)

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Rapid construction of concrete pressure vessels

International Nuclear Information System (INIS)

Limbert, D.; Weatherseed, D.C.

1989-01-01

This paper opens with a general description of the concrete pressure vessel followed by a more detailed examination of the critical elements of the construction, including choice of methods and plant which were selected to ensure its rapid construction. The pressure vessel construction cannot be treated in isolation, because it is very closely linked with its surrounding structures - namely the reactor hall which surrounds it and the charge hall which tops it, as will be seen in the context of this paper. Rate of progress of construction is not entirely in the civil contractor's hands because so many of the operations affecting the civil works are of a mechanical nature, hence a very close liaison and understanding amongst all contractors concerned was of the utmost importance. (author)

FE-simulation of the viscoplastic behaviour of different RPV steels in the frame of in-vessel melt retentions scenarios

International Nuclear Information System (INIS)

Altstadt, E.; Willschuetz, H.G.; Mueller, G.

2004-01-01

Assuming the hypothetical scenario of a severe accident with subsequent core meltdown and formation of a melt pool in the reactor pressure vessel (RPV) lower plenum of a Light Water Reactor (LWR) leads to the question about the behavior of the RPV. One accident management strategy could be to stabilize the in-vessel debris configuration in the RPV as one major barrier against uncontrolled release of heat and radio nuclides. To get an improved understanding and knowledge of the melt pool convection and the vessel creep and possible failure processes and modes occurring during the late phase of a core melt down accident the FOREVER-experiments (Failure Of REactor VEssel Retention) have been performed at the Division of Nuclear Power Safety of the Royal Institute of Technology Stockholm. These experiments are simulating the behavior of the lower head of the RPV under the thermal loads of a convecting melt pool with decay heating, and under the pressure loads that the vessel experiences in a depressurization scenario. The geometrical scale of the experiments is 1:10 compared to a common LWR. This paper deals with the experimental, numerical, and metallographical results of the creep failure experiment EC-FOREVER-4, where the American pressure vessel steel SA533B was applied for the lower head. For comparison the results of the experiment EC-FOREVER-3B, build of the French 16MND5 steel, are discussed, too. Emphasis is put on the differences in the viscoplastic behaviour of different heats of the RPV steel. For this purpose, the creep tests in the frame of the LHF/OLHF experiments are reviewed, too. As a hypothesis it is stated that the sulphur content could be responsible for differences in the creep behaviour. (orig.)

Possible research program on a large scale nuclear pressure vessel

International Nuclear Information System (INIS)

1983-01-01

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

Multiple cell common pressure vessel nickel hydrogen battery

Science.gov (United States)

Zagrodnik, Jeffrey P.; Jones, Kenneth R.

1991-01-01

A multiple cell common pressure vessel (CPV) nickel hydrogen battery was developed that offers significant weight, volume, cost, and interfacing advantages over the conventional individual pressure vessel (IPV) nickel hydrogen configuration that is currently used for aerospace applications. The baseline CPV design was successfully demonstrated though the testing of a 26 cell prototype, which completed over 7,000 44 percent depth of discharge LEO cycles. Two-cell boilerplate batteries have now exceeded 12,500 LEO cycles in ongoing laboratory tests. CPV batteries using both nominal 5 and 10 inch diameter vessels are currently available. The flexibility of the design allows these diameters to provide a broad capability for a variety of space applications.

Pressure vessel lid

International Nuclear Information System (INIS)

Schoening, J.; Elter, C.; Becker, G.; Pertiller, S.

1986-01-01

The invention concerns a lid for closing openings in reactor pressure vessels containing helium, which is made as a circular casting with hollow spaces and a flat floor and is set on the opening and kept down. It consists of helium-tight metal cast material with sufficient temperature resistance. There are at least two concentric heat resistant seals let into the bottom of the lid. The bottom is in immediate contact with the container atmosphere and has hollow spaces in its inside in the area opposite to the opening. (orig./HP) [de

SNPDL work in support of collaborative programmes to study fatigue crack propagation in light water reactor pressure vessel steels

International Nuclear Information System (INIS)

McMinn, A.

1981-07-01

Work performed at Springfields Nuclear Power Development Laboratories (SNL) in support of United Kingdom and international cooperative groups is described. For the UK collaborative group crack growth tests have been performed on A533-B pressure vessel steel in air at 1 Hz, and in a simulated PWR environment at 1 Hz and 0.0167 Hz, all tests being at a stress ratio (R) of 0.7. Tests were performed at SNL at ambient temperature and pressure. The results showed that enhancements in crack growth rates were obtained because of the aqueous environment and the lower cyclic frequency. Good inter-laboratory agreement was obtained for the air test, indicating that there was little variation in the mechanical control between laboratories. There was also good agreement between the results from laboratories where tests were performed under high temperature wet conditions at 1 Hz. SNL room temperature data demonstrated an effect of temperature, as the crack growth rates found were up to a factor of four lower than the high temperature data. Although crack growth could be maintained at 0.0167 Hz and low cyclic stress intensity levels at ambient temperature, this was not found to be possible by the laboratories performing the test at 288 0 C. The first international 'round robin' test at R = 0.2 and 0.0167 Hz (1 cycle/min) in an aqueous environment gave a significant amount of scatter in the results. (author)

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

International Nuclear Information System (INIS)

Dickson, T.L.

1993-01-01

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

Fracture behaviour assessment of a flawed pressure vessel in the hydro-test

Energy Technology Data Exchange (ETDEWEB)

Sarkimo, M; Rintamac, R

1988-12-31

This document deals with the fracture properties of a flawed pressure vessel. The experiment was carried out within the Nordic Countries on a vessel in a Finnish refinery. The instrumentation used included acoustic emission. Some results are provided. (TEC).

Dynamic fracture characterization of a pressure vessel steel

International Nuclear Information System (INIS)

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

1991-01-01

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

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

Science.gov (United States)

2010-10-01

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

Light Water Reactor-Pressure Vessel Surveillance project computer system

International Nuclear Information System (INIS)

Merriman, S.H.

1980-10-01

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

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

Science.gov (United States)

2010-10-01

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

Reactor pressure vessel steels

International Nuclear Information System (INIS)

Van De Velde, J.; Fabry, A.; Van Walle, E.; Chaouuadi, R.

1998-01-01

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

Flaw density examinations of a clad boiling water reactor pressure vessel segment

International Nuclear Information System (INIS)

Cook, K.V.; McClung, R.W.

1986-01-01

Flaw density is the greatest uncertainty involved in probabilistic analyses of reactor pressure vessel failure. As part of the Heavy-Section Steel Technology (HSST) Program, studies have been conducted to determine flaw density in a section of reactor pressure vessel cut from the Hope Creek Unit 2 vessel [nominally 0.7 by 3 m (2 by 10 ft)]. This section (removed from the scrapped vessel that was never in service) was evaluated nondestructively to determine the as-fabricated status. We had four primary objectives: (1) evaluate longitudinal and girth welds for flaws with manual ultrasonics, (2) evaluate the zone under the nominal 6.3-mm (0.25-in.) clad for cracking (again with manual ultrasonics), (3) evaluate the cladding for cracks with a high-sensitivity fluorescent penetrant method, and (4) determine the source of indications detected

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

International Nuclear Information System (INIS)

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

1977-01-01

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

Integrity of Magnox reactor steel pressure vessels

International Nuclear Information System (INIS)

Flewitt, P.E.J.; Williams, G.H.; Wright, M.B.

