Apparatus for carrying out ultrasonic inspection of pressure vessels

International Nuclear Information System (INIS)

Dent, K.H.; Challender, R.S.

1975-01-01

A carriage-supported manipulator for taking an ultrasonic scanner mechanism into a coolant nozzle of a nuclear reactor pressure vessel is described. The manupulator is rotatable about the axis of the nozzle and is radially expansible to urge the scanner mechanism into a scanning position within the nozzle

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

International Nuclear Information System (INIS)

Bryson, J.W.

1979-01-01

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

Application of acoustic emission monitoring to pressure tests of a steam receiver vessel with flawed nozzle welds

International Nuclear Information System (INIS)

Woodward, B.; McDonald, N.R.; Hincksman, M.J.

1976-01-01

As part of the first stage of an Australian Welding Research Association co-operative research project, acoustic emission monitoring has been applied to a steam receiver vessel withdrawn from service owing to severe weld cracking. This technique is used to check acceptance standards for defects in nozzle welds and to apply modern methods of assessing the integrity of pressurised plant. Acoustic emission monitoring has been used, together with strain gauge measurements and ultrasonic scanning, to detect the occurrence of any significant defect growth during cyclic pressurisation of the vessel. During this first stage, no significant defect growth has been produced by 1000 cycles of pressure up to 24.1 MPa (3500 psi), subsequent pressurisation up to 35.8 MPa (5200 psi), or 97 per cent of the expected yield stress of the vessel shell. The small amount of acoustic emission detected was consistent with this result. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1989-07-01

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

Preventive protection device and method for bottom of reactor pressure vessel

International Nuclear Information System (INIS)

Hayashi, Eisaku; Kurosawa, Koichi; Furukawa, Hideyasu; Morinaka, Ren; Enomoto, Kunio; Otaka, Masahiro; Yoshikubo, Fujio; Chiba, Noboru; Sato, Kazunori.

1995-01-01

In a preventive protection device for improving stresses in reactor structural components by jetting highly pressurized water with cavitation bubbles from a jetting nozzle toward structural components in a reactor pressure vessel, a fixed structure to a CRD housing is provided with a rotational body attached to the structure, a multi joint arm and a jetting nozzle supported to the multi joint arm. The jetting nozzle is disposed at a position where the center of the jetting deviates from the center of the CRD housing. In addition, a monitoring camera is disposed for displaying the target for preventive protection. The state of stresses on a plurality of targets for preventive protection can be improved by the preventive protection device at a fixed position in the bottom of a reactor pressure vessel where housings stand densely, thereby enabling to attain the preventive protection operation easily and rapidly. (N.H.)

Corrosion Damage in Penetration Nozzle and Its Weldment of Reactor Pressure Vessel Head

International Nuclear Information System (INIS)

Lim, Yun Soo; Kim, Joung Soo; Kim, Hong Pyo; Hwang, Seong Sik; Yi, Young Sun; Kim, Dong Jin; Jung, Man Kyo

2003-07-01

The recent status on corrosion damage of reactor vessel head (RVH) penetration nozzles at primary water reactors (PWRs), including control rod drive mechanism (CRDM) and thermocouple nozzles, was investigated. The studies for primary water stress corrosion cracking (PWSCC) characteristics of Alloy 600 and Alloy 182/82 were reviewed and summarized in terms of the crack initiation and crack growth rate. The studies on the boric acid corrosion (BAC) of low alloy steels were also included in this report. PWSCC was found to be the main failure mechanism of RVH CRDM nozzles, which are constituted with Alloy 600 base metal and Alloy 182 weld filler materials. Alloy 600 and Alloy 182/82 are very susceptible to intergranular SCC in the PWR environments. The PWSCC crack initiation and growth features in the fusion zone of Alloy 182/82 were strongly dependant on solidification anisotropy during welding, test temperature, weld heat, mechanical loading, stress relief heat treatment, cold work and so on. BAC of low alloy steels is a wastage phenomenon due to general corrosion occurring on the over-all surface area of material. Systematic studies, concerned with structural integrity of RVH penetration nozzles as well as improvement of PWSCC resistance of nickel-based weld metals in the simulated PWR environment, are needed

Assessment of pressurized water reactor control rod drive mechanism nozzle cracking

International Nuclear Information System (INIS)

Shah, V.N.; Ware, A.G.; Porter, A.M.

1994-10-01

This report surveys the field experience related to cracking of pressurized water reactor (PWR) control rod drive mechanism nozzles (Alloy 600 material); evaluates design, fabrication, and operating conditions for the nozzles in US PWR; and evaluates the safety significance of nozzle cracking. Inspection at 78 overseas and one US PWR has revealed mainly axial cracks in 101 nozzles. The cracking is caused by primary water stress corrosion cracking, which requires the simultaneous presence of high tensile stresses, high operating temperatures, and susceptible microstructure. CRDM nozzle cracking is not a short-term safety issue. An axial crack is not likely to grow above the vessel head to a critical length because the stresses are not high enough to support the growth away from the attachment weld. Primary coolant leaking through an axial crack could cause a short circumferential crack on the outside surface. However, this crack is not likely to propagate through the nozzle wall to cause rupture. Leakage of the primary coolant from a through-wall crack could cause boric acid corrosion of the vessel head and challenge the structural integrity of the head, but it is very unlikely that the accumulated deposits of boric acid crystals resulting from such leakage could remain undetected

In service inspection of the reactor pressure vessel coolant and moderator nozzles at Atucha 1. 1998/1999 outages

International Nuclear Information System (INIS)

Antonaccio, Carlos; Conde, Alberto; Fittipaldi, Andres H.; Maniotti, Jorge; Moliterno, Gabriel E.

2000-01-01

During the August 1998 and the August 1999 Atucha 1 outages, two areas were inspected on the Reactor Pressure Vessel: the nozzle inner radii and the nozzle shell welds on all 3 moderator nozzles and all 4 main coolant nozzles. The inspections themselves were carried out by Mitsui Babcock Energy Limited from Scotland. The coordination, maintenance assistant and mounting of the manipulator devices over the nozzles were carried out by NASA personnel. Although it was not the first time the nozzle shell welds were inspected, due to the technologies advances in the ultrasonic field and in the inspection manipulators (magnetic ones), it was possible to inspect more volume than in previous inspections. In the other hand, it was the first time NASA was able to inspect the inner radii. In this last case the mayor problems to inspect them were the nozzles geometry and the small space available to install manipulators. The result of the inspections were: 1) There were no reportable indications at any of the inner radii inspected; 2) The inspection of nozzle to shell welds in main-coolant nozzles R3 and R4 detected flaws (one in each nozzle) which were reported as exceeding the dimensions specified as the acceptance level under Table IWB 3512-1, Section XI of the ASME code. Subsequent analysis requested by NASA and performed by Mitsui Babcock, demonstrated that the flaws were over dimensioned and could be explained as due to 'point' flaws. The analysis was based on theoretical mathematic model and experimental trials. Therefore their dimension were under the acceptance level of the ASME XI code. Although the Mitsui Babcock analysis, and at the same time it was in progress, it was assumed that the flaws were as they were originally presented (exceeding the acceptance level). NASA asked SIEMENS/KWU, the designer of the plant, to perform the fracture assessment according to ASME XI App. A. The assessment shows that the expected crack growth is negligibly small and the safety

Interpretation of Strain Measurements on Nuclear Pressure Vessels

DEFF Research Database (Denmark)

Andersen, Svend Ib Smidt; Engbæk, Preben

1980-01-01

with a negligible zeroshift. However, deviations from linear behaviour are observed in several cases. This nonlinearity can be explained by friction (flange connections) or by gaps (concentrical nozzles) in certain regions, whereas local plastic deformations during the first pressure loadings of the vessel seem...

Experiments and analysis of thermal stresses around the nozzle of the reactor vessel

International Nuclear Information System (INIS)

Song, D.H.; Oh, J.H.; Song, H.K.; Park, D.S.; Shon, K.H.

1981-01-01

This report describes the results of analysis and experiments on the thermal stress around the reactor vessel nozzle performed to establish a capability of thermal stress analysis of pressure vessel subjected to thermal loadings. Firstly, heat conduction analysis during reactor design transients and analysis on the experimental model were performed using computer code FETEM-1 for the purpose of verification of FETEM-1 which was developed in 1979 and will be used to obtain the temperature distribution in a solid body under the steady-state and the transient conditions. The results of the analysis was compared to the results in the Stress Report of Kori-1 reactor vessel and those from experiments on the model, respectively

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-06-01

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

Reactor vessel nozzle cracks: a photoelastic study

International Nuclear Information System (INIS)

Smith, C.W.

1979-01-01

A method consisting of a marriage between the ''frozen stress'' photoelastic approach and the local stress field equations of linear elastic fracture mechanics for estimating stress intensity factor distributions in three dimensional, finite cracked body problems is reviewed and extensions of the method are indicated. The method is then applied to the nuclear reactor vessel nozzle corner crack problem for both Intermediate Test Vessel and Boiling Water Reactor geometries. Results are compared with those of other investigators. 35 refs

Appropriate nominal stresses for use with ASME Code pressure-loading stress indices for nozzles

International Nuclear Information System (INIS)

Rodabaugh, E.C.

1976-06-01

This program is part of a cooperative effort with industry to develop and verify analytical methods for assessing the safety of nuclear pressure-vessel and piping-system design. The study of nominal stresses and stress indices described is part of a continuing study of design rules for nozzles in pressure vessels being coordinated by the PVRC Subcommittee on Reinforced Openings and External Loadings. Results from these studies are used by appropriate ASME Code groups in drafting new and improved design rules

Bursting tests on pressure vessels with cracks differing in configuration and location

International Nuclear Information System (INIS)

Stahlberg, R.

1978-01-01

For assessing the safety of nuclear pressure vessels exhibiting cracks, bursting test were carried out on a series of medium-size pressure vessels with and without welded nozzles and exhibiting cracks differing in configuration and location. The linear-elastic approach proved to be sufficiently accurate for straight strain conditions up to the onset of general yielding. Other analytical methods were successfully used to cover the plastic region. (orig.) [de

Development of Reactor Vessel Bottom Mount Instrumentation Nozzle Routine Inspection Device

Energy Technology Data Exchange (ETDEWEB)

Khaled, Atya Ahmed Abdallah; Ihn, Namgung [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2015-10-15

The primary coolant water of pressurized water reactors has created cracks in j-weld of penetrations with Alloy 600 through a process called primary water stress corrosion cracking. On October 6, 2013, BMI nozzle number 3 at Palo Verde Unit 3 (PVNGS-3) exhibited small white de-posits around the annulus. Nozzle attachment to the RV lower head is by J-groove weld to the inside penetration of the nozzle and the weld material is of Alloy 600 material. Above two cases clearly show the necessity of routine inspection of RV lower head penetration during refueling outage. Nondestructive inspection is generally performed to detect fine cracks or defects that may develop during operation. Defects usually occur at weld regions, hence most non-destructive inspection is to scan and check any defects or crack in the weld region. BMI nozzles at the bottom head of a nuclear reactor vessel (RV) are one of such area for inspection. But BMI nozzles have not been inspected during regular refuel outage due to the relative small size of BMI nozzle and limited impact of the consequences of BMI leak. However, there is growing concern since there have been leaks at nuclear power plants (NPPs) as well as recent operating experience. In this study, we propose a system that is conveniently used for nondestructive inspection of BMI nozzles during regular refueling outage without removing all the reactor internals. A 3D model of the inspection system was also developed along with the RV and internals which permits a virtual 3D simulation to check the design concept and usability of the system.

Analysis of LOFT pressurizer spray and surge nozzles to include a 4500F step transient

International Nuclear Information System (INIS)

Nitzel, M.E.

1978-01-01

This report presents the analysis of the LOFT pressurizer spray and surge nozzles to include a 450 0 F step thermal transient. Previous analysis performed under subcontract by Basic Technology Incorporated was utilized where applicable. The SAASIII finite element computer program was used to determine stress distributions in the nozzles due to the step transient. Computer results were then incorporated in the necessary additional calculations to ascertain that stress limitations were not exceeded. The results of the analysis indicate that both the spray and surge nozzles will be within stress allowables prescribed by subsubarticle NB-3220 of the 1974 edition of the ASME Boiler and Pressure Vessel Code when subjected to currently known design, normal operating, upset, emergency, and faulted condition loads

Nuclear reactor vessel decontamination systems

International Nuclear Information System (INIS)

McGuire, P. J.

1985-01-01

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

Reactor vessel sealing plug

International Nuclear Information System (INIS)

Dooley, R.A.

1986-01-01

This invention relates to an apparatus and method for sealing the cold leg nozzles of a nuclear reactor pressure vessel from a remote location during maintenance and inspection of associated steam generators and pumps while the pressure vessel and refueling canal are filled with water. The apparatus includes a sealing plug for mechanically sealing the cold leg nozzle from the inside of a reactor pressure vessel. The sealing plugs include a primary and a secondary O-ring. An installation tool is suspended within the reactor vessel and carries the sealing plug. The tool telescopes to insert the sealing plug within the cold leg nozzle, and to subsequently remove the plug. Hydraulic means are used to activate the sealing plug, and support means serve to suspend the installation tool within the reactor vessel during installation and removal of the sealing plug

Residual Stress Evaluation of Weld Inlay Process on Reactor Vessel Nozzles

Energy Technology Data Exchange (ETDEWEB)

Cho, Kihyun; Cho, Hong Seok [KEPCO KPS, Naju (Korea, Republic of)

2015-10-15

Weld overlay, weld inlay and stress improvement are mitigation technologies for butt joints. Weld overlay is done on pressurizer nozzles which are the highest potential locations occurring PWSCC due to high temperature in Korea. Reactor vessel nozzles are other big safety concerns for butt joints. Weld overlay and stress improvement should be so difficult to apply to those locations because space is too limited. Weld inlay should be one of the solutions. KEPCO KPS has developed laser welding system and process for reactor nozzles. Welding residual stress analysis is necessary for flaw evaluation. United States nuclear regulatory commission has calculated GTAW(Gas Tungsten Arc Welding) residual stress using ABAQUS. To confirm effectiveness of weld inlay process, welding residual stress analysis was performed. and difference between GTAW and LASER welding process was compared. Evaluation of weld inlay process using ANSYS and ABAQUS is performed. All of the both results are similar. The residual stress generated after weld inlay was on range of 450-500 MPa. Welding residual stresses are differently generated by GTAW and LASER welding. But regardless of welding process type, residual tensile stress is generated on inside surface.

Fracture analyses and test of regions with nozzle and hole and curvature influence in nuclear vessel

International Nuclear Information System (INIS)

Wang Baisong; Xu Dinggen; Ye Weijuan; Hu Yinbiao; Liang Xingyun; Gu Shaode; Zhou Peiying

1993-08-01

For the calculations of stress intensity factor K 1 of surface crack in the regions with nozzle and hole and the curvature influence on nuclear vessel, a improved 3-D collapsed isoparametric singular element with quarter-points was presented. The square root singularity in the vertical planes of crack was derived. The methods of transitional element and calculating K 1 from displacements were extensively used in 3- D case. The SIF K 1 of the corner crack in inner wall of the nozzle of RPV (reactor pressure vessel) for a typical 300 MW nuclear plant was calculated, and it was verified by 3-D photo-elastic test and diffusion of light test. The engineering fracture analysis and evaluation of the outside surface crack in the circular are transitional region of the head flange of RPV are also completed

Stress categorization in nozzle to pressure vessel connections finite elements models; Categorizacao de tensoes em modelos de elementos finitos de conexoes bocal-vaso de pressao

Energy Technology Data Exchange (ETDEWEB)

Albuquerque, Levi Barcelos de

1999-07-01

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

Thermal and stress analyses of the reactor pressure vessel lower head of the Three Mile Island Unit 2

International Nuclear Information System (INIS)

Hashimoto, K.; Onizawa, K.; Kurihara, R.; Kawasaki, S.; Soda, K.

1992-01-01

Thermal and stress analyses were performed using the finite element analysis code ABAQUS to clarify the factors which caused tears in the stainless steel liner of the reactor pressure vessel lower head of the Three Mile Island Unit 2 (TMI-2) reactor pressure vessel during the accident on 28 March 1979. The present analyses covered the events which occurred after approximately 20 tons of molten core material were relocated to the lower head of the reactor pressure vessel. They showed that the tensile stress was highest in the case where the relocated core material consisting of homogeneous UO 2 debris was assumed to attack the lower head and the debris was then quenched. The peak tensile stress was in the vicinity of the welded zone of the penetration nozzle. This result agrees with the findings from the examination of the TMI-2 reactor pressure vessel that major tears in the stainless steel liner were observed around two penetration nozzles of the lower head. (author)

Inspection of dissimilar metal welds in reactor pressure vessels in Spain

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-01

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

Inspection of dissimilar metal welds in reactor pressure vessels in Spain

International Nuclear Information System (INIS)

Gadea, J.R.; Regidor, J.J.; Pelaez, J.A.; Serrano, P.

2011-01-01

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

Remotely replaceable fuel and feed nozzles for the NWCF calciner vessel

International Nuclear Information System (INIS)

Fletcher, R.D.; Carter, J.A.; May, K.W.

1978-01-01

The development and testing of remotely replaceable fuel and feed nozzles for calcination of liquid radioactive wastes in the calciner vessel of the New Waste Calcining Facility (NWCF) being built at the Idaho National Engineering Laboratory are described. A complete fuel nozzle assembly was fabricated and tested at the Remote Maintenance Development Facility to evolve design refinements, identify required support equipment, and develop handling techniques. The design also provided for remote replacement of the nozzle support carriage and adjacent feed and fuel pipe loops using two pairs of master-slave manipulators

Minimum weight designs for reinforcement of spherical pressure vessels with flush radial nozzles

International Nuclear Information System (INIS)

Yeo, K.T.; Robinson, M.

1978-01-01

A cylinder-sphere pressure vessel, reinforced in the sphere by a section of constant thickness, has been analysed from the point of view of minimum weight. The reinforcement is allowed to be offset from the main sphere and the design has to be such that the test pressure of the vessel equals the limit pressure. It is shown that in most circumstances an economy of weight may be obtained by making the reinforcement thicker, but less extensive, than suggested in a previous proposal. Further benefit can be obtained by offsetting the reinforcement radially outwards so that the inside surfaces of main sphere and reinforcement are flush. (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)

Design criteria for piping and nozzles program

International Nuclear Information System (INIS)

Moore, S.E.; Bryson, J.W.

