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

International Nuclear Information System (INIS)

Kotilainen, H.

1980-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Sarkimo, M; Rintamac, R

1988-12-31

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

Influence of crack depth on the fracture toughness of reactor pressure vessel steel

International Nuclear Information System (INIS)

Theiss, T.J.; Bryson, J.W.

1991-01-01

The Heavy Section Steel Technology Program (HSST) at Oak Ridge National Laboratory (ORNL) is investigating the influence of flaw depth on the fracture toughness of reactor pressure vessel (RPV) steel. Recently, it has been shown that, in notched beam testing, shallow cracks tend to exhibit an elevated toughness as a result of a loss of constraint at the crack tip. The loss of constraint takes place when interaction occurs between the elastic-plastic crack-tip stress field and the specimen surface nearest the crack tip. An increased shallow-crack fracture toughness is of interest to the nuclear industry because probabilistic fracture-mechanics evaluations show that shallow flaws play a dominant role in the probability of vessel failure during postulated pressurized-thermal-shock (PTS) events. Tests have been performed on beam specimens loaded in 3-point bending using unirradiated reactor pressure vessel material (A533 B). Testing has been conducted using specimens with a constant beam depth (W = 94 mm) and within the lower transition region of the toughness curve for A533 B. Test results indicate a significantly higher fracture toughness associated with the shallow flaw specimens compared to the fracture toughness determined using deep-crack (a/W = 0.5) specimens. Test data also show little influence of thickness on the fracture toughness for the current test temperature (-60 degree C). 21 refs., 5 figs., 3 tabs

Podoplanin immunopositive lymphatic vessels at the implant interface in a rat model of osteoporotic fractures.

Directory of Open Access Journals (Sweden)

Katrin Susanne Lips

Full Text Available Insertion of bone substitution materials accelerates healing of osteoporotic fractures. Biodegradable materials are preferred for application in osteoporotic patients to avoid a second surgery for implant replacement. Degraded implant fragments are often absorbed by macrophages that are removed from the fracture side via passage through veins or lymphatic vessels. We investigated if lymphatic vessels occur in osteoporotic bone defects and whether they are regulated by the use of different materials. To address this issue osteoporosis was induced in rats using the classical method of bilateral ovariectomy and additional calcium and vitamin deficient diet. In addition, wedge-shaped defects of 3, 4, or 5 mm were generated in the distal metaphyseal area of femur via osteotomy. The 4 mm defects were subsequently used for implantation studies where bone substitution materials of calcium phosphate cement, composites of collagen and silica, and iron foams with interconnecting pores were inserted. Different materials were partly additionally functionalized by strontium or bisphosphonate whose positive effects in osteoporosis treatment are well known. The lymphatic vessels were identified by immunohistochemistry using an antibody against podoplanin. Podoplanin immunopositive lymphatic vessels were detected in the granulation tissue filling the fracture gap, surrounding the implant and growing into the iron foam through its interconnected pores. Significant more lymphatic capillaries were counted at the implant interface of composite, strontium and bisphosphonate functionalized iron foam. A significant increase was also observed in the number of lymphatics situated in the pores of strontium coated iron foam. In conclusion, our results indicate the occurrence of lymphatic vessels in osteoporotic bone. Our results show that lymphatic vessels are localized at the implant interface and in the fracture gap where they might be involved in the removal of

Podoplanin Immunopositive Lymphatic Vessels at the Implant Interface in a Rat Model of Osteoporotic Fractures

Science.gov (United States)

Lips, Katrin Susanne; Kauschke, Vivien; Hartmann, Sonja; Thormann, Ulrich; Ray, Seemun; Kampschulte, Marian; Langheinrich, Alexander; Schumacher, Matthias; Gelinsky, Michael; Heinemann, Sascha; Hanke, Thomas; Kautz, Armin R.; Schnabelrauch, Matthias; Schnettler, Reinhard; Heiss, Christian; Alt, Volker; Kilian, Olaf