1992-01-01

The background to the safety assessment of the steel reactor pressure vessels for Magnox power stations is reviewed. The evolved philosophy adopted for the 1991 safety cases prepared for the continued operation of four Magnox power stations operated by Nuclear Electric plc is described, together with different aspects of the multi-legged integrity argument. The main revisions to the materials mechanical property data are addressed together with the assessment methodology adopted and their implications for the overall integrity argument formulated for the continued safe operation of these reactor pressure vessels. (author)

Pressurized-thermal-shock experiments with thick vessels

International Nuclear Information System (INIS)

Bryan, R.H.; Nanstad, R.K.; Merkle, J.G.; Robinson, G.C.; Whitman, G.D.

1986-01-01

Information is provided on the series of pressurized-thermal-shock experiments at the Oak Ridge National Laboratory, motivated by a concern for the behavior of flaws in reactor pressure vessels having welds or shells exhibiting low upper-shelf Charpy impact energies, approx. 68J or less

The Assembly and Test of Pressure Vessel for Irradiation

International Nuclear Information System (INIS)

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

2009-01-01

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

The Assembly and Test of Pressure Vessel for Irradiation

Energy Technology Data Exchange (ETDEWEB)

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

2009-02-15

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

Safety of steel vessel Magnox pressure circuits

International Nuclear Information System (INIS)

Stokoe, T.Y.; Bolton, C.J.; Heffer, P.J.H.

1991-01-01

The maintenance of pressure circuit integrity is fundamental to nuclear safety at the steel vessel Magnox stations. To confirm continued pressure circuit integrity the CEGB, as part of the Long Term Safety Review, has carried out extensive assessment and inspection in recent years. The assessment methods and inspection techniques employed are based on the most modern available. Reactor pressure vessel integrity is confirmed by a combination of arguments including safety factors inferred from the successful pre-service overpressure test, leak-before-break analysis and probabilistic assessment. In the case of other parts of the pressure circuits that are more accessible, comprising the boiler shells and interconnecting gas duct work, in-service inspection is a major element of the safety substantiation. The assessment and inspection techniques and the materials property data have been underpinned for many years by extensive research and development programmes and in-reactor monitoring of representative samples has also been undertaken. The paper summarises the work carried out to demonstrate the long term integrity of the Magnox pressure circuits and provides examples of the results obtained. (author)

Weld evaluation on spherical pressure vessels using holographic interferometry

International Nuclear Information System (INIS)

Boyd, D.M.; Wilcox, W.W.

1980-01-01

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

Analysis of aging mechanism and management for HTR-PM reactor pressure vessel

International Nuclear Information System (INIS)

Sun Yunxue; Shao Jin

2015-01-01

Reactor pressure vessel is an important part of the reactor pressure boundary, its important degree ranks high in ageing management and life assessment of nuclear power plant. Carrying out systematic aging management to ensure reactor pressure vessel keeping enough safety margins and executing design functions is one of the key factors to guarantee security and stability operation for nuclear power plant during the whole lifetime and prolong life. This paper briefly introduces the structure and aging mechanism of reactor pressure vessel in pressurized water reactor nuclear power plant, and introduces the design principle and structure characteristics of HTR-PM. At the same time, this paper carries out preliminary analysis and exploration. and discusses aging management of HTR-PM reactor pressure vessel. Finally, the advice of carring out aging management for HTR-PM reactor pressure vessel is proposed. (authors)

The relevance of crack arrest phenomena for pressure vessel structural integrity assessment

International Nuclear Information System (INIS)

Connors, D.C.; Dowling, A.R.; Flewitt, P.E.J.

1996-01-01

The potential role of a crack arrest argument for the structural integrity assessments of steel pressure vessels and the relationship between crack initiation and crack arrest philosophies are described. A typical structural integrity assessment using crack initiation fracture mechanics is illustrated by means of a case study based on assessment of the steel pressure vessels for Magnox power stations. Evidence of the occurrence of crack arrest in structures is presented and reviewed, and the applications to pressure vessels which are subjected to similar conditions are considered. An outline is given of the material characterisation that would be required to undertake a crack arrest integrity assessment. It is concluded that crack arrest arguments could be significant in the structural integrity assessment of PWR reactor pressure vessels under thermal shock conditions, whereas for Magnox steel pressure vessels it would be limited in its potential to supporting existing arguments. (author)

Computational analysis of transient gas release from a high pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Pedro, G.; Oshkai, P.; Djilali, N. [Victoria Univ., BC (Canada). Inst. for Integrated Energy Systems; Penau, F. [CERAM Euro-American Inst. of Technology, Sophia Antipolis (France)

2006-07-01

Gas jets exiting from compressed vessels can undergo several regimes as the pressure in the vessel decreases, and a greater understanding of the characteristics of gas jets is needed to determine safety requirements in the transport, distribution, and use of hydrogen. This paper provided a study of the bow shock waves that typically occur during the initial stage of a gas jet incident. The transient behaviour of an initiated jet was investigated using unsteady, compressible flow simulations. The gas was considered to be ideal, and the domain was considered to be axisymmetric. Tank pressure for the analysis was set at a value of 100 atm. Jet structure was examined, as well as the shock structures and separation due to adverse pressure gradients at the nozzle. Shock structure displacement was also characterized.

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.

Problems in manufacturing and transport of pressure vessels of integral reactors

International Nuclear Information System (INIS)

Kralovec, J.

1997-01-01

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)

Interpretation of strain measurements on nuclear pressure vessels

International Nuclear Information System (INIS)

Andersen, S.I.; Engbaek, P.

1979-11-01

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

Ductile fracture toughness of modified A 302 Grade B Plate materials, data analysis. Volume 1

Energy Technology Data Exchange (ETDEWEB)

McCabe, D.E.; Manneschmidt, E.T.; Swain, R.L.