1977-01-01

This report reviews the activities and accomplishments of the Design Criteria for Piping and Nozzles program being conducted by the Oak Ridge National Laboratory for the period July 1, 1975, to September 30, 1976. The objectives of the program are to conduct integrated experimental and analytical stress analysis studies of piping system components and isolated and closely-spaced pressure vessel nozzles in order to confirm and/or improve the adequacy of structural design criteria and analytical methods used to assure the safe design of nuclear power plants. Activities this year included the development of a finite-element program for analyzing two closely spaced nozzles in a cylindrical pressure vessel; a limited-parameter study of vessels with isolated nozzles, finite-element studies of piping elbows, a fatigue test of an out-of-round elbow, summary and evaluation of experimental studies on the elastic-response and fatigue failure of tees, parameter studies on the behavior of flanged joints, publication of fifteen topical reports and papers on various experimental and analytical studies; and the development and acceptance of a number of design rules changes to the ASME Code. 2 figures, 2 tables

Evaluation of J-groove weld residual stress and crack growth rate of PWSCC in reactor pressure vessel closure head

Energy Technology Data Exchange (ETDEWEB)

Oh, Seung Hyuk; Ryu, Tae Young; Park, Seung Hyun; Won, Min Gu; Kang, Seok Jun; Kim, Moon Ki; Choi, Jae Boong [Sungkyunkwan University, Suwon (Korea, Republic of); Lee, Kyoung Soo; Lee, Sung Ho [Korea Hydro and Nuclear Power, Daejeon (Korea, Republic of)

2015-03-15

Over the last decade, primary water stress corrosion cracking (PWSCC) has been frequently found in pressurized water reactor (PWR) applications. Especially, PWSCC has occurred in long-term operated PWRs. As this phenomenon leads to serious accidents, we must be beforehand with the anticipated problems. A typical PWR consists of J-groove welded components such as reactor pressure vessel closure head and nozzles. Reactor pressure vessel closure head is made of SA508 and it is covered by cladding. Alloy 600 is used for nozzles. And J-groove weld is conducted with alloy 82/182. Different material properties of these metals lead to residual stress and PWSCC consequentially. In this study, J-groove weld residual stress was investigated by a three-dimensional finite element analysis with an actual asymmetric J-groove weld model and process of construction. Also crack growth rate of PWSCC was evaluated from cracks applied on the penetration nozzles. Based on these two values, one cannot only improve the structural integrity of PWR, but also explain PWSCC behavior such that high residual stress at the J-groove weld area causes crack initiation and propagation through the surface of nozzles. In addition, crack behavior was predicted at the various points around the nozzle.

Evaluation of J-groove weld residual stress and crack growth rate of PWSCC in reactor pressure vessel closure head

International Nuclear Information System (INIS)

Oh, Seung Hyuk; Ryu, Tae Young; Park, Seung Hyun; Won, Min Gu; Kang, Seok Jun; Kim, Moon Ki; Choi, Jae Boong; Lee, Kyoung Soo; Lee, Sung Ho

2015-01-01

Over the last decade, primary water stress corrosion cracking (PWSCC) has been frequently found in pressurized water reactor (PWR) applications. Especially, PWSCC has occurred in long-term operated PWRs. As this phenomenon leads to serious accidents, we must be beforehand with the anticipated problems. A typical PWR consists of J-groove welded components such as reactor pressure vessel closure head and nozzles. Reactor pressure vessel closure head is made of SA508 and it is covered by cladding. Alloy 600 is used for nozzles. And J-groove weld is conducted with alloy 82/182. Different material properties of these metals lead to residual stress and PWSCC consequentially. In this study, J-groove weld residual stress was investigated by a three-dimensional finite element analysis with an actual asymmetric J-groove weld model and process of construction. Also crack growth rate of PWSCC was evaluated from cracks applied on the penetration nozzles. Based on these two values, one cannot only improve the structural integrity of PWR, but also explain PWSCC behavior such that high residual stress at the J-groove weld area causes crack initiation and propagation through the surface of nozzles. In addition, crack behavior was predicted at the various points around the nozzle.

Novel design for transparent high-pressure fuel injector nozzles

Science.gov (United States)

Falgout, Z.; Linne, M.

2016-08-01

The efficiency and emissions of internal combustion (IC) engines are closely tied to the formation of the combustible air-fuel mixture. Direct-injection engines have become more common due to their increased practical flexibility and efficiency, and sprays dominate mixture formation in these engines. Spray formation, or rather the transition from a cylindrical liquid jet to a field of isolated droplets, is not completely understood. However, it is known that nozzle orifice flow and cavitation have an important effect on the formation of fuel injector sprays, even if the exact details of this effect remain unknown. A number of studies in recent years have used injectors with optically transparent nozzles (OTN) to allow observation of the nozzle orifice flow. Our goal in this work is to design various OTN concepts that mimic the flow inside commercial injector nozzles, at realistic fuel pressures, and yet still allow access to the very near nozzle region of the spray so that interior flow structure can be correlated with primary breakup dynamics. This goal has not been achieved until now because interior structures can be very complex, and the most appropriate optical materials are brittle and easily fractured by realistic fuel pressures. An OTN design that achieves realistic injection pressures and grants visual access to the interior flow and spray formation will be explained in detail. The design uses an acrylic nozzle, which is ideal for imaging the interior flow. This nozzle is supported from the outside with sapphire clamps, which reduces tensile stresses in the nozzle and increases the nozzle's injection pressure capacity. An ensemble of nozzles were mechanically tested to prove this design concept.

Equivalent nozzle in thermomechanical problems

International Nuclear Information System (INIS)

Cesari, F.

1977-01-01

When analyzing nuclear vessels, it is most important to study the behavior of the nozzle cylinder-cylinder intersection. For the elastic field, this analysis in three dimensions is quite easy using the method of finite elements. The same analysis in the non-linear field becomes difficult for designs in 3-D. It is therefore necessary to resolve a nozzle in two dimensions equivalent to a 3-D nozzle. The purpose of the present work is to find an equivalent nozzle both with a mechanical and thermal load. This has been achieved by the analysis in three dimensions of a nozzle and a nozzle cylinder-sphere intersection, of a different radius. The equivalent nozzle will be a nozzle with a sphere radius in a given ratio to the radius of a cylinder; thus, the maximum equivalent stress is the same in both 2-D and 3-D. The nozzle examined derived from the intersection of a cylindrical vessel of radius R=191.4 mm and thickness T=6.7 mm with a cylindrical nozzle of radius r=24.675 mm and thickness t=1.350 mm, for which the experimental results for an internal pressure load are known. The structure was subdivided into 96 finite, three-dimensional and isoparametric elements with 60 degrees of freedom and 661 total nodes. Both the analysis with a mechanical load as well as the analysis with a thermal load were carried out on this structure according to the Bersafe system. The thermal load consisted of a transient typical of an accident occurring in a sodium-cooled fast reactor, with a peak of the temperature (540 0 C) for the sodium inside the vessel with an insulating argon temperature constant at 525 0 C. The maximum value of the equivalent tension was found in the internal area at the union towards the vessel side. The analysis of the nozzle in 2-D consists in schematizing the structure as a cylinder-sphere intersection, where the sphere has a given relation to the

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

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.

Device for positioning ultrasonic probes and/or television cameras on the outer surface of reactor pressure vessels

International Nuclear Information System (INIS)

Zipser, R.; Dose, G.F.

1977-01-01

The device makes possible periodical in-service inspections of welding seams and material of a reactor pressure vessel without local human presence. A 'support ring' encloses the pressure vessel in a horizontal plane with free space. It is vertically moved up and down in the space between pressure vessel and thermal shield by means of tackles. At a control desk placed in a protected area its movement is controlled and its vertical position is indicated. A 'rotating track' with its own drive is rotating remote-controlled on the 'support ring'. By a combination of the vertical with the rotating movement, an ultrasonic probe placed removably on the 'rotating hack', or a television camera will be brought to any position on the cylindrical circumference of the pressure vessel. Special devices extend the radius of action, in upward direction for inspecting the welding seams of the coolant nozzles, and in downward direction for the inspection of welds on the hemispherical bottom of the pressure vessel or on the outlet pipe nozzle placed there. The device remains installed during reactor operation, but is moved down to the lower horizontal surface of the thermal shield. Parts which are sensible to radiation like probes or television cameras and special devices will then be removed respectively mounted before beginning an inspection compaign. This position may be reached by the lower access in the biological shield and through an opening in the horizontal surface of the thermal shield. (HP) [de

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

Reactor vessel sealing plug

International Nuclear Information System (INIS)

Dooley, R.A.

1986-01-01

An apparatus is described for sealing a cold leg nozzle of a nuclear reactor pressure vessel from a remote location comprising: at least one sealing plug for mechanically sealing the nozzle from the inside of the reactor pressure vessel. The sealing plug includes a plate and a cone assembly having an end part receptive in the nozzle, the plate being axially moveable relative to the cone assembly. The plate and cone assembly have confronting bevelled edges defining an opening therebetween. A primary O-ring is disposed about the opening and is supported on the bevelled edges, the plate being guidably mounted to the cone assembly for movement toward the cone assembly to radially expand the primary O-ring into sealing engagement with the nozzle. A means is included for providing relative movement between the outer plate and the cone assembly

Study on steam pressure characteristics in various types of nozzles

Science.gov (United States)

Firman; Anshar, Muhammad

2018-03-01

Steam Jet Refrigeration (SJR) is one of the most widely applied technologies in the industry. The SJR system was utilizes residual steam from the steam generator and then flowed through the nozzle to a tank that was containing liquid. The nozzle converts the pressure energy into kinetic energy. Thus, it can evaporate the liquid briefly and release it to the condenser. The chilled water, was produced from the condenser, can be used to cool the product through a heat transfer process. This research aims to study the characteristics of vapor pressure in different types of nozzles using a simulation. The Simulation was performed using ANSYS FLUENT software for nozzle types such as convergent, convrgent-parallel, and convergent-divergent. The results of this study was presented the visualization of pressure in nozzles and was been validated with experiment data.

Remotely replaceable fuel and feed nozzles for the new waste calcining facility calciner vessel

International Nuclear Information System (INIS)

Fletcher, R.D.; Carter, J.A.; May, K.W.

1978-01-01

The development and testing of remotely replaceable fuel and feed nozzles for calcination of liquid radioactive wastes in the calciner vessel of the New Waste Calcining Facility being built at the Idaho National Engineering Laboratory is described. A complete fuel nozzle assembly was fabricated and tested at the Remote Maintenance Development Facility to evolve design refinements, identify required support equipment, and develop handling techniques. The design also provided for remote replacement of the nozzle support carriage and adjacent feed and fuel pipe loops using two pairs of master-slave manipulators

Behaviour of a pressure vessel nozzle with thermo-sleeve under thermal loading induced by stratified flow

International Nuclear Information System (INIS)

Kussmaul, K.; Mayinger, W.; Diem, H.; Katzenmeier, G.

1993-01-01

Startup at low reactor power may give rise to stratified flow conditions in pipes of boiling water and pressurized water reactors. Stratified flow regimes cause a steep temperature gradient between the cold and the hot fluid layer. This temperature gradient produces high axial stresses which, in the case of intermittent feeding of cold water and an appropriate number of repetitions, in principle may initiate cracking in the feedwater pipe and close to the feeding nozzle. Thermosleeves have been installed in a number of reactors to mitigate thermally induced stresses; they reduce the intensity of thermal transients by means of an insulating fluid annulus developing between the sleeve and the nozzle, in order to measure the temperature and stress gradients occurring in the region of the nozzle edge, the so-called TEMS experiments were carried out under realistic operating conditions, and with different cold water levels within the framework of German research activities in the field of reactor safety at the HDR test facility. The experiments served to simulate the physics phenomena by means of a FE-program and to verify the computational approach by comparisons of measurements and calculations

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)

Experiences concerning reactor pressure vessel head penetration inspections; Erfahrungen mit Pruefungen von Reaktordruckbehaelter-Deckeldurchfuehrungen

Energy Technology Data Exchange (ETDEWEB)

Debnar, Angelika [Westinghouse Electric Germany GmbH, Mannheim (Germany)

2009-07-01

Globally observed damage at the control rod drive mechanism nozzles in PWR-type reactors (Bugey-3, Oconee 1,2,3 and ANO-1, David Besse) have triggered enhanced inspection of reactor pressure vessel (RPV) head penetrations. In Germany the regulations require a periodic inspection especially of dissimilar welds. Westinghouse has developed an automated measuring system for RPV heads aimed to inspect welded joints at open nozzles of nozzles with thermosleeves. The testing technology with remote controlled robotics is supposed to perform a weld inspection as complete as possible, restraints result from constructive difficulties for the accessibility. The new gap-scanner DE2008 was qualified at the mock-up and was implemented into the periodic in-service inspection of Neckarwestheim-1.

Crack of reactor vessel upper head penetration nozzles in Korean nuclear plants

Energy Technology Data Exchange (ETDEWEB)

Doh, E.; Lee, T-S.; Kim, J-Y.; Lee, C-H. [KEPCO Plant Service and Engineering Co., Ltd., Busan (Korea, Republic of)

2014-07-01

Since the first CRDM nozzles of reactor vessel head at Kori unit 1 in Korea were inspected in 2003, no CRDM nozzle cracks had been revealed prior to the inspection at Hanbit unit 3 in October 2012, even though many foreign plants had been reporting PWSCC cracks. In October 2012, seven axial cracks from 6 CRDM nozzles at Hanbit unit 3, and in November 2013, six axial cracks from 6 CRDM nozzles at Hanbit unit 4 were detected by TOFD Ultrasonic testing from ID of nozzles. There were confirmed to be PWSCC by Dye penetrant testing and Replica on the surface of J-groove weld of CRDM nozzles. Both plants are OPR-1000 types. All flaws started from the surface of J-groove weld at interface with OD of nozzle, but did not grow up to the top of J-groove weld, and did not make any Leak path up to head outside. The Performance Demonstration Initiative (PDI) system of CRDM nozzle inspection for Westinghouse type plants has been applied in Korea since July 2011. However, its application for OPR-1000 is still under development in Korea. The experience of PDI inspection for Westinghouse type plant contributed greatly to the detection and evaluation of PWSCC of CRDM nozzles at OPR- 1000 of Hanbit unit 3 & 4. The experimentally based procedure of flaw detection and the enhanced detection technique of examiners made it possible to detect and to determine the PWSCC indications. Embedded Flaw Repair process was approved by government authority, and the repair of the 6 CRDM nozzles in each plant was conducted by a consortium of Westinghouse and KPS. (author)

Experiments on the spray nozzles used in the pressurizer of power reactor

International Nuclear Information System (INIS)

Diao Wentang

1989-04-01

The spray nozzle, which is used in the pressurizer of pressurized water reactor system, usually uses a less differential pressure between the reactor inlet and outlet as the spray drive pressure, but its flow rate is relatively larger. It is difficult to obtain a optimum spray performance of such a nozzle. The experimental results of five types of twenty seven spray nozzles in different structures and sizes with the range of the spray drive pressure from 0.127 to 0.245 MPa and the flow rates from 5 to 50 t/h are given. The main factors affecting spray performances and their distribution characteristics have been found. And some relatively suitable spray structures have been recommended, which can be used as references for improving the spray nozzles used in the pressurizers of existing PWRs or of the PWRs to be built

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

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 ...

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-12-15

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Stress intensities for nozzle cracks in reactor vessels. Reporting period, January 1, 1976--October 31, 1976

International Nuclear Information System (INIS)

Smith, C.W.; Jolles, M.; Peters, W.H.

1976-11-01

A series of six frozen stress photoelastic tests was conducted to investigate the distribution of stress intensity factor (SIF) along a crack which occurred at the juncture of a pipe (nozzle) with a cylindrical pressure vessel. Typical photoelastic fringe patterns are shown for slices which were taken mutually orthogonal to the flaw border and the flaw surface. A typical plot of normalized apparent SIF versus square root of normalized distance from the crack tip is presented. The variation in SIF along the flaw border is given for all six different crack geometries and also the variation of SIF with varying a/T is presented. 40 references

Nozzle erosion characterization and minimization for high-pressure rocket motor applications

Science.gov (United States)

Evans, Brian

Understanding of the processes that cause nozzle throat erosion and developing methods for mitigation of erosion rate can allow higher operating pressures for advanced rocket motors. However, erosion of the nozzle throat region, which is a strong function of operating pressure, must be controlled to realize the performance gains of higher operating pressures. The objective of this work was the study the nozzle erosion rates at a broad range of pressures from 7 to 34.5 MPa (1,000 to 5,000 psia) using two different rocket motors. The first is an instrumented solidpropellant motor (ISPM), which uses two baseline solid propellants; one is a non-metallized propellant called Propellant S and the other is a metallized propellant called Propellant M. The second test rig is a non-metallized solid-propellant rocket motor simulator (RMS). The RMS is a gas rocket with the ability to vary the combustion-product species composition by systematically varying the flow rates of gaseous reactants. Several reactant mixtures were utilized in the study to determine the relative importance of different oxidizing species (such as H2O, OH, and CO2). Both test rigs are equipped with a windowed nozzle section for real-time X-ray radiography diagnostics of the instantaneous throat variations for deducing the instantaneous erosion rates. The nozzle test section for both motors can also incorporate a nozzle boundary-layer control system (NBLCS) as a means of nozzle erosion mitigation. The effectiveness of the NBLCS at preventing nozzle throat erosion was demonstrated for both the RMS and the ISPM motors at chamber pressures up to 34 MPa (4930 psia). All tests conducted with the NBLCS showed signs of coning of the propellant surface, leading to increased mass burning rate and resultant chamber pressure. Two correlations were developed for the nozzle erosion rates from solid propellant testing, one for metallized propellant and one for non-metallized propellants. The non-metallized propellant

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

Learning from EDF investigations on SG divider plates and vessel head nozzles. Evidence of prior deformation effect on stress corrosion cracking

International Nuclear Information System (INIS)

Deforge, D.; Duisabeau, L.; Miloudi, S.; Thebault, Y.; Couvant, T.; Vaillant, F.; Lemaire, E.

2011-01-01

Nickel Based alloys Stress Corrosion Cracking (SCC) has been a major concern for all the Nuclear Power Plants (NPP) utilities since the beginning of the seventies. At EDF, the nineties were marked by the occurrence of cracks on vessel head nozzles. These cracks were responsible for a leak at Bugey 3 vessel head, which was the precursor leading to the replacement of all vessel heads. From 2002, new cases of Stress Corrosion Cracking were reported on Steam Generator (SG) Divider Plates (SGDP) welded junctions. These cracks are periodically inspected inservice and reparations could be performed in case of a significant evolution of the phenomenon even if the safety issue is less relevant than for the vessel head nozzles. Both issues have led to an important non-destructive testing (NDT) program and to destructive investigations campaigns. NDT were performed on an exhaustive basis for all vessel head nozzles and for all the divider plates of 900 MWe plants. Destructive investigations were performed on more than 30 vessel head nozzles and on 6 divider plates. The last investigations were performed on samples from two decommissioned Steam Generators of Chinon B1 which present SCC cracks. In this paper, the main conclusions driven from the analysis of both NDT and destructive investigation results are reported and a comparison of the behaviours of divider plates and vessel head nozzles is given. Results give evidence that prior plastic deformation of the components before operation is fundamental for the further environmental behaviour of the material. Analysis of field experience based on parameters characteristics of prior deformation and parameters characteristics of material microstructure can be used to account for the components which are the most sensitive to SCC cracking. Some perspectives on SCC predictive models are also presented. (authors)

Determination of two dimensional axisymmetric finite element model for reactor coolant piping nozzles

International Nuclear Information System (INIS)

Choi, S. N.; Kim, H. N.; Jang, K. S.; Kim, H. J.