2013-01-01

Insertion of bone substitution materials accelerates healing of osteoporotic fractures. Biodegradable materials are preferred for application in osteoporotic patients to avoid a second surgery for implant replacement. Degraded implant fragments are often absorbed by macrophages that are removed from the fracture side via passage through veins or lymphatic vessels. We investigated if lymphatic vessels occur in osteoporotic bone defects and whether they are regulated by the use of different materials. To address this issue osteoporosis was induced in rats using the classical method of bilateral ovariectomy and additional calcium and vitamin deficient diet. In addition, wedge-shaped defects of 3, 4, or 5 mm were generated in the distal metaphyseal area of femur via osteotomy. The 4 mm defects were subsequently used for implantation studies where bone substitution materials of calcium phosphate cement, composites of collagen and silica, and iron foams with interconnecting pores were inserted. Different materials were partly additionally functionalized by strontium or bisphosphonate whose positive effects in osteoporosis treatment are well known. The lymphatic vessels were identified by immunohistochemistry using an antibody against podoplanin. Podoplanin immunopositive lymphatic vessels were detected in the granulation tissue filling the fracture gap, surrounding the implant and growing into the iron foam through its interconnected pores. Significant more lymphatic capillaries were counted at the implant interface of composite, strontium and bisphosphonate functionalized iron foam. A significant increase was also observed in the number of lymphatics situated in the pores of strontium coated iron foam. In conclusion, our results indicate the occurrence of lymphatic vessels in osteoporotic bone. Our results show that lymphatic vessels are localized at the implant interface and in the fracture gap where they might be involved in the removal of lymphocytes, macrophages

Fracture toughness requirements of reactor vessel material in evaluation of the safety analysis report of nuclear power plants

International Nuclear Information System (INIS)

Widia Lastana Istanto

2011-01-01

Fracture toughness requirements of reactor vessel material that must be met by applicants for nuclear power plants construction permit has been investigated in this paper. The fracture toughness should be described in the Safety Analysis Reports (SARs) document that will be evaluated by the Nuclear Energy Regulatory Agency (BAPETEN). Because BAPETEN does not have a regulations or standards/codes regarding the material used for the reactor vessel, especially in the fracture toughness requirements, then the acceptance criteria that applied to evaluate the fracture toughness of reactor vessel material refers to the regulations/provisions from the countries that have been experienced in the operation of nuclear power plants, such as from the United States, Japan and Korea. Regulations and standards used are 10 CFR Part 50, ASME and ASTM. Fracture toughness of reactor vessel materials are evaluated to ensure compliance of the requirements and provisions of the Regulatory Body and the applicable standards, such as ASME or ASTM, in order to assure a reliability and integrity of the reactor vessels as well as providing an adequate safety margin during the operation, testing, maintenance, and postulated accident conditions over the reactor vessel lifetime. (author)

Prevention of non-ductile fracture in 6061-T6 aluminum nuclear pressure vessels

International Nuclear Information System (INIS)

Yahr, G.T.

1995-01-01

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Committee has approved rules for the use of 6061-T6 and 6061-T651 aluminum for the construction of Class 1 welded nuclear pressure vessels for temperatures not exceeding 149 C (300 F). Nuclear Code Case N-519 allows the use of this aluminum in the construction of low temperature research reactors such as the Advanced Neutron Source. The rules for protection against non-ductile fracture are discussed. The basis for a value of 25.3 MPa √m (23 ksi √in.) for the critical or reference stress intensity factor for use in the fracture analysis is presented. Requirements for consideration of the effects of neutron irradiation on the fracture toughness are discussed

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-01

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

Master curve approach to monitor fracture toughness of reactor pressure vessels in nuclear power plants

International Nuclear Information System (INIS)