1997-01-01

The goal of this work was to develop ductile fracture toughness data in the form of J-R curves for modified A302 grade B plate materials typical of those used in reactor pressure vessels. A previous experimental study on one heat of A302 grade B plate showed decreasing J-R curves with increased specimen thickness. This characteristic has not been observed in tests made on recent production materials of A533 grade B and A508 class 2 pressure vessel steels. It was unknown if the departure from norm for the material was a generic characteristic for all heats of A302 grade B steels or unique to that particular plate. Seven heats of modified A302 grade B steel and one heat of vintage A533 grade B steel were tested for chemical content, tensile properties, Charpy transition temperature curves, drop-weight nil-ductility transition (NDT) temperature, and J-R curves. Tensile tests were made in the three principal orientations and at four temperatures, ranging from room temperature to 550F. Charpy V-notch transition temperature curves were obtained in longitudinal, transverse, and short transverse orientations. J-R curves were made using four specimen sizes (1/2T, 1T, 2T, and 4T). The fracture mechanics-based evaluation method covered three test orientations and three test temperatures (80, 400, and 550F). However, the coverage of these variables was contingent upon the amount of material provided. Drop-weight NDT temperature was determined for the T-L orientation only. None of the heats of modified A302 grade B showed size effects of any consequence on the J-R curve behavior. Crack orientation effects were present, but none were severe enough to be reported as atypical. A test temperature increase from 180 to 550F produced the usual loss in J-R curve fracture toughness. Generic J-R curves and curve fits were generated to represent each heat of material. This volume deals with the evaluation of data and the discussion of technical findings. 8 refs., 18 figs., 8 tabs.

Ductile fracture toughness of modified A 302 Grade B Plate materials, data analysis. Volume 1

International Nuclear Information System (INIS)

McCabe, D.E.; Manneschmidt, E.T.; Swain, R.L.

1997-01-01

The goal of this work was to develop ductile fracture toughness data in the form of J-R curves for modified A302 grade B plate materials typical of those used in reactor pressure vessels. A previous experimental study on one heat of A302 grade B plate showed decreasing J-R curves with increased specimen thickness. This characteristic has not been observed in tests made on recent production materials of A533 grade B and A508 class 2 pressure vessel steels. It was unknown if the departure from norm for the material was a generic characteristic for all heats of A302 grade B steels or unique to that particular plate. Seven heats of modified A302 grade B steel and one heat of vintage A533 grade B steel were tested for chemical content, tensile properties, Charpy transition temperature curves, drop-weight nil-ductility transition (NDT) temperature, and J-R curves. Tensile tests were made in the three principal orientations and at four temperatures, ranging from room temperature to 550F. Charpy V-notch transition temperature curves were obtained in longitudinal, transverse, and short transverse orientations. J-R curves were made using four specimen sizes (1/2T, 1T, 2T, and 4T). The fracture mechanics-based evaluation method covered three test orientations and three test temperatures (80, 400, and 550F). However, the coverage of these variables was contingent upon the amount of material provided. Drop-weight NDT temperature was determined for the T-L orientation only. None of the heats of modified A302 grade B showed size effects of any consequence on the J-R curve behavior. Crack orientation effects were present, but none were severe enough to be reported as atypical. A test temperature increase from 180 to 550F produced the usual loss in J-R curve fracture toughness. Generic J-R curves and curve fits were generated to represent each heat of material. This volume deals with the evaluation of data and the discussion of technical findings. 8 refs., 18 figs., 8 tabs

Design, fabrication and quality assurance of pressure vessels

International Nuclear Information System (INIS)

Kimura, Ichiro; Miki, Masao; Yamazaki, Tsuneji; Tanaka, Yoshikazu; Sato, Misao

1978-01-01

The production facilities, design and manufacturing technologies, and quality assurance in the Toyo Works, Ehime Manufactory, Sumitomo Heavy Industries, Ltd., which manufactures pressure vessels, are described, and especially the actual example of non-destructive tests is shown. The Toyo Works was completed in April, 1973, to manufacture large structures such as pressure vessels, offshore structures and bridges. The total area of the site is 535,000 m 2 , that of factory buildings is 33,600 m 2 , and the outdoor assembling yard is 114,800 m 2 . The large dry dock and main installations such as 12,000 tf hydraulic press, an annealing furnace, a heat treating furnace, a quenching tank, a horizontal boring machine, 6 m vertical lathe, various welding machines, 8 MeV X-ray apparatus, sand blasting and pickling facilities, and two 160 t cranes for shipment are arranged so as to enable smooth flow of production. The standards for chemical pressure vessels in various countries are compared, and considerably high allowable stress is adopted in Europe. The design and stress analysis of pressure vessels are carried out in accordance with ASME Section 8, Div. 1 or Div. 2. As for the materials, attention must be paid to the change of properties due to heat and strain, temper brittleness, low temperature toughness and so on. The quality assurance system must be established to observe the requirements of standards. (Kako, I.)

Analysis of nuclear reactor pressure vessel flanges

International Nuclear Information System (INIS)

Oliveira, C.A.N. de; Augusto, O.B.

1985-01-01

This work proposes a methodology for the structural analysis of high diameter nuclear reactor pressure vessel flanges. In the analysis the vessel is divided into shell-of-revolution elements, the flanges are represented by rigid rings, and the bolts are treated as beams. The flexibility method is used for solving the problem, and the results are compared with results obtained by the finite element method. (Author) [pt

Contribution for the improvement of pressurized thermal shock assessment methodologies in PWR pressure vessels

International Nuclear Information System (INIS)

Gomes, Paulo de Tarso Vida

2005-01-01

The structural integrity assessment of nuclear reactor pressure vessel, concerned to Pressurized Thermal Shock (PTS) accidents, became a necessity and has been investigated since the eighty's. The recognition of the importance of PTS assessment has led the international nuclear technology community to devote a considerable research effort directed to the complete integrity assessment process of the Reactor Pressure Vessels (VPR). Researchers in Europe, Japan and U.S.A. have concentrated efforts in the VPR structural and fracture analysis, conducting experiments to best understand how specific factors act on the behavior of discontinuities, under PTS loading conditions. The main goal of this work is to study de structural behavior of an 'in scale' PWR nuclear reactor pressure vessel model, containing actual discontinuities, under loading conditions generated by a pressurized thermal shock. To construct the pressure vessel model utilized in this research, the approach developed by Barroso (1995) and based on likelihood studies, related to thermal-hydraulic behavior during the PTS was employed. To achieve the objective of this research, a new methodology to generate cracks, with known geometry and localization in the vessel model wall was developed. Additionally, an hydraulic circuit, able to flood the vessel model, heated to 300 deg C, with 10 m 3 of water at 8 deg C, in 170 seconds, was built. Thermo-hydraulic calculations using RELAP5/M0D 3.2.2γ computational code were done, to estimate the temperature profiles during the cooling time. The resulting data subsidized the thermo-structural calculations that were accomplished using ANSYS 7.01 computational code, for both 2D and 3D models. So, the stress profiles obtained with these calculations were associated with fracture mechanics concepts, to assess the crack growth behavior in the VPR model wall. After the PTS test, the VPR model was submitted to destructive and non-destructive inspections. The results

Development of computational methods of design by analysis for pressure vessel components

International Nuclear Information System (INIS)