2000-01-01

The purpose of this paper is to determine a two dimensional axisymmetric model through a comparative study between a three dimensional and an axisymmetric finite element analysis of the reactor coolant piping nozzle subject to internal pressure. The finite element analysis results show that the stress adopting the axisymmetric model with the radius of equivalent spherical vessel are well agree with that adopting the three dimensional model. The radii of equivalent spherical vessel are 3.5 times and 7.3 times of the radius of the reactor coolant piping for the safety injection nozzle and for the residual heat removal nozzle, respectively

Evaluation of flaws or service induced cracks in pressure vessels

International Nuclear Information System (INIS)

Riccardella, P.C.; Copeland, J.F.; Gilman, J.

1987-01-01

An overview of the ASME flaw evaluation procedures for nuclear pressure vessels is presented, with emphasis on fatigue crack growth evaluations. Environmental and load-rate effects are further considered with respect to new crack growth data and a time-dependent crack growth model. This new crack growth model is applied to evaluate feedwater nozzle cracking in boiling water reactors and is compared to current and past ASME crack growth curves. The time-dependent model bounds the observed cracking and indicates that more detailed consideration of material susceptibility, in terms of sulfur content and product form, is needed

Review of in-service thermal annealing of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Server, W.L.

1984-01-01

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. A test reactor pressure vessel has been wet annealed at less than 343 0 C (650 0 F), and annealing of the Belgian BR-3 reactor vessel has recently taken place. An industry survey indicates that dry annealing a reactor vessel in-place is feasible, but solvable engineering problems do exist. The materials with highest radiation sensitivity in the older reactor vessels are submerged-arc weld metals with high copper and nickel concentrations. The limited Charpy V-notch and fracture toughness data available for five such welds were reviewed. The review suggested that significant recovery results from annealing at 454 0 C (850 0 F) for one week. Two of the main concerns with a localized heat treatment at 454 0 C (850 0 F) are the degree of distortion that may occur after the annealing cycle and the extent of residual stresses. A thermal and structural analysis of a reactor vessel for distortions and residual stresses found no problems with the reactor vessel itself but did indicate a rotation at the nozzle region of the vessel that would plastically deform the attached primary piping. Further analytical studies are needed. An American Society for Testing and Materials (ASTM) task group is upgrading and revising the ASTM Recommended Guide for In-Service Annealing of WaterCooled Nuclear Reactor Vessels (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 (for example, the American Society of Mechanical Engineers Boiler and Pressure Vessel Code body)

High Pressure Water Stripping Using Multi-Orifice Nozzles

Science.gov (United States)

Hoppe, David

1999-01-01

The use of multi-orifice rotary nozzles greatly increases the speed and stripping effectiveness of high pressure water blasting systems, but also greatly increases the complexity of selecting and optimizing the operating parameters. The rotational speed of the nozzle must be coupled with its transverse velocity as it passes across the surface of the substrate being stripped. The radial and angular positions of each orifice must be included in the analysis of the nozzle configuration. Orifices at the outer edge of the nozzle head move at a faster rate than the orifices located near the center. The energy transmitted to the surface from the impact force of the water stream from an outer orifice is therefore spread over a larger area than energy from an inner orifice. Utilizing a larger diameter orifice in the outer radial positions increases the total energy transmitted from the outer orifice to compensate for the wider distribution of energy. The total flow rate from the combination of all orifices must be monitored and should be kept below the pump capacity while choosing orifice to insert in each position. The energy distribution from the orifice pattern is further complicated since the rotary path of all the orifices in the nozzle head pass through the center section. All orifices contribute to the stripping in the center of the path while only the outer most orifice contributes to the stripping at the edge of the nozzle. Additional orifices contribute to the stripping from the outer edge toward the center section. With all these parameters to configure and each parameter change affecting the others, a computer model was developed to track and coordinate these parameters. The computer simulation graphically indicates the cumulative affect from each parameter selected. The result from the proper choices in parameters is a well designed, highly efficient stripping system. A poorly chosen set of parameters will cause the nozzle to strip aggressively in some areas

An analytical evaluation for the pressure drop characteristics of bottom nozzle flow holes

International Nuclear Information System (INIS)

Yang, S. G.; Kim, H. J.; Lim, H. T.; Park, E. J.; Jeon, K. L.

2002-01-01

An analytical evaluation for the bottom nozzle flow holes was performed to find a best design concept in terms of pressure drop. For this analysis, Computational Fluid Dynamics (CFD), FLUENT 5.5, code was selected as an analytical evaluation tool. The applicability of CFD code was verified by benchmarking study with Vibration Investigation of Small-scale Test Assemblies (VISTA) test data in several flow conditions and typical flow hole shape. From this verification, the analytical data were benchmarked roughly within 17% to the VISTA test data. And, overall trend under various flow conditions looked very similar between both cases. Based on the evaluated results using CFD code, it is concluded that the deburring and multiple chamfer hole features at leading edge are the excellent design concept to decrease pressure drop across bottom nozzle plate. The deburring and multiple chamfer hole features at leading edge on the bottom nozzle plate have 12% and 17% pressure drop benefit against a single chamfer hole feature on the bottom nozzle plate, respectively. These design features are meaningful and applicable as a low pressure drop design concept of bottom nozzle for Pressurized Water Reactor (PWR) fuel assembly

New intraocular pressure measurement method using reflected pneumatic pressure from cornea deformed by air puff of ring-type nozzle.

Science.gov (United States)

Kim, Hyung Jin; Seo, Yeong Ho; Kim, Byeong Hee

2017-01-01

In this study, a non-contact type intraocular pressure (IOP) measuring system using reflected pneumatic pressure is proposed to overcome the disadvantages of existing measurement systems. A ring-type nozzle, a key component in the proposed system, is designed via computational fluid analysis. It predicts the reflected pneumatic pressure based on the nozzle exit angle and inner and outer diameters of the nozzle, which are 30°, 7 mm, and 9 mm, respectively. Performance evaluation is conducted using artificial eyes fabricated using polydimethylsiloxane with the specifications of human eyes. The IOP of the fabricated artificial eyes is adjusted to 10, 30, and 50 mm Hg, and the reflected pneumatic pressure is measured as a function of the distance between the ring-type nozzle and artificial eye. The measured reflected pneumatic pressure is high when the measurement distance is short and eye pressure is low. The cornea of an artificial eye is significantly deformed at a low IOP, and the applied pneumatic pressure is more concentrated in front of the ring-type nozzle because of the deformed cornea. Thus, the reflected pneumatic pressure at a low IOP has more inflows into the pressure sensor inserted inside the nozzle. The sensitivity of the output based on the IOP at measurement distances between 3-5 mm is -0.0027, -0.0022, -0.0018, -0.0015, and -0.0012. Sensitivity decreases as the measurement distance increases. In addition, the reflected pneumatic pressure owing to the misalignment at the measurement distances of 3-5 mm is not affected within a range of 0.5 mm. Therefore, the measurement range is acceptable up to a 1 mm diameter from the center of an artificial eye. However, the accuracy gradually decreases as the reflected pneumatic pressure from a misalignment of 1 mm or more decreases by 26% or more.

Computation of principal stresses and stress intensity of a nozzle on a spherical pressure vessel

International Nuclear Information System (INIS)

Sun, B.C.; Lyow, B.L.; Koplik, B.

1993-01-01

This paper presents a Stress Computation Table that systematically computes the local stresses at various locations of the sphere-nozzle intersection. The six components of external loading are: radial load, two overturning moments, two horizontal shear forces, and a torsional moment. The radial and overturning moments induce local membrane and bending stresses in both the circumferential and meridional directions of the sphere around the nozzle. The shear forces and torsional moment produce local shear stresses. In addition, the shear forces induce local membrane and bending stresses around the nozzle. The local stress factors from each external loading component are taken from recent publications by Lyow, Sun and Koplik who have studied this subject through the use of the finite element method. These factors are a function of the nozzle-sphere geometrical parameters, beta, β, (nozzle radius/sphere radius) and gamma, γ, (sphere radius/thickness), with the beta value ranging from 0.1 to 0.5, and the gamma value ranging from 10 to 100. The Stress Table summarizes all the normal and shear stresses at eight different locations around the nozzle, and finally the principal stresses and stress intensity are computed. The stress factor plots from previous publications are replotted in this paper to provide a handy reference as well as consistency. A numerical sample employing a FORTRAN program is also given. (author)

A-508 class 3 forgings for pressure vessels

International Nuclear Information System (INIS)

Comon, J.

1977-01-01

The manufacture of the forged parts of the first PWR nuclear pressure vessel installed in France started in the Creusot-Loire's Forge Plant in 1961. Since this date, more than 300 forgings of this type were delivered (flanges, rings, zones, and nozzles). The major part of these forgings were made of Mn, Ni, Mo steel (SA 508 class 3). They represent a population large and homogeneous enough to attempt a statistical analysis of chemical and mechanical test results. The aim of this analysis was double: (1) a better knowledge of the scattering of the results and a better estimate of what can be introduced or accepted in a specification, and (2) the setting up of correlations existing between these results, particularly between chemical analysis and mechanical test results. In addition to this statistical analysis concerning industrial results, several laboratory studies are presented, giving a more complete characterization of SA 508 class 3. All these results form a very complete documentation showing that SA 508 class 3 steel is suitable for the manufacture of large forged vessels requiring a high degree of reliability

Inservice inspection of Halden BWR pressure vessel

International Nuclear Information System (INIS)

Foerli, O.; Hernes, T.

1978-01-01

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

Pressure Distribution and Performance Impacts of Aerospike Nozzles on Rotating Detonation Engines

Science.gov (United States)

2017-06-01

Nozzle Exit Plane at Various Pressure Ratios for the Quiescent Air Hydrogen Fuel Case, PRdesign = 10:1...81 Figure 55. Mach Number Distribution along the Nozzle Exit Plane at Various Pressure Ratios for the Supersonic...budget constraints, have spurred engineers to focus on improving the specific fuel consumption of these engines. One technology that promises

Development of user-friendly structural design system for pressure vessels

International Nuclear Information System (INIS)

Sato, Takuya; Nomoto, Taeko; Kado, Kenichiro; Yagawa, Genki; Yoshimura, Shinobu.

1996-01-01

In this paper we describe a new user-friendly structural design system for pressure vessels, which is based on finite element stress analyses. The basic concept of the developed system is to minimize input data required for the finite element analysis and to perform the analysis quickly. To realize this, the system is equipped with the finite element modeling module based on fuzzy knowledge processing, the input data generation module, the finite element analyzer, the graphic user-interface module for analysis results, and the stress evaluation module. Fundamental performance of the present system is clearly demonstrated through the analysis of a top nozzle. (author)

Stress analysis of PCV nozzle junction

International Nuclear Information System (INIS)

Uchiyama, Shoichi; Oikawa, Tsuneo; Hoshino, Seizo

1976-01-01

Most of various pressure vessels comprise each one cylindrical shell and one or more nozzles. In this study, in order to analyze the stress in the structures of this type as minutely and exactly as possible, the program for stress analysis by the finite element method was made, which is required for the strength analysis for three-dimensional structures. Especially, the problem of the stress distribution around nozzle junctions was solved theoretically with the program. The program for the analysis developed in this study is provided with various functions, such as the input generator for cylindrical, conical and spherical shells, and plotter, and is very covenient. The accuracy of analysis is very good. The method of analysis and the calculation of the rigidity matrices for the deformation in plane and bending are explained. The result of the stress analysis around the nozzle junctions of a containment vessel with this program was in good agreement with experimental data and the result with SAP-4 code, therefore the propriety of the calculated result with this program was proved. Also calculations were carried out on three cases, namely a flat plate fixed at one end with distributed load, a cylinder fixed at one end with internal pressure, and an I-beam fixed at one end with concentrated load. The calculated results agreed well with theoretical solutions in all cases. (Kako, I.)

Effect of carrier gas pressure on condensation in a supersonic nozzle

International Nuclear Information System (INIS)

Wyslouzil, B.E.; Wilemski, G.; Beals, M.G.; Frish, M.B.

1994-01-01

Supersonic nozzle experiments were performed with a fixed water or ethanol vapor pressure and varying amounts of nitrogen to test the hypothesis that carrier gas pressure affects the onset of condensation. Such an effect might occur if nonisothermal nucleation were important under conditions of excess carrier gas in the atmospheric pressure range, as has been suggested by Ford and Clement [J. Phys. A 22, 4007 (1989)]. Although a small increase was observed in the condensation onset temperature as the stagnation pressure was reduced from 3 to 0.5 atm, these changes cannot be attributed to any nonisothermal effects. The pulsed nozzle experiments also exhibited two interesting anomalies: (1) the density profiles for the water and ethanol mixtures were shifted in opposite directions from the dry N 2 profile; (2) a long transient period was required before the nozzle showed good pulse-to-pulse repeatability for condensible vapor mixtures. To theoretically simulate the observed onset behavior, calculations of nucleation and droplet growth in the nozzle were performed that took into account two principal effects of varying the carrier gas pressure: (1) the change in nozzle shape due to boundary layer effects and (2) the variation in the heat capacity of the flowing gas. Energy transfer limitations were neglected in calculating the nucleation rates. The trend of the calculated results matched that of the experimental results very well. Thus, heat capacity and boundary layer effects are sufficient to explain the experimental onset behavior without invoking energy transfer limited nucleation. The conclusions about the rate of nucleation are consistent with those obtained recently using an expansion cloud chamber, but are at odds with results from thermal diffusion cloud chamber measurements

Effect of Pressure on the Uniformity of Nozzles Transverse Distribution and Mathematical Model Development

Directory of Open Access Journals (Sweden)

Vladimir Višacki

2017-01-01

Full Text Available Timely and high-quality application of pesticides contributes to environmental protection, economical production and production of healthy food. The efficacy of pesticide application depends not only on the quality of pesticides but also the quality of the application. One of the factor that most influences the quality of applications, from the standpoint of mechanization, are nozzles. They working liquid applied on the surface the plant resulting in the same volume of pesticide is applied to the entire surface of the plants. To achieve this goal, nozzles must be performed uniform application of working liquid per unit area, or tractor sprayer working width. The variable factor in the application of pesticides may be nozzle and operating pressure. With increasing working pressure obtained smaller droplets. The paper presents test of three different nozzles. Each nozzle is characterized by a flat jet with an angle of 110° and a flow rate of 1.6 l∙min−1 at a pressure of 3 bar. Differ from each other are by the way of disintegration of the jet. Exactly this characteristic causes that with pressure change coming to changes in the uniformity of nozzles transverse distribution. So the best distribution has nozzle with a flat jet. The coefficient of variation is between roughly from 4 to 6 % at the pressure application of 2 to 4 bar. Obtained mathematical model that describes changes in the coefficient of variation depending on pressure applications can be a good basis for easy harmonization parameters in the pesticide application.

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

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.

Supersonic flaw detection device for nozzle

International Nuclear Information System (INIS)

Hata, Moriki.

1996-01-01

In a supersonic flaw detection device to be attached to a body surface of a reactor pressure vessel for automatically detecting flaws of a welded portion of a horizontally connected nozzle by using supersonic waves, a running vehicle automatically running along a circumferential direction of the nozzle comprises a supersonic flaw detection means for detecting flaws of the welded portion of the nozzle by using supersonic waves, and an inclination angle sensor for detecting the inclination angle of the running vehicle relative to the central axis of the nozzle. The running distance of the vehicle running along the circumference of the nozzle, namely, the position of the running vehicle from a reference point of the nozzle can be detected accurately by dividing the distance around the nozzle by the inclination angle detected by the inclination angle sensor. Accordingly, disadvantages in the prior art, for example, that the detected values obtained by using an encoder are changed by slipping or idle running of the magnet wheels are eliminated, and accurate flaw detection can be conducted. In addition, an operation of visually adjusting the reference point for the device can be eliminated. An operator's exposure dose can be reduced. (N.H.)

Phased array concept for the ultrasonic inservice inspection of the spherical bottom of BWR-pressure vessels

International Nuclear Information System (INIS)

Brekow, G.; Wuestenberg, H.; Moehrle, W.; Schulz, E.

1989-01-01

The spherical bottom of BWR-pressure vessels contains holes for the nozzles of control rods and instrumentation. Up to now the detectable areas for the ultrasonic inspection are the accessible ligaments between the nozzles with an orientation parallel and transverse to the manipulator rails. Some licensing authorities demand an inspection technique capable of reliably detecting significant crack initiation in all critical areas near the cladding of the spherical inner surface. By order and in cooperation with the HEW we have developed a computer controlled equipment with two ultrasonic probes containing four linear arrays and a digitized A-scan storage for documentation and evaluation of inspection results. The manipulator guided probe movement in the paths between the nozzles of the spherical bottom is controlled by a computer program. This program determines for each array system and for each coupling position the beam angle as a function of the variable skewing angle to realize detection conditions suited to possible crack positions at the longitudinal, transverse and diagonal ligaments between the nozzles for control rods and instrumentation. (orig./HP)

Thermo-hydraulic behavior of saturated steam-water mixture in pressure vessel during injection of cold water

International Nuclear Information System (INIS)

Aya, Izuo; Kobayashi, Michiyuki; Inasaka, Fujio; Nariai, Hideki.

1983-01-01

The thermo-hydraulic behavior of saturated steam water mixture in a pressure vessel during injection of cold water was experimentally investigated with the Facility for Mixing Effect of Emergency Core Cooling Water. The dimensions of the pressure vessel used in the experiments were 284mm ID and 1,971mm height. 11 experiments were conducted without blowdown in order to comprehend the basic process excluding the effect of blowdown at injection of cold water. The initial pressure and water level, the injection flow rate and the size of injection nozzle were chosen as experimental parameters. Temperatures and void fractions at 6 elevations as well as pressure in the pressure vessel were measured, and new data especially on the pressure undershoot just after the initation of water injection and the vertical distribution of temperature and void fraction were gotten. The transients of pressure, average temperature and void fraction were caluculated using single-volume analysis code BLODAC-1V which is based on thermal equilibrium and so-called bubble gradient model. Some input parameters included in the analysis code were evaluated through the comparison of analysis with experimental data. Moreover, the observed pressure undershoot which is evaluated to be induced by a time lag of vapourization in water due to thermal nonequilibrium, was also discussed with the aid of another simple analysis model. (author)

Effects of Nozzle Configuration on Rock Erosion Under a Supercritical Carbon Dioxide Jet at Various Pressures and Temperatures

Directory of Open Access Journals (Sweden)

Man Huang

2017-06-01

Full Text Available The supercritical carbon dioxide (SC-CO2 jet offers many advantages over water jets in the field of oil and gas exploration and development. To take better advantage of the SC-CO2 jet, effects of nozzle configuration on rock erosion characteristics were experimentally investigated with respect to the erosion volume. A convergent nozzle and two Laval nozzles, as well as artificial cores were employed in the experiments. It was found that the Laval nozzle can enhance rock erosion ability, which largely depends on the pressure and temperature conditions. The enhancement increases with rising inlet pressure. Compared with the convergent nozzle, the Laval-1 nozzle maximally enhances the erosion volume by 10%, 21.2% and 30.3% at inlet pressures of 30, 40 and 50 MPa, respectively; while the Laval-2 nozzle maximally increases the erosion volume by 32.5%, 49.2% and 60%. Moreover, the enhancement decreases with increasing ambient pressure under constant inlet pressure or constant pressure drop. The growth of fluid temperature above the critical value can increase the enhancement. In addition, the jet from the Laval-2 nozzle with a smooth inner profile always has a greater erosion ability than that from the Laval-1 nozzle.