Bao Shiyi; Zhou Yu; He Shuyan; Wu Honglin

2005-01-01

Stress classification is not only one of key steps when pressure vessel component is designed by analysis, but also a difficulty which puzzles engineers and designers at all times. At present, for calculating and categorizing the stress field of pressure vessel components, there are several computation methods of design by analysis such as Stress Equivalent Linearization, Two-Step Approach, Primary Structure method, Elastic Compensation method, GLOSS R-Node method and so on, that are developed and applied. Moreover, ASME code also gives an inelastic method of design by analysis for limiting gross plastic deformation only. When pressure vessel components design by analysis, sometimes there are huge differences between the calculating results for using different calculating and analysis methods mentioned above. As consequence, this is the main reason that affects wide application of design by analysis approach. Recently, a new approach, presented in the new proposal of a European Standard, CEN's unfired pressure vessel standard EN 13445-3, tries to avoid problems of stress classification by analyzing pressure vessel structure's various failure mechanisms directly based on elastic-plastic theory. In this paper, some stress classification methods mentioned above, are described briefly. And the computational methods cited in the European pressure vessel standard, such as Deviatoric Map, and nonlinear analysis methods (plastic analysis and limit analysis), are depicted compendiously. Furthermore, the characteristics of computational methods of design by analysis are summarized for selecting the proper computational method when design pressure vessel component by analysis. (authors)

Bounding the conservatism in flaw-related variables for pressure vessel integrity analyses

International Nuclear Information System (INIS)

Foulds, J.R.; Kennedy, E.L.

1993-01-01

The fracture mechanics-based integrity analysis of a pressure vessel, whether performed deterministically or probabilistically, requires use of one or more flaw-related input variables, such as flaw size, number of flaws, flaw location, and flaw type. The specific values of these variables are generally selected with the intent to ensure conservative predictions of vessel integrity. These selected values, however, are largely independent of vessel-specific inspection results, or are, at best, deduced by ''conservative'' interpretation of vessel-specific inspection results without adequate consideration of the pertinent inspection system performance (reliability). In either case, the conservatism associated with the flaw-related variables chosen for analysis remains examination (NDE) technology and the recently formulated ASME Code procedures for qualifying NDE system capability and performance (as applied to selected nuclear power plant components) now provides a systematic means of bounding the conservatism in flaw-related input variables for pressure vessel integrity analyses. This is essentially achieved by establishing probabilistic (risk)-based limits on the assigned variable values, dependent upon the vessel inspection results and on the inspection system unreliability. Described herein is this probabilistic method and its potential application to: (i) defining a vessel-specific ''reference'' flaw for calculating pressure-temperature limit curves in the deterministic evaluation of pressurized water reactor (PWR) reactor vessels, and (ii) limiting the flaw distribution input to a PWR reactor vessel-specific, probabilistic integrity analysis for pressurized thermal shock loads

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.

Design study on steam generator integration into the VVER reactor pressure vessel

International Nuclear Information System (INIS)

Hort, J.; Matal, O.

2004-01-01

The primary circuit of VVER (PWR) units is arranged into loops where the heat generated by the reactor is removed by means of main circulating pumps, loop pipelines and steam generators, all located outside the reactor pressure vessel. If the primary circuit and reactor core were integrated into one pressure vessel, as proposed, e.g., within the IRIS project (WEC), a LOCA situation would be limited by the reactor pressure vessel integrity only. The aim of this design study regarding the integration of the steam generator into the reactor pressure vessel was to identify the feasibility limits and some issues. Fuel elements and the reactor pressure vessel as used in the Temelin NPP were considered for the analysis. From among the variants analyzed, the variant with steam generators located above the core and vertically oriented circulating pumps at the RPV lower bottom seems to be very promising for future applications

Light-water-reactor pressure-vessel surveillance dosimetry using solid-state track recorders

International Nuclear Information System (INIS)

Ruddy, F.H.; Roberts, J.H.; Gold, R.; Preston, C.C.

1983-07-01

The accumulation of neutron dose by the pressure vessel of an operating nuclear power plant results in damage in the form of steel embrittlement. In order to ascertain the safe operating lifetime of the reactor pressure vessel, dosimetric measurements must be made to evaluate the neutron dose to the pressure vessel and relate this dose to the cumulative radiation damage. Advanced dosimetry techniques are being evaluated for surveillance of operating reactors. Solid-state track recorder (SSTR) techniques are included among these advanced dosimetry techniques. Described herein are low neutron fluence calibration and standardization measurements that are being carried out in pressure vessel mockup benchmark neutron fields in the USA, Belgium, and England. In addition, high fluence SSTR dosimetry capsules have been irradiated with metallurgical specimens in a pressure vessel mockup facility. The design and deployment of advances SSTR dosimetry capsules in operating power reactors are also described

Stress categorization in nozzle to pressure vessel connections finite elements models

International Nuclear Information System (INIS)

Albuquerque, Levi Barcelos de

1999-01-01

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

A novel high pressure, high temperature vessel used to conduct long-term stability measurements of silicon MEMS pressure transducers

Science.gov (United States)

Wisniewiski, David

2014-03-01

The need to quantify and to improve long-term stability of pressure transducers is a persistent requirement from the aerospace sector. Specifically, the incorporation of real-time pressure monitoring in aircraft landing gear, as exemplified in Tire Pressure Monitoring Systems (TPMS), has placed greater demand on the pressure transducer for improved performance and increased reliability which is manifested in low lifecycle cost and minimal maintenance downtime through fuel savings and increased life of the tire. Piezoresistive (PR) silicon MEMS pressure transducers are the primary choice as a transduction method for this measurement owing to their ability to be designed for the harsh environment seen in aircraft landing gear. However, these pressure transducers are only as valuable as the long-term stability they possess to ensure reliable, real-time monitoring over tens of years. The "heart" of the pressure transducer is the silicon MEMS element, and it is at this basic level where the long-term stability is established and needs to be quantified. A novel High Pressure, High Temperature (HPHT) vessel has been designed and constructed to facilitate this critical measurement of the silicon MEMS element directly through a process of mechanically "floating" the silicon MEMS element while being subjected to the extreme environments of pressure and temperature, simultaneously. Furthermore, the HPHT vessel is scalable to permit up to fifty specimens to be tested at one time to provide a statistically significant data population on which to draw reasonable conclusions on long-term stability. With the knowledge gained on the silicon MEMS element, higher level assembly to the pressure transducer envelope package can also be quantified as to the build-effects contribution to long-term stability in the same HPHT vessel due to its accommodating size. Accordingly, a HPHT vessel offering multiple levels of configurability and robustness in data measurement is presented, along

Further fields of application for prestressed cast iron pressure vessels (PCIV)

International Nuclear Information System (INIS)

Guelicher, L.; Schilling, F.E.

1977-01-01

The redundancy of the prestressing system of prestressed structures as well as the clear separation of sealing and load-carrying functions of prestressed cast iron pressure vessels offer substantial advantages over conventional welded steel pressure vessels. Because of the temperature resistance of cast iron up to 400 0 C it is possible to build prestressed pressure vessels commercially as hot-working structures. The compressive strength of cast iron, which is 25 times as high as that of concrete allows for a very compact design of the PCIV. Further specific properties of the PCIV like pre-fabrication of the vessel in the production plant - made possible by a structure assembled from segments - short assembly periods at the construction site etc., may open more fields of application. - PCIV as pressurized storage tanks for the emergency shut down system in nuclear power stations. - PCIV as high pressure vessel for the chemical industry. - PCIV as energy storage. - PCIV for light water reactors. - PCIV as burst protection. It is concluded that the application of prestressed cast iron promises to be successful where either structures with large volumes and high pressures and/or temperatures are required or where aspects of safety allow for efficient use of prestressed structures. (Auth.)