Manufacturing and material properties of forgings for reactor pressure vessel of high temperature engineering test reactor

International Nuclear Information System (INIS)

Sato, I.; Suzuki, K.

1994-01-01

For the reactor pressure vessel (RPV) of high temperature engineering test reactor (HTTR) which has been developed by Japan Atomic Energy Research Institute (JAERI), 2 1/4Cr-1Mo steel is used first in the world. Material confirmation test has been carried out to demonstrate good applicability of forged low Si 2 1/4Cr-1Mo steel to the RPV of HTTR. Recently, JSW has succeeded in the manufacturing of large size ring forgings and large size forged cover dome integrated with nozzles for stand pipe for the RPV. This paper describes the results of the material confirmation test as well as the manufacturing and material properties of the large forged cover dome integrated with nozzles for stand pipe. (orig.)

Cavitation inception in nozzle-plate and wire mesh pressure droppers in water and sodium

International Nuclear Information System (INIS)

Collinson, A.E.

1976-01-01

Cavitation tests on multi-hole nozzle plates and wire meshes approximately 100mm diameter in water at 20 deg C and sodium at 300 deg C are described. These pressure dropping elements were mounted in recirculating loops where cavitation was induced by gradually lowering the back-ground pressure at constant flow. Cavitation was detected acoustically using wall mounted piezoelectric microphones, the signal being displayed on a ratemeter recording individual cavitation events. For nozzle plates, cavitation started intermittently as the pressure was lowered, the noise level suddenly increasing at a critical cavitation number sigma. For meshes the intermittent region was absent. Values of sigma for nozzles and meshes were similar in water and sodium for the conditions prevailing during the tests. It was apparent that cavitation took place on the axes of vortices both in the free stream and close to nozzle curved surfaces

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.)

Development of automatic Ultrasonic testing equipment for reactor pressure vessel

International Nuclear Information System (INIS)

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

1996-06-01

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

Development of automatic Ultrasonic testing equipment for reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

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

1996-06-01

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

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

Determine spray droplets on water sensitive paper (WSP) for low pressure deflector nozzle using image J

Science.gov (United States)

Sies, M. F.; Madzlan, N. F.; Asmuin, N.; Sadikin, A.; Zakaria, H.

2017-09-01

In this study, determine of spray droplets size (SMD) using water sensitive paper (WSP) at low fluid pressure with deflector nozzle or tangential flow nozzle model Delavan AL75 and New Design Nozzle with two different type of swirl (ND2.5 A1.0 & ND2.5 B1.0). These three deflected flat sprays have used at different liquid mixing ratio. These liquid mixture ratios are pure water, 10% of lime juice + 90% of water (L10W90) and 30% of lime juice + 70% of water (L30W70). WSP is used to collect the spray droplets from nozzles. The operational liquid pressure of each nozzle is 3 bar, while air operational pressures are 3 bar and 6 bar. Then, the WSP were scanned using scanner then it was analyzed using ImageJ software. ImageJ can be used for determining the diameter of droplets size on the WSP. As the results from an experiment, the AL75 nozzle recorded the lowest Sauter mean diameter which is 193.69μm at 6 bar of pressurized air while ND2.5 A1.0 recorded the highest Sauter mean diameter which is 353.61µm at 3 bar of pressurized air. Summary from the experiment shows that the higher of droplet size is because of the lower air pressure (3 Bar). Then, increasing of liquid viscosity also increase the SMD. The orifice diameter for New Design nozzle (ND-2.5) is smaller than AL75, which are 2.5mm and 2.8mm respectively. The different nozzle design also gives effect the SMD. WSP is an alternative method to determine SMD for spray droplets with the low cost if compared to Phase Doppler Anemometry (PDA).

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.

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.)

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.

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

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

Feasibility of local stress relieving close to main shell of a large vessel

International Nuclear Information System (INIS)

Hancinsky, O.A.

1978-01-01

This work determines the feasibility of local stress relieving for a circumferential pipe-to-nozzle field weld positioned close to the main shell of a large pressure vessel. This is applicable to nuclear as well as conventional vessels. ANSYS computer program is utilized to perform thermal and thermal stress analysis and ASME Pressure Vessels Code is adhered to. Conclusions and recommendations are made with a view on their applicability in practice

Characterization of the full cone pressure swirl spray nozzles for the nuclear reactor containment spray system

Energy Technology Data Exchange (ETDEWEB)

Jain, Manish [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); John, Benny [Nuclear Power Corporation of India Limited, Mumbai (India); Iyer, K.N. [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Prabhu, S.V., E-mail: svprabhu@iitb.ac.in [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India)

2014-07-01

Highlights: • Full cone spray pressure swirl nozzle with X-Vane is studied. • Laser illuminated imaging technique is used. • Correlations for coefficient of discharge, spray cone angle and SMD are suggested. • Droplet size and mass fraction distribution is measured. • Inviscid theory predicts the coefficient of discharge. - Abstract: The objective of the present study is to characterize a full cone pressure swirl nozzle for the Containment Spray System (CSS) of Indian Pressurized heavy Water reactors (IPHWR). The influence of Reynolds number and geometric parameters on the coefficient of discharge, spray cone angle, mass flux density distribution, droplet size distribution, Sauter mean diameter (SMD is studied for full cone pressure swirl full cone nozzles. The nozzles of orifice diameter range from 1.3 to 7.2 mm are studied. Experiments are conducted with water at room temperature as the working medium. The nozzles are operated with the pressure ranging from 1 to 8 bar. The measurements of the drop size distributions are performed with laser illuminated imaging technique. The spray cone-angle of the full cone nozzles is measured by the evaluation of images recorded with a camera using IMAGE J software. Correlations for coefficient of discharge, spray cone angle and Sauter mean diameter are suggested on the basis of the experimental results. Rosin–Rammler model and Nukiyama–Tanasawa distributions predict the mass fraction distribution reasonably well. However, the droplet size distribution is predicted by Nukiyama-Tanasawa model only.

The effect of nozzle diameter, injection pressure and ambient temperature on spray characteristics in diesel engine

Science.gov (United States)

Rhaodah Andsaler, Adiba; Khalid, Amir; Sharifhatul Adila Abdullah, Nor; Sapit, Azwan; Jaat, Norrizam

2017-04-01

Mixture formation of the ignition process is a key element in the diesel combustion as it influences the combustion process and exhaust emission. Aim of this study is to elucidate the effects of nozzle diameter, injection pressure and ambient temperature to the formation of spray. This study investigated diesel formation spray using Computational Fluid Dynamics. Multiphase volume of fluid (VOF) behaviour in the chamber are determined by means of transient simulation, Eulerian of two phases is used for implementation of mixing fuel and air. The detail behaviour of spray droplet diameter, spray penetration and spray breakup length was visualised using the ANSYS 16.1. This simulation was done in different nozzle diameter 0.12 mm and 0.2 mm performed at the ambient temperature 500 K and 700 K with different injection pressure 40 MPa, 70 MPa and 140 MPa. Results show that high pressure influence droplet diameter become smaller and the penetration length longer with the high injection pressure apply. Smaller nozzle diameter gives a shorter length of the breakup. It is necessary for nozzle diameter and ambient temperature condition to improve the formation of spray. High injection pressure is most effective in improvement of formation spray under higher ambient temperature and smaller nozzle diameter.

Elastic stresses at reinforced nozzles in spherical shells with pressure and moment loading

International Nuclear Information System (INIS)

Rodabaugh, E.C.; Gwaltney, R.D.

1976-01-01

Calculated elastic stresses at reinforced nozzles in spherical shells with pressure and moment loading are presented. The models used in the calculations represent a wide variety of reinforced shapes; all meeting Code requirements. The results show Code stress indices for pressure loading for nozzles with local reinforcement are acceptable with some modification in coverage. Simple equations for stress indices for moment loading are developed. Potential application of the moment-loading stress indices is discussed. Several recommendations for Code changes are included

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

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

Pressure drop performance evaluation for test assemblies with the newly developed top and bottom nozzles

International Nuclear Information System (INIS)

Lee, S. K.; Park, N. K.; Su, J. M.; Kim, H. K.; Lee, J. N.; Kim, K. T.

2003-01-01

To perform the hydraulic test for the newly developed top and bottom nozzles, two kinds of test assemblies were manufactured i. e. one is the test assembly which has the newly developed top and bottom nozzles and the other is Guardian test assembly which is commercially in mass production now. The test results show that the test assembly with one top nozzle and two bottom nozzles has a greater pressure loss coefficient than Guardian test assembly by 60.9% and 90.4% at the bottom nozzle location. This cause is due to the debris filtering plate for bottom nozzle to improve a filtering efficiency aginst foreign material. In the region of mid grid and top nozzle, there is no difference in pressure loss coefficient between the test assemblies since the componet features in these regions are very similar or same each other. The loss coefficients are 14.2% and 21.9% for model A and B respectively in the scale of test assembly, and the value would be within the 10% in the scale of real fuel assembly. As a result of hydraulic performance evaluation, model A is superior to model B

Evaluation of the integrity of SEP reactor vessels

International Nuclear Information System (INIS)

Hoge, K.G.

1979-12-01

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

Transient heating effects in high pressure Diesel injector nozzles

International Nuclear Information System (INIS)

Strotos, George; Koukouvinis, Phoevos; Theodorakakos, Andreas; Gavaises, Manolis; Bergeles, George

2015-01-01

Highlights: • Simulation of friction-induced heating in high pressure Diesel fuel injectors. • Injection pressures up to 3000 bar. • Simulations with variable fuel properties significantly affect predictions. • Needle motion affects flow and temperature fields. • Possible heterogeneous boiling as injection pressures increase above 2000 bar. - Abstract: The tendency of today’s fuel injection systems to reach injection pressures up to 3000 bar in order to meet forthcoming emission regulations may significantly increase liquid temperatures due to friction heating; this paper identifies numerically the importance of fuel pressurization, phase-change due to cavitation, wall heat transfer and needle valve motion on the fluid heating induced in high pressure Diesel fuel injectors. These parameters affect the nozzle discharge coefficient (C d ), fuel exit temperature, cavitation volume fraction and temperature distribution within the nozzle. Variable fuel properties, being a function of the local pressure and temperature are found necessary in order to simulate accurately the effects of depressurization and heating induced by friction forces. Comparison of CFD predictions against a 0-D thermodynamic model, indicates that although the mean exit temperature increase relative to the initial fuel temperature is proportional to (1 − C d 2 ) at fixed needle positions, it can significantly deviate from this value when the motion of the needle valve, controlling the opening and closing of the injection process, is taken into consideration. Increasing the inlet pressure from 2000 bar, which is the pressure utilized in today’s fuel systems to 3000 bar, results to significantly increased fluid temperatures above the boiling point of the Diesel fuel components and therefore regions of potential heterogeneous fuel boiling are identified

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

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.)

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.)

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.

Three dimensional stress analysis of nozzle-to-shell intersections by the finite element method and a auto-mesh generation program

International Nuclear Information System (INIS)

Fujihara, Hirohiko; Ueda, Masahiro

1975-01-01

In the design of chemical reactors or nuclear pressure vessels it is often important to evaluate the stress distribution in nozzle-to-shell intersections. The finite element method is a powerful tool for stress analysis, but it has a defects to require troublesome work in preparing input data. Specially, the mesh data of oblique nozzles and tangential nozzles, in which stress concentration is very high, are very difficult to be prepared. The authors made a mesh generation program which can be used to any nozzle-to-shell intersections, and combining this program with a three dimensional stress analysis program by the finite element method they made the stress analysis of nozzle-to-shell intersections under internal pressure. Consequently, stresses, strains and deformations of nozzles nonsymmetrical to spherical shells and nozzles tangential to cylindrical shells were made clear and it was shown that the curvature of the inner surface of the nozzle corner was a controlling factor in reducing stress concentration. (auth.)

Effects of ambient pressure on dynamics of near-nozzle diesel sprays studied by ultrafast x-radiography

International Nuclear Information System (INIS)

Cheong, S. K.; Liu, J.; Shu, D.; Wang, J.; Powell, C. F.; Experimental Facilities Division

2004-01-01

A time-resolved x-radiographic technique has been employed for measuring the fuel distribution close to a single-hole nozzle fitted in a high-pressure diesel injector. Using a monochromatic synchrotron x-ray beam, it is possible to perform quantitative x-ray absorption measurements and obtain two-dimensional projections of the mass of the fuel spray. We have completed a series of spray measurements in the optically dense, near-nozzle region (ml 15 mm from the nozzle orifice) under ambient pressures of 1, 2, and 5.2 bar Nd2 and 1 bar SFd6 at room temperature with injection pressures of 500 and 1000 bar. The focus of the measurements is on the dynamical behaviors of the fuel jets with an emphasis on their penetration in the near-nozzle region. Careful analysis of the time-resolved, x-radiographic data revealed that the spray penetration in this near-nozzle region was not significantly affected by the limited change of the ambient pressure. In addition, well-defined features of the spray, such as the leading and trailing edges, and fluctuations of fuel mass density in the spray body, allowed us to calculate the leading, trailing, and internal speeds of the sprays

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)

Computational study of a High Pressure Turbine Nozzle/Blade Interaction

Science.gov (United States)

Kopriva, James; Laskowski, Gregory; Sheikhi, Reza

2015-11-01

A downstream high pressure turbine blade has been designed for this study to be coupled with the upstream uncooled nozzle of Arts and Rouvroit [1992]. The computational domain is first held to a pitch-line section that includes no centrifugal forces (linear sliding-mesh). The stage geometry is intended to study the fundamental nozzle/blade interaction in a computationally cost efficient manner. Blade/Nozzle count of 2:1 is designed to maintain computational periodic boundary conditions for the coupled problem. Next the geometry is extended to a fully 3D domain with endwalls to understand the impact of secondary flow structures. A set of systematic computational studies are presented to understand the impact of turbulence on the nozzle and down-stream blade boundary layer development, resulting heat transfer, and downstream wake mixing in the absence of cooling. Doing so will provide a much better understanding of stage mixing losses and wall heat transfer which, in turn, can allow for improved engine performance. Computational studies are performed using WALE (Wale Adapted Local Eddy), IDDES (Improved Delayed Detached Eddy Simulation), SST (Shear Stress Transport) models in Fluent.

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)

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.)

Break location influence in pressure vessel SBLOCA scenarios

Energy Technology Data Exchange (ETDEWEB)

Querol, Andrea; Gallardo, Sergio; Verdú, Gumersindo, E-mail: anquevi@upv.es, E-mail: sergalbe@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Instituto Universitario de Seguridad Industrial, Radiofísica y Medioambiental (ISIRYM), Universitat Politècnica de València (Spain)

2017-07-01

The inspections performed in Davis Besse and in the South Texas Project Unit-I reactors pointed out safety issues regarding the structural integrity of the Pressure Vessel (PV). In these inspections, two anomalies were found: a wall thinning and degradation in the PV upper head of the Davis Besse reactor and a small amount of residue around of two instrument-tube penetration nozzles located in the PV lower plenum of the South Texas Project Unit-I reactor. The evolution of these defects could have resulted in Small Break Loss-Of-Coolant Accidents (SBLOCA) if they had not been detected in time. In this frame, the OECD/NEA considered the necessity to simulate these accidental sequences in the OECD/NEA ROSA Project using the Large Scale Test Facility (LSTF). This work is focused in simulating different hypothetical accidental scenarios in the PV using the thermalhydraulic code TRACE5. These simulations allow studying the break localization influence in the transient and the effectiveness of the accident management (AM) actions considered mitigating the consequences of these hypothetical accidental scenarios. (author)

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...

The design of lifting attachments for the erection of large diameter and heavy wall pressure vessels

International Nuclear Information System (INIS)

Antalffy, Leslie P.; Miller, George A.; Kirkpatrick, Kenneth D.; Rajguru, Anil; Zhu, Yong

2016-01-01

Lifting attachments for the erection of large diameter and heavy wall pressure vessels require special consideration to ensure that their attachment to their vessel shells or heads do not overstress the vessel during the erection process when lifting these from grade onto their respective foundations. Today, in refinery and petrochemical services, large diameter vessels with diameters ranging up to 15 m and reactors with lifting weights in the range of 700–1400 tons are not uncommon. In today's fabrication market, these vessels may be purchased and fabricated in shops dispersed globally and will require unique equipment for their safe handling, transportation and subsequent erection. The challenge is to design the lifting attachments in such a manner that the attachments provide a safe, cost effective and effective solution based upon the limitations of the job site lift equipment available for erection. Such equipment for the transportation and subsequent lifting of large diameter and heavy wall pressure equipment is usually scarce and quite expensive. Planning ahead, well in advance of the lift date is almost a mandatory requirement. Usually, the specific parameters of the vessel to be lifted and the lifting equipment available at the site will dictate the type of lifting attachments to be designed for the vessel. Once the type of vessel attachment has been chosen, careful consideration must be given to the design of attachments to the pressure vessel in consideration to ensure that the vessel and lifting components are not overstressed during the lifting process. The paper also discusses different types of lifting attachments that may be attached to each end of the vessel either by bolting or welding and discusses the pros and cons of each. The paper also provides an example of a finite element analysis (FEA) of a top nozzle, a FEA of a pair of lifting trunnions and a FEA of welded on lifting lugs for buried pipe. The purpose of the paper is to outline the

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.

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

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.

Investigation of impulsively loaded pressure vessels

Energy Technology Data Exchange (ETDEWEB)

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

1963-10-15

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

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)

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.)

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

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

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

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.)

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

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

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

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.)

Base Flow and Heat Transfer Characteristics of a Four-Nozzle Clustered Rocket Engine: Effect of Nozzle Pressure Ratio

Science.gov (United States)

Nallasamy, R.; Kandula, M.; Duncil, L.; Schallhorn, P.

2010-01-01

The base pressure and heating characteristics of a four-nozzle clustered rocket configuration is studied numerically with the aid of OVERFLOW Navier-Stokes code. A pressure ratio (chamber pressure to freestream static pressure) range of 990 to 5,920 and a freestream Mach number range of 2.5 to 3.5 are studied. The qualitative trends of decreasing base pressure with increasing pressure ratio and increasing base heat flux with increasing pressure ratio are correctly predicted. However, the predictions for base pressure and base heat flux show deviations from the wind tunnel data. The differences in absolute values between the computation and the data are attributed to factors such as perfect gas (thermally and calorically perfect) assumption, turbulence model inaccuracies in the simulation, and lack of grid adaptation.