Cylindrical reinforced-concrete pressure vessel for nuclear reactors

International Nuclear Information System (INIS)

Vaessen, F.

1975-01-01

The cylindrical pressure vessel has got a wall and an isolating layer composed of blocks of heat-resistant concrete or of ceramic material. The side of the isolating layer facing the interior of the presssure vessel is coated by a liner made of metallic material. In cold state and without internal pressure, the radius of this liner is smaller by a differential amount than that of the isolating layer. By means of radially displaceable fixing elements consisting of an anchoring tube and a holding tube inserted in it, the liner can be made to rest against the isolating layer. This occurs if the pressure vessel is brought to operational temperature. The anchoring tube is attached to the isolating layer whereas the displaceable holding tube is connected with the liner. The possible relative travelling distance of these two elements is equal to the difference of length of the two radii. In addition, the liner may consist of single parts connected with each other through compensating flanges. There may also be additional springs arranged between the isolating layer and the liner. (DG/PB) [de

Structural analysis and evaluation for the design of pressure vessel

International Nuclear Information System (INIS)

Arai, K.; Uragami, K.; Funada, T.; Baba, K.; Kira, T.

1977-01-01

For the design of pressure vessel, the detailed structural analysis such as the fatigue analysis under operating conditions is required by ASME Code or Japanese regulation. Accordingly, it should be verified by the analysis that the design of the pressure vessel is in compliance with the stress limitation defined in the Code or the regulation. However, it was apparent that the analysis is very complicated and takes a lot of time to evaluate in accordance with the Code requirements. Thereupon we developed the computer program by which we can perform the stress analysis with correctness and comparatively in a short period of design work reflecting the calculation results on detailed drawings to be used for fabrication. The computer program is controlled in combination with the system of the design work and out put list of the program can be directly used for the stress analysis report which is issued to customers. In addition to the above computer program, we developed the specific three dimensional finite element computer program to make sure of the structural integrity of the vessel head and flanges which are most complex for the analysis compared with the stress distribution measured by strain gauges on the vessel head and flange. Besides the structural analysis, the fracture mechanics analysis for the purpose of preventing the pressure vessel from the brittle fracture during heat-up and cool-down operation is also important and thereby we showed herein that the pressure vessel is in safety against the brittle fracture for the specified operating conditions. As a result of the above-mentioned analysis, the pressure vessel is designed with safety from the stand-points of the structural intensity and the fracture mechanics. (auth.)

Problems in Pressure Vessel Design and Manufacture

Energy Technology Data Exchange (ETDEWEB)

Hellstroem, O [Uddeholms AB, Degerfors (Sweden); Nilson, Ragnar [AB Atomenergi, Nykoeping (Sweden)

1963-05-15

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.

Problems in Pressure Vessel Design and Manufacture

International Nuclear Information System (INIS)

Hellstroem, O.; Nilson, Ragnar

1963-05-01

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

Structural features and in-service inspection of the LTHR-200 pressure vessel

International Nuclear Information System (INIS)

Xiong Dunshi; He Shuyan; Liu Junjie; Yu Suyuan

1993-01-01

LTHR-200 is a low temperature district-heating reactor. It adopts double-shell design pressure vessel and metal containment. Because of the safety and structural features of the reactor, the in-service inspection of the pressure vessel can be simplified greatly. LTHR-200 is an integrated arrangement. Both its core components and the main heat exchangers are contained in the reactor pressure vessel. The coolant of the main loop is run by a full-power natural circulation and there need no main pumps and pipes. Thus, the reactor pressure vessel constitutes the pressure boundary of the reactor's main loop coolant. In regard to these features, a small-sized containment is designed for the reactor. The metal safety container with a small volume is placed closely around the reactor pressure vessel. Outside the metal containment, there is a large reinforced concrete construction for the reactor. Their main operation and design parameters are as follows: The pressure vessel: operation pressure = 2.4 MPa; design pressure = 3.0 MPa; design temperature = 250 deg C; 40 year fast neutron (E>1MeV) fluence in the belt-line region = < 10E16n/cm; internal diameter = 5000 mm; material SA516-70; shell thickness 65 mm; The metal containment: maximum operation pressure = 1.8 MPa; design pressure = 1.8 MPa; design temperature = 250 deg. C; upper internal diameter 7000 mm; lower internal diameter = 5600 mm; material = SA516-70; shell thickness, upper part = 80 mm; lower part = 50 mm. All penetrating pipes through the pressure vessel are located at the top penetration section of the shell. All the internal diameters of penetrating pipes are less than 50 mm. Inside and outside the metal containment wall respectively, isolating valves are connected to the reactor coolant pipe which passes through the containment. These two isolating valves use different driving methods. Every penetrating part of the reactor construction uses a proper form of structure according to safety requirements

Renovation of the sealing planes of WWER-400 reactors pressure vessel

International Nuclear Information System (INIS)

Jablonicky, P.; Pilat, P.

2007-01-01

An article describes technical solution for renovation of the sealing planes of WWER-440 reactor's pressure vessel. Four nickel sealing rings placed in four concentric grooves are providing hermetic sealing between the vessel and the lid of this type of the reactor. Impeccable seal of the reactor's pressure vessel, where the fission reaction takes place, represents a basic security factor for safe electric energy production. Principle of renovation of the reactor's pressure vessel and lid sealing planes is based on mechanical enlargement of defective grooves and following cladding of the new material by TIG welding. Final step for renovation includes machining of new grooves according to geometrical and surface quality requirements (Authors)

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

International Nuclear Information System (INIS)

Duijvestijn, G.; Birchley, J.; Reichlin, K.

1997-01-01

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

Reactor Pressure Vessel Steels

Energy Technology Data Exchange (ETDEWEB)

Van de Velde, J.; Fabry, A.; Van Walle, E.; Chaoudi, R

1998-07-01

SCK-CEN's R and D programme on Reactor Pressure Vessel (RPV) Steels in performed in support of the RVP integrity assessment. Its main objectives 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 the applied methodology on a broad database; (3) to achieve regulatory acceptance and industrial use. Progress and achievements in 1999 are reported.

Needs for evaluated covariance data for reactor pressure vessel dosimetry

International Nuclear Information System (INIS)

Maerker, R.E.; Broadhead, B.L.; Wagschal, J.J.