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.

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

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...

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

International Nuclear Information System (INIS)

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

1979-01-01

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

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

Influence of transit water flow rate on its dispensation and on inflow through nozzles in pressure pipeline under action of external pressure

Science.gov (United States)

Cherniuk, V. V.; Riabenko, O. A.; Ivaniv, V. V.

2017-12-01

The influence of transit flow rate of water upon operative of the equipped with nozzles pressure pipeline is experimentally investigated. External pressure, which varies in the range of 1465-2295 mm, acted upon the pipeline. The angle β between vectors of velocities of the stream in the pipeline and jets which branch off through nozzles were given the value: 0° ; 45° ; 90° ; 135° ; 180°. The diameter of the pipeline was of D=20.18 mm, the diameter of nozzles d=6.01 mm. The distances between the nozzles were 180 mm, and the number of them 11. The value of the transit flow rate at input into the pipeline varied from 4.05 to 130.20 cm3 / s. The increase in flow rate of the transit flux Qtr caused increase in non-uniformity of distribution of operating heads and increase in flow rate of water along the pipeline over the segment of its dispensation. On the segment of collecting of water, inverse tendency was observed. The number of nozzles through which water became to be dispensed increased with the increase in Qtr.

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.

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)

Axisymmetric thrust-vectoring nozzle performance prediction

International Nuclear Information System (INIS)

Wilson, E. A.; Adler, D.; Bar-Yoseph, P.Z

1998-01-01

Throat-hinged geometrically variable converging-diverging thrust-vectoring nozzles directly affect the jet flow geometry and rotation angle at the nozzle exit as a function of the nozzle geometry, the nozzle pressure ratio and flight velocity. The consideration of nozzle divergence in the effective-geometric nozzle relation is theoretically considered here for the first time. In this study, an explicit calculation procedure is presented as a function of nozzle geometry at constant nozzle pressure ratio, zero velocity and altitude, and compared with experimental results in a civil thrust-vectoring scenario. This procedure may be used in dynamic thrust-vectoring nozzle design performance predictions or analysis for civil and military nozzles as well as in the definition of initial jet flow conditions in future numerical VSTOL/TV jet performance studies

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

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)

Measurement of the residual stresses in a PWR Control Rod Drive Mechanism nozzle

OpenAIRE

Coules, Harry; Smith, David

2018-01-01

Residual stress in the welds that attach Control Rod Drive Mechanism nozzles into the upper head of a PWR reactor vessel can influence the vessel's structural integrity and initiate Primary Water Stress Corrosion Cracking. PWSCC at Alloy 600 CRDM nozzles has caused primary coolant leakage in operating PWRs. We have used Deep Hole Drilling to characterise residual stresses in a PWR vessel head. Measurements of the internal cladding and nozzle attachment weld showed that although modest tensile...

Application of a general purpose finite element program system in pressure vessel technology

International Nuclear Information System (INIS)

Aamodt, B.; Sandsmark, N.; Medonos, S.

1977-01-01

Main advantages of using general purpose finite element program systems in structural analysis are summarized. Several illustrative applications of the program system SESAM-69 to pressure vessel problems are described. The first example is a dynamic analysis of the motor housing of the internal main circulation pump of a BWR nuclear reactor. The next example is a transient heat conduction and stress analysis of deflector of feeding nozzle of PWR nuclear reactor. Then, numerical calculations of stress intensity factors and fatigue crack growth of semi-elliptical surface cracks are discussed. And finally, an elasto-plastic analysis of a thick plate with edge-cracks is considered. It is concluded that due to the fact that general purpose finite element program systems are general and user-orientated, they will gain increasingly higher popularity in the years ahead

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

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

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

Effect of airstream velocity on mean drop diameters of water sprays produced by pressure and air atomizing nozzles

Science.gov (United States)

Ingebo, R. D.

1977-01-01

A scanning radiometer was used to determine the effect of airstream velocity on the mean drop diameter of water sprays produced by pressure atomizing and air atomizing fuel nozzles used in previous combustion studies. Increasing airstream velocity from 23 to 53.4 meters per second reduced the Sauter mean diameter by approximately 50 percent with both types of fuel nozzles. The use of a sonic cup attached to the tip of an air assist nozzle reduced the Sauter mean diameter by approximately 40 percent. Test conditions included airstream velocities of 23 to 53.4 meters per second at 293 K and atmospheric pressure.

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.

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

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

Fractal analysis of agricultural nozzles spray

Directory of Open Access Journals (Sweden)

Francisco Agüera

2012-02-01

Full Text Available Fractal scaling of the exponential type is used to establish the cumulative volume (V distribution applied through agricultural spray nozzles in size x droplets, smaller than the characteristic size X. From exponent d, we deduced the fractal dimension (Df which measures the degree of irregularity of the medium. This property is known as 'self-similarity'. Assuming that the droplet set from a spray nozzle is self-similar, the objectives of this study were to develop a methodology for calculating a Df factor associated with a given nozzle and to determine regression coefficients in order to predict droplet spectra factors from a nozzle, taking into account its own Df and pressure operating. Based on the iterated function system, we developed an algorithm to relate nozzle types to a particular value of Df. Four nozzles and five operating pressure droplet size characteristics were measured using a Phase Doppler Particle Analyser (PDPA. The data input consisted of droplet size spectra factors derived from these measurements. Estimated Df values showed dependence on nozzle type and independence of operating pressure. We developed an exponential model based on the Df to enable us to predict droplet size spectra factors. Significant coefficients of determination were found for the fitted model. This model could prove useful as a means of comparing the behavior of nozzles which only differ in not measurable geometric parameters and it can predict droplet spectra factors of a nozzle operating under different pressures from data measured only in extreme work pressures.

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

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.

Pressure Distribution on Inner Wall of Parabolic Nozzle in Laser Propulsion with Single Pulse

Science.gov (United States)

Cui, Cunyan; Hong, Yanji; Wen, Ming; Song, Junling; Fang, Juan

2011-11-01

A system based of dynamic pressure sensors was established to study the time resolved pressure distribution on the inner wall of a parabolic nozzle in laser propulsion. Dynamic calibration and static calibration of the test system were made and the results showed that frequency response was up to 412 kHz and linear error was less than 10%. Experimental model was a parabolic nozzle and three test points were preset along one generating line. This study showed that experimental results agreed well with those obtained by numerical calculation way in pressure evolution tendency. The peak value of the calculation was higher than that of the experiment at each tested orifice because of the limitation of the numerical models. The results of this study were very useful for analyzing the energy deposition in laser propulsion and modifying numerical models.

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.

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

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.

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

International Nuclear Information System (INIS)

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

1976-08-01

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

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

Flow-throttling orifice nozzle

International Nuclear Information System (INIS)

Sletten, H.L.

1975-01-01

A series-parallel-flow type throttling apparatus to restrict coolant flow to certain fuel assemblies of a nuclear reactor is comprised of an axial extension nozzle of the fuel assembly. The nozzle has a series of concentric tubes with parallel-flow orifice holes in each tube. Flow passes from a high pressure plenum chamber outside the nozzle through the holes in each tube in series to the inside of the innermost tube where the coolant, having dissipated most of its pressure, flows axially to the fuel element. (U.S.)

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)

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)

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

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)

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.

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...

Pitot-Pressure Measurements in Flow Fields Behind a Rectangular Nozzle with Exhaust Jet for Free-Stream Mach Numbers of 0.00, 0.60, and 1.20

Science.gov (United States)

Putnam, L. E.; Mercer, C. E.

1986-01-01

An investigation has been conducted in the Langley 16-Foot Transonic Tunnel to measure the flow field in and around the jet exhaust from a nonaxisymmetric nozzle configuration. The nozzle had a rectangular exit with a width-to-height ratio of 2.38. Pitot-pressure measurements were made at five longitudinal locations downstream of the nozzle exit. The maximum distance downstream of the exit was about 5 nozzle heights. These measurements were made at free-stream Mach numbers of 0.00, 0.60, and 1.20 with the nozzle operating at a ratio of nozzle total pressure to free-stream static pressure of 4.0. The jet exhaust was simulated with high-pressure air that had an exit total temperature essentially equal to the free-stream total temperature.

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.)

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.

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.)

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...

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)

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.)

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 ...

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

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

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

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

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)

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...

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.

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)

Pitot pressure measurements in flow fields behind circular-arc nozzles with exhaust jets at subsonic free-stream Mach numbers. [langley 16 foot transonic tunnel

Science.gov (United States)

Mason, M. L.; Putnam, L. E.

1979-01-01

The flow field behind a circular arc nozzle with exhaust jet was studied at subsonic free stream Mach numbers. A conical probe was used to measure the pitot pressure in the jet and free stream regions. Pressure data were recorded for two nozzle configurations at nozzle pressure ratios of 2.0, 2.9, and 5.0. At each set of test conditions, the probe was traversed from the jet center line into the free stream region at seven data acquisition stations. The survey began at the nozzle exit and extended downstream at intervals. The pitot pressure data may be applied to the evaluation of computational flow field models, as illustrated by a comparison of the flow field data with results of inviscid jet plume theory.

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)

Estimation of residual stress distribution for pressurizer nozzle of Kori nuclear power plant considering safe end

Energy Technology Data Exchange (ETDEWEB)

Song, Tae Kwang; Bae, Hong Yeol; Chun, Yun Bae; Oh, Chang Young; Kim, Yun Jae [Korea University, Seoul (Korea, Republic of); Lee, Kyoung Soo; Park, Chi Yong [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

2008-08-15

In nuclear power plants, ferritic low alloy steel nozzle was connected with austenitic stainless steel piping system through alloy 82/182 butt weld. Accurate estimation of residual stress for weldment is important in the sense that alloy 82/182 is susceptible to stress corrosion cracking. There are many results which predict residual stress distribution for alloy 82/182 weld between nozzle and pipe. However, nozzle and piping system usually connected through safe end which has short length. In this paper, residual stress distribution for pressurizer nozzle of Kori nuclear power plant was predicted using FE analysis, which considered safe end. As a result, existing residual stress profile was redistributed and residual stress of inner surface was decreased specially. It means that safe end should be considered to reduce conservatism when estimating the piping system.

Investigation of stress distribution in normal and oblique partial penetration. Welded nozzles by 3-D photoelastic stress freezing method

International Nuclear Information System (INIS)

Miyamoto, H.; Kubo, M.; Katori, T.

1981-01-01

Experimental investigation by 3-D photoelasticity has been carried out to measure the stress distribution of partial penetration welded nozzles attached to the bottom head of a pressure vessel. A 3-D photoelastic stress freezing method was chosen as the most effective means of observation of the stress distribution in the vicinity of the nozzle/wall weld. The experimental model was a 1:20 scale spherical bottom head. Both an axisymmetric nozzle and an asymmetric nozzle were investigated. Epoxy resin, which is a thermosetting plastic, was used as the model material. The oblique effect was examined by comparing the stress distribution of the asymmetric nozzle with that of the axisymmetric nozzle. Furthermore, the experimental results were compared with the analytical results using 3-D finite element method (FEM). The stress distributions obtained from the frozen fringe pattern of the 3-D photoelastic model were in good agreement with those by 3-D FEM. (orig.)

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)

Development of a Remotely-operated Visual Inspection System for Reactor Vessel Bottommounted Instrument Penetrations of KSNP and Lessons Learned

International Nuclear Information System (INIS)

Jeong, Kyungmin; Choi, Youngsu; Lee, Sunguk; Seo, Yongchil; Kang, Jong Gyu; Kim, Seungho; Jung, Seungho

2006-01-01

In April 2003, South Texas Project Unit 1 made a surprising discovery of boron acid leakage from two nozzles from a bare-metal examination of the reactor vessel bottom-mounted instrument penetrations during a routine refueling outage. A small powdery substance about 150mg was found on the outside of two instrument guide penetration nozzles on the bottom of the reactor. The primary coolant water of pressurized water reactors has caused cracking in penetrations with Alloy 600 through a process called primary water stress corrosion cracking. In South Korea, it is required to conduct 100% visual inspection of the outside of instrument guide penetration nozzles on the bottom of PWRs to confirm the integrity of reactor vessel. This paper describes the remotely-operated visual inspection systems for reactor vessel bottom-mounted instrument penetrations dispatched two times to Youngkwang NPPs and discusses the lessons learned

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.

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

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

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

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

Firefighter Nozzle Reaction

DEFF Research Database (Denmark)

Chin, Selena K.; Sunderland, Peter B.; Jomaas, Grunde

2017-01-01

to anchor forces, the hose becomes straight. The nozzle reaction is found to equal the jet momentum flow rate, and it does not change when an elbow connects the hose to the nozzle. A forward force must be exerted by a firefighter or another anchor that matches the forward force that the jet would exert...... on a perpendicular wall. Three reaction expressions are derived, allowing it to be determined in terms of hose diameter, jet diameter, flow rate, and static pressure upstream of the nozzle. The nozzle reaction predictions used by the fire service are 56% to 90% of those obtained here for typical firefighting hand...

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...

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.

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.

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)

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...

External Cylindrical Nozzle with Controlled Vacuum

Directory of Open Access Journals (Sweden)

V. N. Pil'gunov

2015-01-01

Full Text Available There is a developed design of the external cylindrical nozzle with a vacuum camera. The paper studies the nozzle controllability of flow rate via regulated connection of the evacuated chamber to the atmosphere through an air throttle. Working capacity of the nozzle with inlet round or triangular orifice are researched. The gap is provided in the nozzle design between the external wall of the inlet orifice and the end face of the straight case in the nozzle case. The presented mathematical model of the nozzle with the evacuated chamber allows us to estimate the expected vacuum amount in the compressed section of a stream and maximum permissible absolute pressure at the inlet orifice. The paper gives experimental characteristics of the fluid flow process through the nozzle for different values of internal diameter of a straight case and an extent of its end face remoteness from an external wall of the inlet orifice. It estimates how geometry of nozzle constructive elements influences on the volume flow rate. It is established that the nozzle capacity significantly depends on the shape of inlet orifice. Triangular orifice nozzles steadily work in the mode of completely filled flow area of the straight case at much more amounts of the limit pressure of the flow. Vacuum depth in the evacuated chamber also depends on the shape of inlet orifice: the greatest vacuum is reached in a nozzle with the triangular orifice which 1.5 times exceeds the greatest vacuum with the round orifice. Possibility to control nozzle capacity through the regulated connection of the evacuated chamber to the atmosphere was experimentally estimated, thus depth of flow rate regulation of the nozzle with a triangular orifice was 45% in comparison with 10% regulation depth of the nozzle with a round orifice. Depth of regulation calculated by a mathematical model appeared to be much more. The paper presents experimental dependences of the flow coefficients of nozzle input orifice

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

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

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)

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...

Development of automatic ultrasonic testing equipment for reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Jang, Kee Ok; Park, Dae Yung; Park, Moon Hoh; Koo, Kil Mo; Park, Kwang Heui; Kang, Sang Sin; Bang, Heui Song; Noh, Heui Choong; Kong, Woon Sik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-08-01

The selected weld areas of reactor pressure vessel and adjacent piping are examined by remote mechanized ultrasonic testing(MUT) equipment. Since the MUT equipment was purchased from Southwest Research Institute (SwRI) in April 1985, we have performed 15 inservice inspections and 5 preservice inspections. However, the reliability of examination was recently decreased rapidly as the problems which results from the old age of equipment and the frequent movement to plant site to site have occurred frequently. Therefore, the 3-axis control system hardware in occurring many problems among the equipments of mechanized ultrasonic testing (MUT) was designed and developed to cover the examination areas of nozzle-shell weld as specified in ASME Code Section XI and to improve the examination reliability. The new 3-axis control system hardware with the performance of this project was developed to be compatible with the old one and it was used as dual system or spare parts of the old system. Furthermore, the established technologies are expected to be applied to the similar control systems in nuclear power plant. 17 figs, 2 pix, 2 tabs, 10 refs. (Author).

Development of automatic ultrasonic testing equipment for reactor pressure vessel

International Nuclear Information System (INIS)

Jang, Kee Ok; Park, Dae Yung; Park, Moon Hoh; Koo, Kil Mo; Park, Kwang Heui; Kang, Sang Sin; Bang, Heui Song; Noh, Heui Choong; Kong, Woon Sik

1994-08-01

The selected weld areas of reactor pressure vessel and adjacent piping are examined by remote mechanized ultrasonic testing(MUT) equipment. Since the MUT equipment was purchased from Southwest Research Institute (SwRI) in April 1985, we have performed 15 inservice inspections and 5 preservice inspections. However, the reliability of examination was recently decreased rapidly as the problems which results from the old age of equipment and the frequent movement to plant site to site have occurred frequently. Therefore, the 3-axis control system hardware in occurring many problems among the equipments of mechanized ultrasonic testing (MUT) was designed and developed to cover the examination areas of nozzle-shell weld as specified in ASME Code Section XI and to improve the examination reliability. The new 3-axis control system hardware with the performance of this project was developed to be compatible with the old one and it was used as dual system or spare parts of the old system. Furthermore, the established technologies are expected to be applied to the similar control systems in nuclear power plant. 17 figs, 2 pix, 2 tabs, 10 refs. (Author)

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...

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)

Fuselage and nozzle pressure distributions on a 1/12-scale F-15 propulsion model at transonic speeds. [conducted in langley 16 foot transonic tunnel

Science.gov (United States)

Pendergraft, O. C., Jr.

1979-01-01

Static pressure coefficient distributions on the forebody, afterbody, and nozzles of a 1/12 scale F-15 propulsion model were determined. The effects of nozzle power setting and horizontal tail deflection angle on the pressure coefficient distributions were investigated.

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)

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

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.)

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.)

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

Micromechanisms of ductile stable crack growth in nuclear pressure vessel steels

Energy Technology Data Exchange (ETDEWEB)

Belcher, W.P.A.; Druce, S.G.

1981-10-01

The objective of this work was to investigate the relationship between the micromechanisms of ductile crack growth, the microstructural constituent phases present in nuclear pressure vessel steel, and the observed fracture behavior as determined by impact and fracture mechanics tests. Results from a microstructural and mechanical property comparison of an A508 Class 3 pressurized water reactor nozzle forging cutout and a 150-mm-thick A533B Class 1 plate are reported. The variation of upper-shelf toughness between the two steels and its orientation sensitivity are discussed on the basis of inclusion and precipitate distributions. Inclusion clusters in A533B, deformed to elongated disks in the rolling plane, have a profound effect on short transverse fracture properties. Data derived using the multi-specimen J-integral method to characterize the initiation of ductile crack extension and resistance to stable crack growth are compared with equivalent Charpy results. Results of the J /SUB R/ -curve analyses indicate (1) that the A533B short transverse crack growth resistance is approximately half that observed from transverse and longitudinal specimen orientations, and (2) that the A508 initiation toughness and resistance to stable crack growth are insensitive to position through the forging wall, and are higher than exhibited by A533B at any orientation in the midthickness position.