1992-01-01

This report discusses new methodology for quantifying and then reducing uncertainties in the calculated pressure vessel fluences of a pressurized water reactor (PWR). The technique involves combining the integral results of the calculated and measured PWR surveillance dosimetry activities with the differential data used in the calculations, along with covariances of all the quantities, into a generalized linear least-squares adjustment procedure. Based on analysis of both PWRs and test reactor benchmarks, substantial evidence now exists to support the conclusion that, of all the nuclear as well as non-nuclear differential data considered, ENDF/B-VI values of the total inelastic iron cross sections and their covariances are the most important data controlling the outcome of the adjustment procedure. Predicted adjustments in these cross sections provided the stimulus for new measurements, the results of which impacted the ENDF/B-VI evaluation of iron 56

Method of detecting construction faults in concrete pressure vessels

International Nuclear Information System (INIS)

Robertson, S.A.; Duhoux, M.; Dawance, G.; Carrie, C.; Morel, D.

1976-01-01

A major problem in the design and construction of concrete pressure vessels for nuclear power stations is the risk of excessive air leaks through the concrete itself, due to faulty construction. The 'sonic coring' method of non-destructive concrete testing has been used successfully in pile and diaphragm wall construction control for several years, and the potential use of this method to control the presence of faults in concrete pressure vessels is here described. (author)

Large inelastic deformation analysis of steel pressure vessels at high temperature

International Nuclear Information System (INIS)

Ikonen, K.

2001-01-01

This publication describes the calculation methodology developed for a large inelastic deformation analysis of pressure vessels at high temperature. Continuum mechanical formulation related to a large deformation analysis is presented. Application of the constitutive equations is simplified when the evolution of stress and deformation state of an infinitesimal material element is considered in the directions of principal strains determined by the deformation during a finite time increment. A quantitative modelling of time dependent inelastic deformation is applied for reactor pressure vessel steels. Experimental data of uniaxial tensile, relaxation and creep tests performed at different laboratories for reactor pressure vessel steels are investigated and processed. An inelastic deformation rate model of strain hardening type is adopted. The model simulates well the axial tensile, relaxation and creep tests from room temperature to high temperature with only a few fitting parameters. The measurement data refined for the inelastic deformation rate model show useful information about inelastic deformation phenomena of reactor pressure vessel steels over a wide temperature range. The methodology and calculation process are validated by comparing the calculated results with measurements from experiments on small scale pressure vessels. A reasonably good agreement, when taking several uncertainties into account, is obtained between the measured and calculated results concerning deformation rate and failure location. (orig.)

Testing of Full Scale Flight Qualified Kevlar Composite Overwrapped Pressure Vessels

Science.gov (United States)

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

2007-01-01

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

12 CFR 533.5 - Related agreements considered a single agreement.

Science.gov (United States)

2010-01-01

... DISCLOSURE AND REPORTING OF CRA-RELATED AGREEMENTS § 533.5 Related agreements considered a single agreement... entered into within the same 12-month period; and (3) Are each in fulfillment of the CRA. (b... in fulfillment of the CRA, if the contracts were negotiated in a coordinated fashion and a NGEP is a...

Interpreting ASME limits and philosophy in FEA of pressure vessel parts

International Nuclear Information System (INIS)

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

1995-01-01

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

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

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false General requirements for boilers and pressure... NONMETAL MINES Compressed Air and Boilers § 57.13001 General requirements for boilers and pressure vessels. All boilers and pressure vessels shall be constructed, installed, and maintained in accordance with...

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

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false General requirements for boilers and pressure... MINES Compressed Air and Boilers § 56.13001 General requirements for boilers and pressure vessels. All boilers and pressure vessels shall be constructed, installed, and maintained in accordance with the...

Structural integrity evaluation of PWR nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Cruz, Julio R.B.; Mattar Neto, Miguel

1999-01-01

The reactor pressure vessel (RPV) is the most important structural component of a PWR nuclear power plant. It contains the reactor core and is the main component of the primary system pressure boundary, the system responsible for removing the heat generated by the nuclear reactions. It is considered not replaceable and, therefore, its lifetime is a key element to define the plant life as a whole. Three critical issues related to the reliability of the RPV structural integrity come out by reason of the radiation damage imposed to the vessel material during operation. These issues concern the definition of pressure versus temperature limits for reactor heatup and cooldown, pressurized thermal shock evaluation and assessment of reactor vessels with low upper shelf Charpy impact energy levels. This work aims to present the major aspects related to these topics. The requirements for preventing fracture of the RPV are reviewed as well as the available technology for assessing the safety margins. For each mentioned problem, the several steps for structural integrity evaluation are described and the analysis methods are discussed. (author)

Probabilistic assessment of pressure vessel and piping reliability

International Nuclear Information System (INIS)

Sundararajan, C.

1986-01-01

The paper presents a critical review of the state-of-the-art in probabilistic assessment of pressure vessel and piping reliability. First the differences in assessing the reliability directly from historical failure data and indirectly by a probabilistic analysis of the failure phenomenon are discussed and the advantages and disadvantages are pointed out. The rest of the paper deals with the latter approach of reliability assessment. Methods of probabilistic reliability assessment are described and major projects where these methods are applied for pressure vessel and piping problems are discussed. An extensive list of references is provided at the end of the paper

Proposal of Ex-Vessel dosimetry for pressure vessel Atucha II

International Nuclear Information System (INIS)

Chiaraviglio, N.; Bazzana, S.

2013-01-01

Nuclear reactor dosimetry has the purpose of guarantee that changes in material mechanical properties of critical materials do not compromise the reactor safety. In PWR in which the top of the reactor vessel is open once a year, is possible to use Charpy specimens to measure the change in mechanical properties. Atucha II nuclear power plant is a reactor with on-line refueling so there is no access to the inside of the pressure vessel. Because of this, ex-vessel dosimetry must be performed and mechanical properties changes must be inferred from radiation damage estimations. This damage can be calculated using displacement per atom cross sections and a transport code such as MCNP. To increase results reliability it is proposed to make a neutron spectrum unfolding using activation dosimeters irradiated during one operation cycle of the power plant. In this work we present a dosimetry proposal for such end, made in base of unfolding procedures and experimental background. (author) [es

In-place thermal annealing of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Server, W.L.

1985-04-01

Radiation embrittlement of ferritic pressure vessel steels increases the ductile-brittle transition temperature and decreases the upper shelf level of toughness as measured by Charpy impact tests. A thermal anneal cycle well above the normal operating temperature of the vessel can restore most of the original Charpy V-notch energy properties. The Amry SM-1A test reactor vessel was wet annealed in 1967 at less than 343 0 C (650 0 F), and wet annealing of the Belgian BR-3 reactor vessel at 343 0 C (650 0 F) has recently taken place. An industry survey indicates that dry annealing a reactor vessel in-place at temperatures as high as 454 0 C (850 0 F) is feasible, but solvable engineering problems do exist. Economic considerations have not been totally evaluated in assessing the cost-effectiveness of in-place annealing of commercial nuclear vessels. An American Society for Testing and Materials (ASTM) task group is upgrading and revising guide ASTM E 509-74 with emphasis on the materials and surveillance aspects of annealing rather than system engineering problems. System safety issues are the province of organizations other than ASTM (e.g., the American Society of Mechanical Engineers Boiler and Pressure Vessel Code body)

Variability of mechanical properties of nuclear pressure vessel steels

International Nuclear Information System (INIS)

Petrequin, P.; Soulat, P.