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.

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

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)

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

Study of impact of the AP1000{sup Registered-Sign} reactor vessel upper internals design on fuel performance

Energy Technology Data Exchange (ETDEWEB)

Xu Yiban; Conner, Michael; Yuan Kun; Dzodzo, Milorad B.; Karoutas, Zeses; Beltz, Steven A.; Ray, Sumit; Bissett, Teresa A. [Westinghouse Electric Company, Cranberry Township, PA 16066 (United States); Chieng, Ching-Chang, E-mail: cchieng@ess.nthu.edu.tw [National Tsing Hua University, Hsinchu 30043, Taiwan (China); Kao, Min-Tsung; Wu, Chung-Yun [National Tsing Hua University, Hsinchu 30043, Taiwan (China)

2012-11-15

One aspect of the AP1000{sup Registered-Sign} reactor design is the reduction in the number of major components and simplification in manufacturing. One design change relative to current Westinghouse reactors of similar size is the reduction in the number of reactor vessel outlet nozzles/hot legs leaving the upper plenum from three to two. With regard to fuel performance, this design difference creates a different flow field in the AP1000 reactor vessel upper plenum (the region above the core). The flow exiting core and entering the upper plenum must turn 90 Degree-Sign , flow laterally through the upper plenum around support structures, and exit through one of the two outlet nozzles. While the flow in the top of the core is mostly axial, there is some lateral flow component as the core flow reacts to the flow field and pressure distribution in the upper plenum. The pressure distribution in the upper plenum varies laterally depending upon various factors including the proximity to the outlet nozzles. To determine how the lateral flow in the top of the AP1000 core compares to current Westinghouse reactors, a computational fluid dynamics (CFD) model of the flow in the upper portion of the AP1000 reactor vessel including the top region of the core, the upper plenum, the reactor vessel outlet nozzles, and a portion of the hot legs was created. Due to geometric symmetry, the computational domain was reduced to a quarter (from the top view) that includes Vulgar-Fraction-One-Quarter of the top of the core, Vulgar-Fraction-One-Quarter of the upper plenum, and Vulgar-Fraction-One-Half of an outlet nozzle. Results from this model include predicted velocity fields and pressure distributions throughout the model domain. The flow patterns inside and around guide tubes clearly demonstrate the influence of lateral flow due to the presence of the outlet nozzles. From these results, comparisons of AP1000 flow versus current Westinghouse plants were performed. Field performance

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...

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

Residual stress reduction in the penetration nozzle weld joint by overlay welding

International Nuclear Information System (INIS)

Jiang, Wenchun; Luo, Yun; Wang, B.Y.; Tu, S.T.; Gong, J.M.

2014-01-01

Highlights: • Residual stress reduction in penetration weld nozzle by overlay welding was studied. • The overlay weld can decrease the residual stress in the weld root. • Long overlay welding is proposed in the actual welding. • Overlay weld to decrease residual stress is more suitable for thin nozzle. - Abstract: Stress corrosion cracking (SCC) in the penetration nozzle weld joint endangers the structural reliability of pressure vessels in nuclear and chemical industries. How to decrease the residual stress is very critical to ensure the structure integrity. In this paper, a new method, which uses overlay welding on the inner surface of nozzle, is proposed to decrease the residual stresses in the penetration joint. Finite element simulation is used to study the change of weld residual stresses before and after overlay welding. It reveals that this method can mainly decrease the residual stress in the weld root. Before overlay welding, large tensile residual stresses are generated in the weld root. After overlay weld, the tensile hoop stress in weld root has been decreased about 45%, and the radial stress has been decreased to compressive stress, which is helpful to decrease the susceptibility to SCC. With the increase of overlay welding length, the residual stress in weld root has been greatly decreased, and thus the long overlay welding is proposed in the actual welding. It also finds that this method is more suitable for thin nozzle rather than thick nozzle

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

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

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

RNL automated ultrasonic inspection of the PISC II PWR inlet nozzle (Plate 3)

International Nuclear Information System (INIS)

Rogerson, A.; Poulter, L.N.J.; Clough, P.; Cooper, A.G.

1987-01-01

In June 1984, Risley Nuclear Laboratories (RNL) performed an automated ultrasonic inspection of the Pressurized Water Reactor (PWR) inlet nozzle (plate 3) from the international Programme of Inspection of Steel Components (PISC II) round-robin inspection programme. High-sensitivity pulse-echo detection and predominantly time-of-flight diffraction sizing techniques were employed from the clad inner surface of the nozzle using digital data collection, analysis, and display facilities developed at RNL. RNL detected 30 out of 31 intended weld flaws, achieved one hundred per cent correct acceptance of all acceptable flaws and had a correct rejection frequency on all rejectable flaws of 0.86. The results confirm that well-conceived automated inspection procedures, similar to those used by RNL in this nozzle inspection, could form the basis of a PSI/ISI procedure for reactor pressure vessel nozzle regions. Analysis of the RNL results with regard to the influence of flaw characteristics on inspection performance lends strong support to the general conclusions drawn by the PISC Data Analysis Group. In particular, the most difficult flaws to accurately size were circular smooth and rough flaws. Examination of the RNL results on individual flaws reveals valuable information on the strengths and weaknesses of the adopted procedures and points towards procedural changes that would improve inspection performance. This report describes the procedures adopted by RNL, in the inspection, and reviews the results in the light of definitive flaw information. (author)

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...

Firefighter Nozzle Reaction

DEFF Research Database (Denmark)

Chin, Selena K.; Sunderland, Peter B.; Jomaas, Grunde

2017-01-01

Nozzle reaction and hose tension are analyzed using conservation of fluid momentum and assuming steady, inviscid flow and a flexible hose in frictionless contact with the ground. An expression that is independent of the bend angle is derived for the hose tension. If this tension is exceeded owing...... to anchor forces, the hose becomes straight. The nozzle reaction is found to equal the jet momentum flow rate, and it does not change when an elbow connects the hose to the nozzle. A forward force must be exerted by a firefighter or another anchor that matches the forward force that the jet would exert...... on a perpendicular wall. Three reaction expressions are derived, allowing it to be determined in terms of hose diameter, jet diameter, flow rate, and static pressure upstream of the nozzle. The nozzle reaction predictions used by the fire service are 56% to 90% of those obtained here for typical firefighting hand...

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

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)

Numerical Simulation of the Pressure Distribution in the Reactor Vessel Downcomer Region Fluctuated by the Reactor Coolant Pump

Energy Technology Data Exchange (ETDEWEB)

Jeong, Dong Hwa; Jung, Byung Ryul; Jang, Ho Cheol; Yune, Seok Jeong; Kim, Eun Kee [KEPCO EnC, Daejeon (Korea, Republic of)

2015-10-15

In this study the numerical simulation of the pressure distribution in the downcomer region resulting from the pressure pulsation by the Reactor Coolant Pump (RCP) is performed using the Finite Difference Method (FDM). Simulation is carried out for the cylindrical shaped 2-dimensional model equivalent to the outer surface of the Core Support Barrel (CSB) of APR1400 and a 1/2 model is adopted based on the bilateral symmetry by the inlet nozzle. The fluid temperature is 555 .deg. F and the forcing frequencies are 120Hz, 240Hz, 360Hz and 480Hz. Simulation results of the axial pressure distributions are provided as the Root Mean Square (RMS) values at the five locations of 0°, 45°, 90°, 135° and 180° in the circumferential direction from the inlet nozzle location. In the study, the numerical simulation of pressure distributions in the downcomer region induced by the RCP was performed using FDM and the results were reviewed. The interference of the waves returned from both boundaries in the axial direction and the source of the sinusoidal wave is shown on the inlet nozzle interface pressure point. It seems that the maximum pressures result from the superposition of the waves reflected from the seating surface and the waves newly arrived from the inlet nozzle interface pressure location.

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

International Nuclear Information System (INIS)

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

1985-01-01

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

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.

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

International Nuclear Information System (INIS)

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

1978-01-01

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

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

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

Design and Checkout of a High Speed Research Nozzle Evaluation Rig

Science.gov (United States)

Castner, Raymond S.; Wolter, John D.

1997-01-01

The High Flow Jet Exit Rig (HFJER) was designed to provide simulated mixed flow turbojet engine exhaust for one- seventh scale models of advanced High Speed Research test nozzles. The new rig was designed to be used at NASA Lewis Research Center in the Nozzle Acoustic Test Rig and the 8x6 Supersonic Wind Tunnel. Capabilities were also designed to collect nozzle thrust measurement, aerodynamic measurements, and acoustic measurements when installed at the Nozzle Acoustic Test Rig. Simulated engine exhaust can be supplied from a high pressure air source at 33 pounds of air per second at 530 degrees Rankine and nozzle pressure ratios of 4.0. In addition, a combustion unit was designed from a J-58 aircraft engine burner to provide 20 pounds of air per second at 2000 degrees Rankine, also at nozzle pressure ratios of 4.0. These airflow capacities were designed to test High Speed Research nozzles with exhaust areas from eighteen square inches to twenty-two square inches. Nozzle inlet flow measurement is available through pressure and temperature sensors installed in the rig. Research instrumentation on High Speed Research nozzles is available with a maximum of 200 individual pressure and 100 individual temperature measurements. Checkout testing was performed in May 1997 with a 22 square inch ASME long radius flow nozzle. Checkout test results will be summarized and compared to the stated design goals.

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)

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

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)

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

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

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.)

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...

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

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.

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

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

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)

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.)

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)

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.)

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)

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)

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

Residual stress measurements in the dissimilar metal weld in pressurizer safety nozzle of nuclear power plant

International Nuclear Information System (INIS)

Campos, Wagner R.C.; Rabello, Emerson G.; Mansur, Tanius R.; Scaldaferri, Denis H.B.; Paula, Raphael G.; Souto, Joao P.R.S.; Carvalho Junior, Ideir T.

2013-01-01

Weld residual stresses have a large influence on the behavior of cracking that could possibly occur under normal operation of components. In case of an unfavorable environment, both stainless steel and nickel-based weld materials can be susceptible to stress-corrosion cracking (SCC). Stress corrosion cracks were found in dissimilar metal welds of some pressurized water reactor (PWR) nuclear plants. In the nuclear reactor primary circuit the presence of tensile residual stress and corrosive environment leads to so-called Primary Water Stress Corrosion Cracking (PWSCC). The PWSCC is a major safety concern in the nuclear power industry worldwide. PWSCC usually occurs on the inner surface of weld regions which come into contact with pressurized high temperature water coolant. However, it is very difficult to measure the residual stress on the inner surfaces of pipes or nozzles because of inaccessibility. A mock-up of weld parts of a pressurizer safety nozzle was fabricated. The mock-up was composed of three parts: an ASTM A508 C13 nozzle, an ASTM A276 F316L stainless steel safe-end, an AISI 316L stainless steel pipe and different filler metals of nickel alloy 82/182 and AISI 316L. This work presents the results of measurements of residual strain from the outer surface of the mock-up welded in base metals and filler metals by hole-drilling strain-gage method of stress relaxation. (author)

Effect of nozzle geometry for swirl type twin-fluid water mist nozzle on the spray characteristic

Energy Technology Data Exchange (ETDEWEB)

Yoon, Soon Hyun; Kim, Do Yeon; Kim, Dong Keon [Pusan National University, Busan (Korea, Republic of); Kim, Bong Hwan [Jinju National University, Jinju (Korea, Republic of)

2011-07-15

Experimental investigations on the atomization characteristics of twin-fluid water mist nozzle were conducted using particle image velocimetry (PIV) system and particle motion analysis system (PMAS). The twin-fluid water mist nozzles with swirlers designed two types of swirl angles such as 0 .deg. , 90 .deg. and three different size nozzle hole diameters such as 0.5mm, 1mm, 1.5mm were employed. The experiments were carried out by the injection pressure of water and air divided into 1bar, 2bar respectively. The droplet size of the spray was measured using PMAS. The velocity and turbulence intensity were measured using PIV. The velocity, turbulence intensity and SMD distributions of the sprays were measured along the centerline and radial direction. As the experimental results, swirl angle controlled to droplet sizes. It was found that SMD distribution decreases with the increase of swirl angle. The developed twin-fluid water mist nozzle was satisfied to the criteria of NFPA 750, Class 1. It was proven that the developed nozzle under low pressures could be applied to fire protection system.

Effect of nozzle geometry for swirl type twin-fluid water mist nozzle on the spray characteristic

International Nuclear Information System (INIS)

Yoon, Soon Hyun; Kim, Do Yeon; Kim, Dong Keon; Kim, Bong Hwan

2011-01-01

Experimental investigations on the atomization characteristics of twin-fluid water mist nozzle were conducted using particle image velocimetry (PIV) system and particle motion analysis system (PMAS). The twin-fluid water mist nozzles with swirlers designed two types of swirl angles such as 0 .deg. , 90 .deg. and three different size nozzle hole diameters such as 0.5mm, 1mm, 1.5mm were employed. The experiments were carried out by the injection pressure of water and air divided into 1bar, 2bar respectively. The droplet size of the spray was measured using PMAS. The velocity and turbulence intensity were measured using PIV. The velocity, turbulence intensity and SMD distributions of the sprays were measured along the centerline and radial direction. As the experimental results, swirl angle controlled to droplet sizes. It was found that SMD distribution decreases with the increase of swirl angle. The developed twin-fluid water mist nozzle was satisfied to the criteria of NFPA 750, Class 1. It was proven that the developed nozzle under low pressures could be applied to fire protection system

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.

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)

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

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

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

International Nuclear Information System (INIS)

Kobayashi, Ken-ichi; Yamada, Jun-ichi

2010-01-01

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

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

Design of pressure vessels. Part 1

International Nuclear Information System (INIS)

Grandemange, J.M.

2008-01-01

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

CFD Analysis On The Performance Of Wind Turbine With Nozzles

Directory of Open Access Journals (Sweden)

Chunkyraj Kh

2015-08-01

Full Text Available In this paper an effort has been made in dealing with fluid characteristic that enters a converging nozzle and analysis of the nozzle is carried out using Computational Fluid Dynamics package ANSYS WORKBENCH 14.5. The paper is the continuation of earlier work Analytical and Experimental performance evaluation of Wind turbine with Nozzles. First the CFD analysis will be carried out on nozzle in-front of wind turbine where streamline velocity at the exit volume flow rate in the nozzle and pressure distribution across the nozzle will be studied. Experiments were conducted on the Wind turbine with nozzles and the corresponding power output at different air speed and different size of nozzles were calculated. Different shapes and dimensions with special contours and profiles of nozzles were studied. It was observed that the special contour nozzles have superior outlet velocity and low pressure at nozzle exit the design has maximum Kinetic energy. These indicators conclude that the contraction designed with the new profile is a good enhancing of the nozzle performance.

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

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)

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

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)

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

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

Generation of reconstruction algorithms for the testing of complex geometries with the ALOK technique: Testing of the nozzle inner corner from the outside with regard to cladding influence and manipulator movements

International Nuclear Information System (INIS)

Stanger, K.H.; Licht, R.

1989-01-01

At a nozzle of the full-size pressure vessel which contains six artificial defects in the nozzle inner corner ALOK measurement data were recorded and the defect geometry was reconstructed with a special software package. The acoustic irradiation conditions and the operating parameters of the manipulators were selected in such a way that an essentially improved signal-to-noise ratio was achieved by typical ALOK noise suppression in the measurement data. (orig.) [de

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)

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.)

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

Heritability of retinal vessel diameters and blood pressure

DEFF Research Database (Denmark)

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

2006-01-01

PURPOSE: To assess the relative influence of genetic and environmental effects on retinal vessel diameters and blood pressure in healthy adults, as well as the possible genetic connection between these two characteristics. METHODS: In 55 monozygotic and 50 dizygotic same-sex healthy twin pairs......%-80%) for CRAE, 83% (95% CI: 73%-89%) for CRVE, and 61% (95% CI: 44%-73%) for mean arterial blood pressure (MABP). Retinal artery diameter decreased with increasing age and increasing arterial blood pressure. Mean vessel diameters in the population were 165.8 +/- 14.9 microm for CRAE, 246.2 +/- 17.7 microm...... for CRVE, and 0.67 +/- 0.05 microm for AVR. No significant influence on artery or vein diameters was found for gender, smoking, body mass index (BMI), total cholesterol, fasting blood glucose, or 2-hour oral glucose tolerance test values. CONCLUSIONS: In healthy young adults with normal blood pressure...

Effect of Suction Nozzle Pressure Drop on the Performance of an Ejector-Expansion Transcritical CO2 Refrigeration Cycle

Directory of Open Access Journals (Sweden)

Zhenying Zhang

2014-08-01

Full Text Available The basic transcritical CO2 systems exhibit low energy efficiency due to their large throttling loss. Replacing the throttle valve with an ejector is an effective measure for recovering some of the energy lost in the expansion process. In this paper, a thermodynamic model of the ejector-expansion transcritical CO2 refrigeration cycle is developed. The effect of the suction nozzle pressure drop (SNPD on the cycle performance is discussed. The results indicate that the SNPD has little impact on entrainment ratio. There exists an optimum SNPD which gives a maximum recovered pressure and COP under a specified condition. The value of the optimum SNPD mainly depends on the efficiencies of the motive nozzle and the suction nozzle, but it is essentially independent of evaporating temperature and gas cooler outlet temperature. Through optimizing the value of SNPD, the maximum COP of the ejector-expansion cycle can be up to 45.1% higher than that of the basic cycle. The exergy loss of the ejector-expansion cycle is reduced about 43.0% compared with the basic cycle.

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

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

Design and testing of low-divergence elliptical-jet nozzles

Energy Technology Data Exchange (ETDEWEB)

Rouly, Etienne; Warkentin, Andrew; Bauer, Robert [Dalhousie University, Halifax (China)

2015-05-15

A novel approach was developed to design and fabricate nozzles to produce high-pressure low-divergence fluid jets. Rapid-prototype fabrication allowed for myriad experiments investigating effects of different geometric characteristics of nozzle internal geometry on jet divergence angle and fluid distribution. Nozzle apertures were elliptical in shape with aspect ratios between 1.00 and 2.45. The resulting nozzle designs were tested and the lowest elliptical jet divergence angle was 0.4 degrees. Nozzle pressures and flowrates ranged from 0.32 to 4.45 MPa and 13.6 to 37.9 LPM, respectively. CimCool CimTech 310 machining fluid was used in all experiments at a Brix concentration of 6.6 percent.

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.)

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

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)

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.)

East/west steels for reactor pressure vessels

International Nuclear Information System (INIS)

Davies, M.; Kryukov, A.; Nikolaev, Y.; English, C.