1980-01-01

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.) [de

Recent experiences and problems in conducting pressure vessel surveillance examinations

International Nuclear Information System (INIS)

Perrin, J.S.

1979-01-01

Each of the commercial power reactors in the U.S.A. has a pressure vessel surveillance program. The purpose of the programs is to monitor the effects of radiation on the mechanical properties on the steel pressure vessels. A program for a given reactor includes a series of irradiation capsules containing neutron dosimeters and mechanical property specimens. The capsules are periodically removed during the life of the reactor and evaluated. The surveillance capsule examinations conducted to date have been valuable in assessing the effects of radiation on pressure vessels. However, a number of problems have been observed in the course of capsule examinations which potentially could reduce the maximum value of the data obtained. These problems are related to specimen design and preparation, capsule design and preparation, capsule installation and removal, capsule disassembly, specimen testing and evaluation, program documentation, and quality assurance. Examples of problems encountered in the preceding areas are presented in the present paper, and recommendations are made for minimization or prevention of these problems in future programs. Included in the recommendations is that appropriate ASTM standards, ASME Boiler and Pressure Vessel Code sections, and NRC regulations provide the appropriate framework for prevention of problems

Advanced Approach of Reactor Pressure Vessel In-service Inspection

International Nuclear Information System (INIS)

Matokovic, A.; Picek, E.; Pajnic, M.

2006-01-01

The most important task of every utility operating a nuclear power plant is the continuously keeping of the desired safety and reliability level. This is achieved by the performance of numerous inspections of the components, equipment and system of the nuclear power plant in operation and in particular during the scheduled maintenance periods at re-fueling time. Periodic non-destructive in-service inspections provide most relevant criteria of the integrity of primary circuit pressure components. The task is to reliably detect defects and realistically size and characterize them. One of most important and the most extensive examination is a reactor pressure vessel in-service inspection. That inspection demand high standards of technology and quality and continual innovation in the field of non-destructive testing (NDT) advanced technology as well as regarding reactor pressure vessel tool and control systems. A remote underwater contact ultrasonic technique is employed for the examination of the defined sections (reactor welds), whence eddy current method is applied for clad surface examinations. Visual inspection is used for examination of the vessel inner surface. The movement of probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with NDT systems. The successful performance is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen NDT techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state of the art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. The advanced approach as presented in this paper offer more flexibility of application (non-destructive tests, local grinding action as well as taking of boat samples

Pool critical assembly pressure vessel facility benchmark

International Nuclear Information System (INIS)

Remec, I.; Kam, F.B.K.

1997-07-01

This pool critical assembly (PCA) pressure vessel wall facility benchmark (PCA benchmark) is described and analyzed in this report. Analysis of the PCA benchmark can be used for partial fulfillment of the requirements for the qualification of the methodology for pressure vessel neutron fluence calculations, as required by the US Nuclear Regulatory Commission regulatory guide DG-1053. Section 1 of this report describes the PCA benchmark and provides all data necessary for the benchmark analysis. The measured quantities, to be compared with the calculated values, are the equivalent fission fluxes. In Section 2 the analysis of the PCA benchmark is described. Calculations with the computer code DORT, based on the discrete-ordinates method, were performed for three ENDF/B-VI-based multigroup libraries: BUGLE-93, SAILOR-95, and BUGLE-96. An excellent agreement of the calculated (C) and measures (M) equivalent fission fluxes was obtained. The arithmetic average C/M for all the dosimeters (total of 31) was 0.93 ± 0.03 and 0.92 ± 0.03 for the SAILOR-95 and BUGLE-96 libraries, respectively. The average C/M ratio, obtained with the BUGLE-93 library, for the 28 measurements was 0.93 ± 0.03 (the neptunium measurements in the water and air regions were overpredicted and excluded from the average). No systematic decrease in the C/M ratios with increasing distance from the core was observed for any of the libraries used

Babcock experience of automated ultrasonic non-destructive testing of PWR pressure vessels during manufacture

International Nuclear Information System (INIS)

Dikstra, B.J.; Farley, J.M.; Scruton, G.

1990-01-01

Major developments in ultrasonic techniques, equipment and systems for automated inspection have lead, over a period of about ten years, to the regular application of sophisticated computer-controlled systems during the manufacture of nuclear reactor pressure vessels. Ten years ago the use of procedures defined in a code such as ASME XI might have been considered sufficient, but it is now necessary, as was demonstrated by the results of the UKAEA defect detection trials and the PISC II trials, to apply more comprehensive arrays of probes and higher test sensitivities. The ultrasonic techniques selected are demonstrated to be adequate by modelling or test-block exercises, the automated systems applied are subject to stringent quality assurance testing, and very rigorous inspection procedures are used in conjunction with a high degree of automation to ensure reproducibility of inspection quality. The state-of-the-art in automated ultrasonic testing of pressure vessels by Babcock is described. Current developments by the company, including automated flaw recognition, integrated modelling of inspection capability, and the use of electronically scanned variable-angle probes are reviewed. Examples quoted include the automated ultrasonic inspections of the Sizewell B pressurized water reactor vessel. (author)

Preliminary investigation of ultrasonic shear wave holography with a view to the inspection of pressure vessels

International Nuclear Information System (INIS)

Aldridge, E.E.; Clare, A.B.; Shepherd, D.A.

1975-01-01

The manner in which holography would fit into the general scheme of pressure vessel inspection is discussed. Compared to conventional A, B and C presentations holography requires a different processing of the ultrasonic signal and a mechanical scan which may be more demanding than that normally provided for a C display. Preliminary results are presented of the examination of artificial defects in steel plate using shear wave holography. (author)

Heat treatments in a conventional steel to reproduce the microstructure of a nuclear grade steel

International Nuclear Information System (INIS)

Rosalio G, M.

2014-01-01

The ferritic steels used in the manufacture of pressurized vessels of Boiling Water Reactors (BWR) suffer degradation in their mechanical properties due to damage caused by the neutron fluxes of high energy bigger to a Mega electron volt (E> 1 MeV) generated in the reactor core. The materials with which the pressurized vessels of nuclear reactors cooled by light water are built correspond to low alloy ferritic steels. The effect of neutron irradiation on these steels is manifested as an increase in hardness, mechanical strength, with the consequent decrease in ductility, fracture toughness and an increase in temperature of ductile-brittle transition. The life of a BWR is 40 years, its design must be considered sufficient margin of safety because pressure forces experienced during operation, maintenance and testing of postulated accident conditions. It is necessary that under these conditions the vessel to behave ductile and likely to propagate a fracture is minimized. The vessels of light water nuclear reactors have a bainite microstructure. Specifically, the reactor vessels of the nuclear power plant of Laguna Verde (Veracruz, Mexico) are made of a steel Astm A-533, Grade B Class 1. At present they are carrying out some welding tests for the construction of a model of a BWR, however, to use nuclear grade steel such as Astm A-533 to carry out some of the welding tests, is very expensive; perform these in a conventional material provides basic information. Although the microstructure present in the conventional material does not correspond exactly to the degree of nuclear material, it can take of reference. Therefore, it is proposed to conduct a pilot study to establish the thermal treatment that reproduces the microstructure of nuclear grade steel, in conventional steel. The resulting properties of the conventional steel samples will be compared to a JRQ steel, that is a steel Astm A-533, Grade B Class 1, provided by IAEA. (Author)

Automated ultrasonic shop inspection of reactor pressure vessel forgings

International Nuclear Information System (INIS)

Farley, J.M.; Dikstra, B.J.; Hanstock, D.J.; Pople, C.H.