1997-01-01

The report consist of three parts dealing with comparison of the irradiation behaviour of 'Eastern' and 'Western' steels, mechanisms of irradiation embrittlement and the role of compositional variations on the irradiation sensitivity of pressure vessels. Nickel, copper and phosphorus are the elements rendering the most essential influence on behaviour of pressure vessel steels under irradiation and subsequent thermal annealing. For WWER-440 reactor pressure vessel (RPV) steels in which nickel content does nor exceed 0.3% the main affecting factors are phosphorous and copper. For WWER-1000 RPV welds in which nickel content generally exceed 1.5% the role of nickel in radiation embrittlement is decisive. In 'Western' type steels main influencing elements are nickel and copper. The secondary role of phosphorus in radiation embrittlement of 'Western' steels is caused by lower relative content compared to 'Eastern' steels. The process of how copper, phosphorus and nickel contents affect the irradiation sensitivity of both types of steel seem to be similar. Some distinctions between the observed radiation effects is apparently caused by differences in the irradiation conditions and ratios of the contents of above mentioned elements in both types of steel. For 'Eastern' RPV steels the dependence of the recovery degree of irradiated steels due to postirradiation thermal annealing id obviously dependent on phosphorus contents and the influence of nickel contents on this process is detectable

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)

Irradiation embrittlement of pressure vessel steels

International Nuclear Information System (INIS)

Brumovsky, M.; Vacek, M.

1975-01-01

A Standard Research Programme on Irradiation Embrittlement of Pressure Vessel Steels was approved by the Coordinating Meeting on the 12th May 1972 at the Working Group on Engineering Aspects of Irradiation Embrittlement of Pressure Vessel Steels. This Working Group was set up by the International Atomic Energy Agency in Vienna. Seven countries with their research institutes agreed on doing irradiation experiments according to the approved programme on steel A533 B from the U.S. HSST Programme. The Czechoslovak contribution covering tensile and impact testing of non-irradiated steel and steel irradiated at 280degC to 1.3 x 10 23 n/m 2 (E above 1 MeV) is presented in this report. As an additional part the same set of experiments was carried out on two additional steels - A 542 and A 543, made in SKODA Works for comparison of their irradiation embrittlement and hardening with A533 B steel. (author)

Structure Optimization and Numerical Simulation of Nozzle for High Pressure Water Jetting

Directory of Open Access Journals (Sweden)

Shuce Zhang

2015-01-01

Full Text Available Three kinds of nozzles normally used in industrial production are numerically simulated, and the structure of nozzle with the best jetting performance out of the three nozzles is optimized. The R90 nozzle displays the most optimal jetting properties, including the smooth transition of the nozzle’s inner surface. Simulation results of all sample nozzles in this study show that the helix nozzle ultimately displays the best jetting performance. Jetting velocity magnitude along Y and Z coordinates is not symmetrical for the helix nozzle. Compared to simply changing the jetting angle, revolving the jet issued from the helix nozzle creates a grinding wheel on the cleaning surface, which makes not only an impact effect but also a shearing action on the cleaning object. This particular shearing action improves the cleaning process overall and forms a wider, effective cleaning range, thus obtaining a broader jet width.

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)

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.

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.

Rupture tests with reactor pressure vessel head models

International Nuclear Information System (INIS)

Talja, H.; Keinaenen, H.; Hosio, E.; Pankakoski, P.H.; Rahka, K.

2003-01-01

In the LISSAC project (LImit Strains in Severe ACcidents), partly funded by the EC Nuclear Fission and Safety Programme within the 5th Framework programme, an extensive experimental and computational research programme is conducted to study the stress state and size dependence of ultimate failure strains. The results are aimed especially to make the assessment of severe accident cases more realistic. For the experiments in the LISSAC project a block of material of the German Biblis C reactor pressure vessel was available. As part of the project, eight reactor pressure vessel head models from this material (22 NiMoCr 3 7) were tested up to rupture at VTT. The specimens were provided by Forschungszentrum Karlsruhe (FzK). These tests were performed under quasistatic pressure load at room temperature. Two specimens sizes were tested and in half of the tests the specimens contain holes describing the control rod penetrations of an actual reactor pressure vessel head. These specimens were equipped with an aluminium liner. All six tests with the smaller specimen size were conducted successfully. In the test with the large specimen with holes, the behaviour of the aluminium liner material proved to differ from those of the smaller ones. As a consequence the experiment ended at the failure of the liner. The specimen without holes yielded results that were in very good agreement with those from the small specimens. (author)

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

Improvements in the UT Inspection of vessel nozzles. Array application

International Nuclear Information System (INIS)

Tanarro, A.; Garcia, A.; Izquierdo, J.

1998-01-01

Automatic ultrasonic inspection of certain components in nuclear power plants, together with problems related to access of same, result in other difficulties due to the complexity of their geometry and the apparent orientation of possible defects. Array technology, recently developed on the basis of the theoretical principals of phased array technique, has meant that it is now possible to advance in the characterisation, localisation, and sizing of the defects in these components. This has been possible thanks to the discovery of synthetic materials which have allowed us to design and manufacture a new group of ultrasonic transducers. To these we may add new developments in electronics and computer sciences which have facilitated the building of high-powered control systems. This report discusses the work carried out by Tecnatom and Iberdrola in the field of automatic ultrasonic inspection of the vessel nozzles by means of array technology in the BWR at the Cofrentes Nuclear Power Station. The aims of this work were: - To facilitate the detection, characterisation, sizing and positioning of defects - To simplify and improve ultrasonic inspection in order to reduce acquisition times and the cost of same In order to achieve these results the following items were developed: - New array transducers were designed and manufactured - A new data acquisition system was developed - New programs for analysing data and for simulating ultrasonic testing was developed - The results have been validated in mock up. (Author)

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

International Nuclear Information System (INIS)

Scharge, J.; Wehowsky, P.; Zeibig, H.

1978-01-01

The in-service inspection program of reactor pressure vessels is mainly based on the ultra-sonic method, visual checking of inner and outer surfaces as well as pressure and leak tests. The test procedure require a design of the pressure vessel suitable for the test methods and the possibility to remove the pressure vessel internals. For the outside inspection a gap of sufficient width is mandatory. The present status of the ultra-sonic method and of the inner and outer manipulators affords to conduct the in-service inspection program in form of automatic checkings. The in-service inspection program for NCS-80, the Nuclear Container-Ship design of 80,000 shp, is integrated in the refueling periods due to the request for a high availability of the ship and reactor plant

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.

Initiation and arrest - two approaches to pressure vessel safety

International Nuclear Information System (INIS)

Brumovsky, M.; Filip, R.; Stepanek, S.

1976-01-01

The safety analysis is described of the reactor pressure vessel related to brittle fracture based on the fracture mechanics theory using two different approximations, i.e., the Crack Arrest Temperature (CAT) or Nil Ductility Temperature (NDT), and fracture toughness. The variation of CAT with stress was determined for different steel specimens of 120 to 200 mm in thickness. A diagram is shown of CAT variation with stress allowing the determination of crack arrest temperature for all types of commonly used steels independently of the NDT initial value. The diagram also shows that the difference between fracture transition elastic (FTE) and NDT depends on the type of material and determines the value of the ΔTsub(sigma) factor typical of the safety coefficient. The so-called fracture toughness reference value Ksub(IR) is recommended for the computation of pressure vessel criticality. Also shown is a defect analysis diagram which may be used for the calculation of pressure vessel safety prior to and during operation and which may also be used in making the decision on what crack sizes are critical, what cracks may be arrested and what cracks are likely to expand. The diagram is also important for the fact that it is material-independent and may be employed for the estimates of pre-operational and operational inspections and for pressure vessel life prediction. It is generally applicable to materials of greater thickness in the region where the validity of linear elastic fracture mechanics is guaranteed. (J.P.)

Combustor exhaust-emissions and blowout-limits with diesel number 2 and Jet A fuels utilizing air-atomizing and pressure-atomizing nozzles

Science.gov (United States)

Ingebo, R. D.; Norgren, C. T.

1975-01-01

The effect of fuel properties on exhaust emissions and blowout limits of a high-pressure combustor segment is evaluated using a splash-groove air-atomizing fuel injector and a pressure-atomizing simplex fuel nozzle to burn both diesel number 2 and Jet A fuels. Exhaust emissions and blowout data are obtained and compared on the basis of the aromatic content and volatility of the two fuels. Exhaust smoke number and emission indices for oxides of nitrogen, carbon monoxide, and unburned hydrocarbons are determined for comparison. As compared to the pressure-atomizing nozzle, the air-atomizing nozzle is found to reduce nitrogen oxides by 20%, smoke number by 30%, carbon monoxide by 70%, and unburned hydrocarbons by 50% when used with diesel number 2 fuel. The higher concentration of aromatics and lower volatility of diesel number 2 fuel as compared to Jet A fuel appears to have the most detrimental effect on exhaust emissions. Smoke number and unburned hydrocarbons are twice as high with diesel number 2 as with Jet A fuel.

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

Helium leak testing of large pressure vessels or subassemblies

International Nuclear Information System (INIS)

Hopkins, J.S.; Valania, J.J.

1977-01-01

Specifications for pressure-vessel components [such as the intermediate heat exchangers (IHX)] for service in the liquid metal fast breeder reactor facilities require helium leak testing of pressure boundaries to very exacting standards. The experience of Foster Wheeler Energy Corporation (FWEC) in successfully leak-testing the IHX shells and bundle assemblies now installed in the Fast Flux Test Facility at Richland, WA is described. Vessels of a somewhat smaller size for the closed loop heat exchanger system in the Fast Flux Test Facility have also been fabricated and helium leak tested for integrity of the pressure boundary by FWEC. Specifications on future components call for helium leak testing of the tube to tubesheet welds of the intermediate heat exchangers

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.

Assessment of the integrity of WWER type reactor pressure vessels

International Nuclear Information System (INIS)

Brumovsky, M.

1995-01-01

Procedures are given for the assessment of the residual lifetime of reactor pressure vessels with respect to a sudden failure, the lifetime of vessels with defects disclosed during in-service inspections, and the fatigue or corrosion-mechanical lifetime. Also outlined are the ways of assessing the effects of major degradation mechanisms, i.e. radiation embrittlement, thermal aging, and fatigue damage, including the use of calculated values and experimental examination, by means of surveillance specimens in particular. All results of assessment performed so far indicate that the life of reactor pressure vessels at the Dukovany, Jaslovske Bohunice, and Temelin nuclear power plants is well secured. 7 figs., 3 refs

Numerical study of base pressure characteristic curve for a four-engine clustered nozzle configuration

Science.gov (United States)

Wang, Ten-See

1993-07-01

Excessive base heating has been a problem for many launch vehicles. For certain designs such as the direct dump of turbine exhaust in the nozzle section and at the nozzle lip of the Space Transportation Systems Engine (STME), the potential burning of the turbine exhaust in the base region has caused tremendous concern. Two conventional approaches have been considered for predicting the base environment: (1) empirical approach, and (2) experimental approach. The empirical approach uses a combination of data correlations and semi-theoretical calculations. It works best for linear problems, simple physics and geometry. However, it is highly suspicious when complex geometry and flow physics are involved, especially when the subject is out of historical database. The experimental approach is often used to establish database for engineering analysis. However, it is qualitative at best for base flow problems. Other criticisms include the inability to simulate forebody boundary layer correctly, the interference effect from tunnel walls, and the inability to scale all pertinent parameters. Furthermore, there is a contention that the information extrapolated from subscale tests with combustion is not conservative. One potential alternative to the conventional methods is computational fluid dynamics (CFD), which has none of the above restrictions and is becoming more feasible due to maturing algorithms and advancing computer technology. It provides more details of the flowfield and is only limited by computer resources. However, it has its share of criticisms as a predictive tool for base environment. One major concern is that CFD has not been extensively tested for base flow problems. It is therefore imperative that CFD be assessed and benchmarked satisfactorily for base flows. In this study, the turbulent base flowfield of a experimental investigation for a four-engine clustered nozzle is numerically benchmarked using a pressure based CFD method. Since the cold air was the

Stochastic simulation of PWR vessel integrity for pressurized thermal shock conditions

International Nuclear Information System (INIS)

Jackson, P.S.; Moelling, D.S.

1984-01-01

A stochastic simulation methodology is presented for performing probabilistic analyses of Pressurized Water Reactor vessel integrity. Application of the methodology to vessel-specific integrity analyses is described in the context of Pressurized Thermal Shock (PTS) conditions. A Bayesian method is described for developing vessel-specific models of the density of undetected volumetric flaws from ultrasonic inservice inspection results. Uncertainty limits on the probabilistic results due to sampling errors are determined from the results of the stochastic simulation. An example is provided to illustrate the methodology

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

International Nuclear Information System (INIS)

Takahashi, W.K.; Al-Qureshi, H.A.

1982-01-01

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

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

International Nuclear Information System (INIS)

Bryan, R.H.; Merkle, J.G.; Raftenberg, M.N.; Robinson, G.C.; Smith, J.E.

1975-11-01

The second series of intermediate vessel tests were crack initiation fracture tests of 6-in.-thick 39-in.-OD steel vessels with sharp surface flaws approximately 2 1 / 2 in. deep by 8 in. long in the longitudinal weld seams of the test cylinders. Fracture was initiated by means of hydraulic pressurization. One vessel was tested at each of three temperatures: 75, 130, and 190 0 F. Pretest analyses were made to predict the failure pressures and strains. Fracture toughness data obtained by equivalent-energy analysis of precracked Charpy-V tests and compact-tension specimen tests were used in the fracture analyses. The vessels behaved generally as had been expected. Posttest fracture analyses were also performed for each vessel. Detailed discussions of the fracture analysis methods developed in support of the vessel tests described are included. 34 references

Milestones in pressure vessel technology

International Nuclear Information System (INIS)

Spence, J.; Nash, D.H.

2004-01-01

The progress of pressure vessel technology over the years has been influenced by many important events. This paper identifies a number of 'milestones' which have provided a stimulus to analysis methods, manufacturing, operational processes and new pressure equipment. The formation of a milestone itself along with its subsequent development is often critically dependent on the work of many individuals. It is postulated that such developments takes place in cycles, namely, an initial idea, followed sometimes by unexpected failures, which in turn stimulate analysis or investigation, and when confidence is established, followed finally by the emergence of codes ad standards. Starting from the industrial revolution, key milestones are traced through to the present day and beyond

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

Study of a flapper-nozzle system for a water hydraulic servovalve; Suiatsu survo ben ni mochiiru nozzle flapper kei no kenkyu

Energy Technology Data Exchange (ETDEWEB)

Urata, E.; Nakao, Y. [Kanagawa University, Kanagawa (Japan). Faculty of Engineering

1997-06-25

The paper discusses the characteristics of a flapper-nozzle system for water hydraulic servovalves. High pressure water at the supply port is first used as the working fluid for the hydrostatic bearings supporting the spool. Spool valve stiction induced by poor lubrication with water is thus avoided. The fluid is then led to the ends of the spool and is used as the working fllid of the flapper-nozzle system. In the new flapper-nozzle system the circumferential clearance of the spool becomes a laminar restriction that substitutes fixed orifice used in conventional servovalves. The linearity in the pressure-displacement relationship of the new flapper-nozzle system is better than that of conventional fixed orifice systems. 7 refs., 11 figs., 2 tabs.

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

Cylindrical pressure vessel constructed of several layers

International Nuclear Information System (INIS)

Yamauchi, Takeshi.

1976-01-01

For a cylindrical pressure vessel constructed of several layers whose jacket has at least one circumferential weld joining the individual layers, it is proposed to provide this at least at the first bending line turning point (counting from the weld between the jacket and vessel floor), which the sinusoidally shaped jacket has. The section of the jacket extending in between should be made as a full wall section. The proposal is based on calculations of the bending stiffness of cylindrical jackets, which could not yet be confirmed for jackets having several layers. (UWI) [de

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

Pressure vessels fabricated with high-strength wire and electroformed nickel

Science.gov (United States)

Roth, B.

1966-01-01

Metal pressure vessels of various shapes having high strength-to-weight ratios are fabricated by using known techniques of filament winding and electroforming. This eliminates nonuniform wall thickness and unequal wall strength which resulted from welding formed vessel segments together.

Reliability aspects of radiation damage in reactor pressure vessel mterials

International Nuclear Information System (INIS)

Brumovsky, M.

1985-01-01

The service life estimate is a major factor in the evaluation of the operating reliability and safety of a nuclear reactor pressure vessel. The evaluation of the service life of the pressure vessel is based on a comparison of fracture toughness values with stress intensity factors. Notch toughness curves are used for the indirect determination of fracture toughness. The dominant degradation effect is radiation embrittlement. Factors having the greatest effect on the result are the properties of the starting material of the vessel and the impurity content, mainly the Cu and P content. The design life is affected by the evaluation of residual lifetime which is made by periodical nondestructive inspections and using surveillance samples. (M.D.)

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)

Microstructural evolution in neutron irradiated reactor pressure vessel steels

International Nuclear Information System (INIS)

English, C.A.; Phythian, W.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. The microstructural evolution in neutron irradiated reactor pressure vessel steels is described. The damage mechanisms are elaborated and techniques for examining the microstructure are suggested. The importance of the initial damage event is analysed, and the microstructural evolution in RPV steels is examined

Experimental study of micron size droplets in a two phase flow in a converging - diverging nozzle

International Nuclear Information System (INIS)

Jurski, Kristine

1997-01-01

The fluid present in a pressurized vessel in normal operation is generally a mono-phase one. In accidental regime (a breach for example), a two-phase (ring and/or dispersed) flow appears and the flow is submitted to large accelerations when passing through the breach, and is then dispersed in the atmosphere. This research thesis reports an experimental simulation of an accident by generating, through a discharge of an upstream vessel into a downstream vessel, a strongly accelerated gaseous-liquid two-phase flow, with an essentially dispersed configuration in a convergent-divergent nozzle. In order to characterize the speed and diameter evolution of the dispersed liquid phase, the author reports a comparative study of two different liquid aerosols: micron-size droplets of di-octyl phthalate (DOP) of known concentration and diameter, and water droplets obtained by heterogeneous spontaneous condensation [fr

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.

Nuclear reactor vessel inspection apparatus

International Nuclear Information System (INIS)

Blackstone, E.G.; Lofy, R.A.; Williams, L.P.

1979-01-01

Apparatus for the in situ inspection of a nuclear reactor vessel to detect the location and character of flaws in the walls of the vessel, in the welds joining the various sections of the vessel, in the welds joining attachments such as nozzles, elbows and the like to the reactor vessel and in such attachments wherein an inspection head carrying one or more ultrasonic transducers follows predetermined paths in scanning the various reactor sections, welds and attachments

TMI-2 instrument nozzle examinations at Argonne National Laboratory

International Nuclear Information System (INIS)

Neimark, L.A.; Shearer, T.L.; Purohit, A.; Hins, A.G.