1986-01-01

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

Investigation of the design of a metal-lined fully wrapped composite vessel under high internal pressure

Science.gov (United States)

Kalaycıoğlu, Barış; Husnu Dirikolu, M.

2010-09-01

In this study, a Type III composite pressure vessel (ISO 11439:2000) loaded with high internal pressure is investigated in terms of the effect of the orientation of the element coordinate system while simulating the continuous variation of the fibre angle, the effect of symmetric and non-symmetric composite wall stacking sequences, and lastly, a stacking sequence evaluation for reducing the cylindrical section-end cap transition region stress concentration. The research was performed using an Ansys® model with 2.9 l volume, 6061 T6 aluminium liner/Kevlar® 49-Epoxy vessel material, and a service internal pressure loading of 22 MPa. The results show that symmetric stacking sequences give higher burst pressures by up to 15%. Stacking sequence evaluations provided a further 7% pressure-carrying capacity as well as reduced stress concentration in the transition region. Finally, the Type III vessel under consideration provides a 45% lighter construction as compared with an all metal (Type I) vessel.

Pressure thermal shock analysis for nuclear reactor pressure vessel

International Nuclear Information System (INIS)

Galik, G.; Kutis, V.; Jakubec, J.; Paulech, J.; Murin, J.

2015-01-01

The appearance of structural weaknesses within the reactor pressure vessel or its structural failure caused by crack formation during pressure thermal shock processes pose as a severe environmental hazard. Coolant mixing during ECC cold water injection was simulated in a detailed CFD analysis. The temperature distribution acting on the pipe wall internal surface was calculated. Although, the results show the formation of high temperature differences and intense gradients, an additional structural analysis is required to determine the possibility of structural damage from PTS. Such an analysis will be the subject of follow-up research. (authors)

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.

Effect of a new specimen size on fatigue crack growth behavior in thick-walled pressure vessels

International Nuclear Information System (INIS)

Shariati, Mahmoud; Mohammadi, Ehsan; Masoudi Nejad, Reza

2017-01-01

Fatigue crack growth in thick-walled pressure vessels is an important factor affecting their fracture. Predicting the path of fatigue crack growth in a pressure vessel is the main issue discussed in fracture mechanics. The objective of this paper is to design a new geometrical specimen in fatigue to define the behavior of semi-elliptical crack growth in thick-walled pressure vessels. In the present work, the importance of the behavior of fatigue crack in test specimen and real conditions in thick-walled pressure vessels is investigated. The results of fatigue loading on the new specimen are compared with the results of fatigue loading in a cylindrical pressure vessel and a standard specimen. Numerical and experimental methods are used to investigate the behavior of fatigue crack growth in the new specimen. For this purpose, a three-dimensional boundary element method is used for fatigue crack growth under stress field. The modified Paris model is used to estimate fatigue crack growth rates. In order to verify the numerical results, fatigue test is carried out on a couple of specimens with a new geometry made of ck45. A comparison between experimental and numerical results has shown good agreement. - Highlights: • This paper provides a new specimen to define the behavior of fatigue crack growth. • We estimate the behavior of fatigue crack growth in specimen and pressure vessel. • A 3D finite element model has been applied to estimate the fatigue life. • We compare the results of fatigue loading for cylindrical vessel and specimens. • Comparison between experimental and numerical results has shown a good agreement.

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

Science.gov (United States)

2010-10-01

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

Stress analysis in a non axisymmetric loaded reactor pressure vessel

International Nuclear Information System (INIS)

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

1995-01-01

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

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

International Nuclear Information System (INIS)

Grover, J.L.; Cipolla, R.C.

1985-01-01

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)

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

International Nuclear Information System (INIS)

Silva, M.C.A.T. da.

1986-01-01

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) [pt

Reliability analysis of pipelines and pressure vessels at nuclear power plants

International Nuclear Information System (INIS)

Klemin, A.I.; Shiverskij, E.A.

1979-01-01

Reliability analysis of pipelines and pressure vessels at NPP is given. The main causes and failure mechanisms of these elements, the ways of reliability improvement and preventing of great damages are considered. The reliability estimation methods both according to the statistical operation data and under the conditions of absence of failure statistics are given. The main characteristics and actual reliability factors of pipelines and pressure vessels of three home NPP: the first in the world NPP, VK-50 and Beloyarsk NPP, are presented. From the start-up there were practically no failures of the pipelines and pressure vessels at the VK-50 pilot installation. The analysis of the operation experience of the first and second blocks of the Beloyarsk NPP, as well as the first in the world NPP, shows that the most part of failures of the pipelines and pressure vessels of these energy blocks with the channel reactors is connected with the coolant leakage at minority pipelines of a small diameter. The most part of failures at individual pipelines of the first and second blocks of the Beloyarsk NPP are connected with the leakages of stuffing boxes of switching off devices. It is noted that serious failures of large pipelines and pressure vessels at all home NPP under operation have not been observed

Completely integrated prestressed-concrete reactor pressure vessel, type 'Star'

International Nuclear Information System (INIS)

Neunert, B.; Jueptner, G.; Kumpf, H.

1975-01-01

The star support vessel is suitable for the connection to all primary circuit systems consisting of a main vessel and a number of satellite vessels around and connected to it, i.e. for LWR, HTR and process reactor. It must be made clear, however, that the PWR in particular with its components does not appear to be suited for the optimum incorporation in a prestressed-concrete pressure vessel system, no matter what kind. There are clear concepts about modifications which, however, require considerable development expenditure. (orig./LH) [de

7 CFR 53.3 - Authority.

Science.gov (United States)

2010-01-01

... 7 Agriculture 3 2010-01-01 2010-01-01 false Authority. 53.3 Section 53.3 Agriculture Regulations of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (Standards, Inspections... STANDARDS) Regulations Administration § 53.3 Authority. The Director is charged with the administration of...

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

Science.gov (United States)

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

2006-01-01

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

A classification system for pressure vessel shell failures

International Nuclear Information System (INIS)

Harrop, L.P.

1989-01-01

A system for classifying failures of the shells of pressure vessels is presented. The classification system is based on the way a failure physically manifests itself and not on imputed economic or safety significance. It is believed the described way of classifying the failures is useful for transferring information from one situation to another. In assigning names to types of failure, the intention has been to adopt explicit definitions rather than supposed colloquial usage. (author)

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.

In-service inspection program for the NCS-80 reactor pressure vessel