1993-09-01

Six of the 14 instrument-penetration-tube nozzles removed from the lower head of TMI-2 were examined to identify damage mechanisms, provide insight to the fuel relocation scenario, and provide input data to the margin-to-failure analysis. Visual inspection, gamma scanning, metallography, microhardness measurements, and scanning electron microscopy were used to obtain the desired information. The results showed varying degrees of damage to the lower head nozzles, from ∼50% melt-off to no damage at all to near-neighbor nozzles. The elevations of nozzle damage suggested that the lower elevations (near the lower head) were protected from molten fuel, apparently by an insulating layer of fuel debris. The pattern of nozzle damage was consistent with fuel movement toward the hot-spot location identified in the vessel wall. Evidence was found for the existence of a significant quantity of control assembly debris on the lower head before the massive relocation of fuel occurred

Final report for the 2nd Ex-Vessel Neutron Dosimetry Installations and Evaluations for Yonggwang Unit 2 Reactor Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Gong, Un Sik; Choi, Kwon Jae; Chung, Kyoung Ki; Kim, Kwan Hyun; Chang, Jong Hwa; Ha, Jea Ju

2008-01-15

This report describes a neutron fluence assessment performed for the Yonggwang Unit 2 pressure vessel beltline region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the beltline region of the pressure vessel. During Cycle 16 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Yonggwang Unit 2 to provide continuous monitoring of the beltline region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 16.

Final report for the 1st ex-vessel neutron dosimetry installations and evaluations for Kori unit 2 reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

2006-11-15

This report describes a neutron fluence assessment performed for the Kori unit 2 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 20 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 2 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 20.

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

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Li, Nam Jin; Hong, Joon Wha

2007-01-15

This report describes a neutron fluence assessment performed for the Yonggwang Unit 1 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During Cycle 16 of reactor operation, 2nd Ex-Vessel Neutron Dosimetry Program was instituted at Yonggwang Unit 1 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 16.

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

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Gong, Un Sik; Choi, Kwon Jae; Chung, Kyoung Ki; Kim, Kwan Hyun; Chang, Jong Hwa; Ha, Jea Ju

2008-01-15

This report describes a neutron fluence assessment performed for the Kori Unit 2 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During Cycle 21 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 2 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 21.

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

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

2006-11-15

This report describes a neutron fluence assessment performed for the Kori unit 4 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 16 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 4 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 16.

U.S. and French approaches to reactor pressure vessel integrity

International Nuclear Information System (INIS)

Griesbach, T.J.; Buchalet, C.; Server, W.L.

1990-01-01

The effects of radiation embrittlement on the reactor pressure vessel must be considered for continued safe operation of nuclear power plants. The consequences of radiation embrittlement require detailed assessments of the margins of safety against brittle fracture of the vessel. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code and U.S. Regulations often use conservative approaches for these assessments which can eventually lead to severe operational hardships for some plants. Taking a look at alternative integrity approaches, such as those demonstrated in France, could ultimately result in improved ASME Code and Regulatory limits. The French studies have shown the significance of performing proper in- service inspections to reliably show that no defects larger than a predetermined size (or class) exist in the inspected region of a vessel. The predetermined size is based upon previous studies on the types of manufacturing defects which can potentially exist in French vessels. Enhanced linear elastic and elastic-plastic fracture mechanics methodologies can be applied to evaluate such defects to assure that brittle fracture will not occur

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

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.)

Strain measurement in and analysis for hydraulic test of CPR1000 reactor pressure vessel

International Nuclear Information System (INIS)

Zhou Dan; Zhuang Dongzhen

2013-01-01

The strain measurement in hydraulic test of CPR1000 reactor pressure vessel performed in Dongfang Heavy Machinery Co., Ltd. is introduced. The detail test scheme and method was introduced and the measurement results of strain and stress was given. Meanwhile the finite element analysis was performed for the pressure vessel, which was generally matched with the measurement results. The reliability of strain measurement was verified and the high strength margin of vessel was shown, which would give a good reference value for the follow-up hydraulic tests and strength analysis of reactor pressure vessel. (authors)

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

International Nuclear Information System (INIS)

Dickson, T.L.; Cheverton, R.D.; Bryson, J.W.; Bass, B.R.; Shum, D.K.M.; Keeney, J.A.

1993-08-01

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

Structural considerations in design of lightweight glass-fiber composite pressure vessels

Science.gov (United States)

Faddoul, J. R.

1973-01-01

The design concepts used for metal-lined glass-fiber composite pressure vessels are described, comparing the structural characteristics of the composite designs with each other and with homogeneous metal pressure vessels. Specific design techniques and available design data are identified. The discussion centers around two distinctly different design concepts, which provide the basis for defining metal lined composite vessels as either (1) thin-metal lined, or (2) glass fiber reinforced (GFR). Both concepts are described and associated development problems are identified and discussed. Relevant fabrication and testing experience from a series of NASA-Lewis Research Center development efforts is presented.

Nature of convection-stabilized dc arcs in dual-flow nozzle geometry

International Nuclear Information System (INIS)

Serbetci, I.; Nagamatsu, H.T.

1990-01-01

In this paper, an experimental investigation of the steady-state low-current air arcs in a dual-flow nozzle system is presented. First, the cold flow with no arc as determined for various nozzle geometries, i.e., two- and three-dimensional and orifice nozzles, and nozzle pressure ratios. Supersonic flow separation and oblique and detached shock waves were observed in the flow field. Using a finite-element computer program, the Mach number contours were determined in the flow field for various nozzle-gap spacings and pressure ratios. In addition, the dc arc voltage and current measurements were made for an electrode gap spacing of ∼ 5.5 cm and current levels of I ∼ 25, 50, and 100 A for the three nozzle geometries. The arc voltage and arc power increased rapidly as the flow speed increased from zero to sonic velocity at the nozzle throat. The shock waves in the converging-diverging nozzles resulted in a decrease in the overall resistance by about 15 percent

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)

Development of acoustically lined ejector technology for multitube jet noise suppressor nozzles by model and engine tests over a wide range of jet pressure ratios and temperatures

Science.gov (United States)

Atvars, J.; Paynter, G. C.; Walker, D. Q.; Wintermeyer, C. F.

1974-01-01

An experimental program comprising model nozzle and full-scale engine tests was undertaken to acquire parametric data for acoustically lined ejectors applied to primary jet noise suppression. Ejector lining design technology and acoustical scaling of lined ejector configurations were the major objectives. Ground static tests were run with a J-75 turbojet engine fitted with a 37-tube, area ratio 3.3 suppressor nozzle and two lengths of ejector shroud (L/D = 1 and 2). Seven ejector lining configurations were tested over the engine pressure ratio range of 1.40 to 2.40 with corresponding jet velocities between 305 and 610 M/sec. One-fourth scale model nozzles were tested over a pressure ratio range of 1.40 to 4.0 with jet total temperatures between ambient and 1088 K. Scaling of multielement nozzle ejector configurations was also studied using a single element of the nozzle array with identical ejector lengths and lining materials. Acoustic far field and near field data together with nozzle thrust performance and jet aerodynamic flow profiles are presented.

Pressure vessel design

International Nuclear Information System (INIS)

Annaratone, D.

2007-01-01

This book guides through general and fundamental problems of pressure vessel design. It moreover considers problems which seem to be of lower importance but which turn out to be crucial in the design phase. The basic approach is rigorously scientific with a complete theoretical development of the topics treated, but the analysis is always pushed so far as to offer concrete and precise calculation criteria that can be immediately applied to actual designs. This is accomplished through appropriate algorithms that lead to final equations or to characteristic parameters defined through mathematical equations. The first chapter describes how to achieve verification criteria, the second analyzes a few general problems, such as stresses of the membrane in revolution solids and edge effects. The third chapter deals with cylinders under pressure from the inside, while the fourth focuses on cylinders under pressure from the outside. The fifth chapter covers spheres, and the sixth is about all types of heads. Chapter seven discusses different components of particular shape as well as pipes, with special attention to flanges. The eighth chapter discusses the influence of holes, while the ninth is devoted to the influence of supports. Finally, chapter ten illustrates the fundamental criteria regarding fatigue analysis. Besides the unique approach to the entire work, original contributions can be found in most chapters, thanks to the author's numerous publications on the topic and to studies performed ad hoc for this book. (orig.)

Reactor water spontaneous circulation structure in reactor pressure vessel

International Nuclear Information System (INIS)

Takahashi, Kazumi

1998-01-01

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

Shock unsteadiness in a thrust optimized parabolic nozzle

Science.gov (United States)

Verma, S. B.

2009-07-01

This paper discusses the nature of shock unsteadiness, in an overexpanded thrust optimized parabolic nozzle, prevalent in various flow separation modes experienced during start up {(δ P0 /δ t > 0)} and shut down {(δ P0/δ t The results are based on simultaneously acquired data from real-time wall pressure measurements using Kulite pressure transducers, high-speed schlieren (2 kHz) of the exhaust flow-field and from strain-gauges installed on the nozzle bending tube. Shock unsteadiness in the separation region is seen to increase significantly just before the onset of each flow transition, even during steady nozzle operation. The intensity of this measure ( rms level) is seen to be strongly influenced by relative locations of normal and overexpansion shock, the decrease in radial size of re-circulation zone in the back-flow region, and finally, the local nozzle wall contour. During restricted shock separation, the pressure fluctuations in separation region exhibit periodic characteristics rather than the usually observed characteristics of intermittent separation. The possible physical mechanisms responsible for the generation of flow unsteadiness in various separation modes are discussed. The results are from an experimental study conducted in P6.2 cold-gas subscale test facility using a thrust optimized parabolic nozzle of area-ratio 30.

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

International Nuclear Information System (INIS)

1998-10-01

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. A separate abstract was prepared for the introduction and for each of the eleven chapters, which are: 1. Reactor Pressure Vessel Design, 2. Reactor Pressure Materials, 3. WWER Pressure Vessels, 4. Determination of Mechanical Properties, 5. Neutron Exposure, 6. Methodology of Irradiation Experiments, 7. Effect of Irradiation on Mechanical Properties, 8. Mechanisms of Irradiation Embrittlement, 9. Modelling of Irradiation Damage, 10. Annealing of Irradiation Damage, 11. Safety Assessment using Surveillance Programmes and Data Bases

Energy and impacts of pressure vessel explosions

International Nuclear Information System (INIS)

Kurttila, H.

1999-01-01

In this paper the explosion energy is considered to be same as the energy of pressure vessel discharge. This is the maximum energy which can be obtained from the process. The energy can be used or it can cause the violence of an explosion accident. (orig.)

Application of fracture mechanics to fatigue in pressure vessels

International Nuclear Information System (INIS)

Ghavami, K.

1982-01-01

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

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)

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

AE/flaw characterization for nuclear pressure vessels

International Nuclear Information System (INIS)

Hutton, P.H.; Kurtz, R.J.; Pappas, R.A.

1984-01-01

This chapter discusses the use of acoustic emission (AE) detected during continuous monitoring to identify and evaluate growing flaws in pressure vessels. Off-reactor testing and on-reactor testing are considered. Relationships for identifying acoustic emission (AE) from crack growth and using the AE data to estimate flaw severity have been developed experimentally by laboratory testing. The purpose of the off-reactor vessel test is to evaluate AE monitoring/interpretation methodology on a heavy section steel vessel under simulated reactor operating conditions. The purpose of on-reactor testing is to evaluate the capability of a monitor system to function in the reactor environment, calibrate the ability to detect AE signals, and to demonstrate that a meaningful criteria can be established to prevent false alarms. An expanded data base is needed from application testing and methodology standardization

Computer Graphic Design Using Auto-CAD and Plug Nozzle Research

Science.gov (United States)

Rogers, Rayna C.

2004-01-01

The purpose of creating computer generated images varies widely. They can be use for computational fluid dynamics (CFD), or as a blueprint for designing parts. The schematic that I will be working on the summer will be used to create nozzles that are a part of a larger system. At this phase in the project, the nozzles needed for the systems have been fabricated. One part of my mission is to create both three dimensional and two dimensional models on Auto-CAD 2002 of the nozzles. The research on plug nozzles will allow me to have a better understanding of how they assist in the thrust need for a missile to take off. NASA and the United States military are working together to develop a new design concept. On most missiles a convergent-divergent nozzle is used to create thrust. However, the two are looking into different concepts for the nozzle. The standard convergent-divergent nozzle forces a mixture of combustible fluids and air through a smaller area in comparison to where the combination was mixed. Once it passes through the smaller area known as A8 it comes out the end of the nozzle which is larger the first or area A9. This creates enough thrust for the mechanism whether it is an F-18 fighter jet or a missile. The A9 section of the convergent-divergent nozzle has a mechanism that controls how large A9 can be. This is needed because the pressure of the air coming out nozzle must be equal to that of the ambient pressure other wise there will be a loss of performance in the machine. The plug nozzle however does not need to have an A9 that can vary. When the air flow comes out it can automatically sense what the ambient pressure is and will adjust accordingly. The objective of this design is to create a plug nozzle that is not as complicated mechanically as it counterpart the convergent-divergent nozzle.

Nozzle flow calculation for real gases

International Nuclear Information System (INIS)

Bier, K.; Ehrler, F.; Hartz, U.; Kissau, G.

1977-01-01

The flow of CHF 2 Cl vapor (refrigerant R 22) through a Laval nozzle of annular geometry has been investigated in the region near the saturation line with stagnation pressures up to 85 per cent of the critical pressure. Static pressure profiles measured along the nozzle axis were found in good agreement with profiles calculated for one-dimensional isentropic flow of the real gas the thermal properties of which were derived from an equation of state proposed previously by Rombusch. Minor deviations between measured and calculated static pressure curves occur in the supersonic part of the mozzle, especially when supersaturated states of the vapour are passed. These deviations can be attributed to uncertainties in the calculation of the enthalpy and to a small influence of the static pressure probe. An additional investigation was concerned with an approximate calculation of the nozzle flow of real gases. In this approximation the well known relations of ideal gas dynamics are applied, the ratio of specific heats for the ideal gas being replaced, however, by a suitably adapted isentropic exponent, which can be determined e.g. from measured values of the Laval pressure or of the mass flow. For pressure ratios p/po between 1 and approximately 0.1, corresponding to Mach numbers up to approximately 2.2, all the interesting properties of the investigated flow of CHF 2 Cl vapour are approximated within a few per cent. (orig.) [de

Single pressure vessel (SPV) nickel-hydrogen battery design

Energy Technology Data Exchange (ETDEWEB)

Coates, D.; Grindstaff, B.; Fox, C. [Eagle-Picher Industries, Inc., Joplin, MO (United States)

1995-07-01

Single pressure vessel (SPV) technology combines an entire multi-cell nickel-hydrogen (NiH{sub 2}) space battery within a single pressure vessel. SPV technology has been developed to improve the performance (volume/mass) of the NiH{sub 2} system at the battery level and ultimately to reduce overall battery cost and increase system reliability. Three distinct SPV technologies are currently under development and in production. Eagle-Picher has license to the COMSAT Laboratories technology, as well as internally developed independent SPV technology. A third technology resulted from the acquisition of Johnson Controls NiH{sub 2} battery assets in June, 1994. SPV batteries are currently being produced in 25 ampere-hour (Ah), 35 Ah and 50 Ah configurations. The battery designs have an overall outside diameter of 10 inches (25.4 centimeters).

INVESTIGATION OF FLOW BEHAVIOR IN MINIMUM QUANTITY LUBRICATION NOZZLE FOR END MILLING PROCESSES

Directory of Open Access Journals (Sweden)

M.S. Najiha

2012-12-01

Full Text Available Minimum quantity lubrication (MQL is a sustainable manufacturing technique that has replaced conventional flooded lubrication methods and dry machining. In the MQL technique, the lubricant is sprayed onto the friction surfaces through nozzles through small pneumatically-operated pumps. This paper presents an investigation into the flow behavior of the lubricant and air mixture under certain pressures at the tip of a nozzle specially designed for MQL. The nozzle used is an MQL stainless steel nozzle, 6.35 mm in diameter. Computational fluid dynamics is used to determine the flow pattern at the tip of the nozzle where the lubricant and compressed air are mixed to form a mist. The lubricant volume flow is approximately 0.08 ml/cycle of the pump. A transient, pressure-based, three-dimensional analysis is performed with a viscous, realizable k-ε model. The results are obtained in the form of vector plots and flow fields. The flow mixing at the tip of the nozzle is wholly shown through the flow fields and vector plots. This study provides an insight into the flow distribution at the tip of the nozzle for a certain pressure to aid modifications in the design of the nozzle for future MQL studies. It attainable aids to determine the correct pressure for the air jet at the nozzle tip.

New paradigm for prediction of radiation life-time of reactor pressure vessel

International Nuclear Information System (INIS)

Kotrechko, S.A.; Meshkov, Yu.Ya.; Neklyudov, I.M.; Revka, V.N.

2011-01-01

New paradigm for prediction of radiation life-time of reactor pressure vessel is presented. Equation for limiting state of reactor pressure vessel wall with crack-like defect is obtained. It is exhibited that the value of critical fluence Φ c may be determined not by shift of critical temperature of fracture of surveillance specimen, which is indirect characteristic, but by direct method, namely, by the condition of initiation of brittle fracture of irradiated metal ahead of a crack in RPV wall. Within the framework of engineering version of LA to fracture the technique for Φ c ascertainment is developed. Prediction of Φ c for WWER pressure vessels demonstrates potentialities of this technique.

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)

EDF field experience of 182 J-Groove welds on CRDMs and SG channel head nozzles

International Nuclear Information System (INIS)

Duisabeau, L.; Deforge, D.; Thebault, Y.; Stindel, M.; Lemaire, E.

2011-01-01

The Reactor Pressure Vessel Head (RPVH) replacement program, which began after a leak occurrence in a vessel head nozzle in Alloy 600 at Bugey Unit 3, was a unique opportunity to perform an extended inspection program on the welds from the decommissioned RPV heads. This paper presents the actual results of this program. More than 800 CRDM J groove welds from 18 decommissioned RPV heads were inspected by automatic dye penetrant testing. Detected indications were characterized by viewing tools specifically developed and in some specific cases, by destructive investigations in hot lab. Some welding defects were observed but no indication corresponding to stress corrosion cracking (SCC) was detected at the welds wet surface nor propagation from welding manufacturing defects, including the weld with the longest operating time on EDF power plants (170 000 h). Very few cases of SCC propagation from Alloy 600 to Alloy 182 are reported. One case of initiation at the weld root pass was observed. From design, the weld root pass (mechanically loaded) of CRDM (Control Rod Drive Mechanism) nozzles is not in contact with primary water and the cracking observed occurred after a through wall cracking of the Alloy 600 tube, enabling primary water to wet the root pass. Concerning the steam generator (SG) drain nozzle, the alloy 182 weld root is directly in contact with primary water. In June 2008, a primary water leakage was suspected on a steam generator bowl drain while conducting a bare metal visual examination during the plant's outage. Dye penetrant testing of the weld and metallographic replica were implemented during the 2008 and 2009 refuelling outages to confirm a leakage by SCC. Manufacturing reports analyses revealed that the drain nozzle weld was repaired and had not been stress relieved during manufacturing. EDF has decided to plug this nozzle and to enforce the maintenance policy for similar components with the same manufacturing specificity. Regarding national and

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)