2009-10-01

A series of coordinated research projects (CRPs) have been sponsored by the IAEA, starting in the early 1970s, focused on neutron radiation effects on reactor pressure vessel (RPV) steels. The purpose of the CRPs was to develop correlative comparisons to test the uniformity of results through coordinated international research studies and data sharing. The overall scope of the eighth CRP (CRP-8), Master Curve Approach to Monitor Fracture Toughness of Reactor Pressure Vessels in Nuclear Power Plants, has evolved from previous CRPs which have focused on fracture toughness related issues. The ultimate use of embrittlement understanding is application to assure structural integrity of the RPV under current and future operation and accident conditions. The Master Curve approach for assessing the fracture toughness of a sampled irradiated material has been gaining acceptance throughout the world. This direct measurement of fracture toughness approach is technically superior to the correlative and indirect methods used in the past to assess irradiated RPV integrity. Several elements have been identified as focal points for Master Curve use: (i) limits of applicability for the Master Curve at the upper range of the transition region for loading quasi-static to dynamic/impact loading rates; (ii) effects of non-homogeneous material or changes due to environment conditions on the Master Curve, and how heterogeneity can be integrated into a more inclusive Master Curve methodology; (iii) importance of fracture mode differences and changes affect the Master Curve shape. The collected data in this report represent mostly results from non-irradiated testing, although some results from test reactor irradiations and plant surveillance programmes have been included as available. The results presented here should allow utility engineers and scientists to directly measure fracture toughness using small surveillance size specimens and apply the results using the Master Curve approach

Ductile fracture of cylindrical vessels containing a large flaw

Science.gov (United States)

Erdogan, F.; Irwin, G. R.; Ratwani, M.

1976-01-01

The fracture process in pressurized cylindrical vessels containing a relatively large flaw is considered. The flaw is assumed to be a part-through or through meridional crack. The flaw geometry, the yield behavior of the material, and the internal pressure are assumed to be such that in the neighborhood of the flaw the cylinder wall undergoes large-scale plastic deformations. Thus, the problem falls outside the range of applicability of conventional brittle fracture theories. To study the problem, plasticity considerations are introduced into the shell theory through the assumptions of fully-yielded net ligaments using a plastic strip model. Then a ductile fracture criterion is developed which is based on the concept of net ligament plastic instability. A limited verification is attempted by comparing the theoretical predictions with some existing experimental results.

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Effects of low upper shelf fracture toughness on reactor vessel integrity during pressurized thermal shock events

International Nuclear Information System (INIS)

Bamford, W.H.; Heinecke, C.C.; Balkey, K.R.

1988-01-01

For the past decade, significant attention has been focused on the subject of nuclear rector vessel integrity during pressurized thermal shock (PTS) events. The issue of low upper shelf fracture toughness at operating temperatures has been a consideration for some reactor vessel materials since the early 1970's. Deterministic and probabilistic fracture mechanics sensitivity studies have been completed to evaluate the interaction between the PTS and lower upper shelf toughness issues that result from neutron embrittlement of the critical beltline region materials. This paper presents the results of these studies to show the interdependency of these fracture considerations in certain instances and to identify parameters that need to be carefully treated in reactor vessel integrity evaluations for these subjects. This issue is of great importance to those vessels which have low upper shelf toughness, both for demonstrating safety during the original design life and in life extension assessments

The treatment of residual stress in fracture assessment of pressure vessels

International Nuclear Information System (INIS)

Green, D.; Knowles, J.

1992-01-01

The treatment of weld residual stress in the fracture assessment of cylindrical pressure vessels is considered through partitioning the stress into membrane, bending and self-balancing through wall components. The influence of each on fracture behavior is discussed. Stress intensity factor solutions appropriate to each type of stress are presented. Short range, medium range and long range stress categories are identified according to simple rules relating the effect of increasing crack length to stress intensity factor and ligament net stress. Proposals are made on how the stress intensity factor from these stress types may be incorporated into a Kr, Lr based fracture assessment

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

Fracture-mechanics data deduced from thermal-shock and related experiments with LWR pressure-vessel material

International Nuclear Information System (INIS)

Cheverton, R.D.; Canonico, D.A.; Iskander, S.K.; Bolt, S.E.; Holz, P.P.; Nanstad, R.K.; Stelzman, W.J.

1982-01-01

Pressurized water reactors (PWRs) are susceptible to certain types of hypothetical accidents that can subject the reactor pressure vessel to severe thermal shock, that is, a rapid cooling of the inner surface of the vessel wall. The thermal-shock loading, coupled with the radiation-induced reduction in the material fracture toughness, introduces the possibility of propagation of preexistent flaws and what at one time were regarded as somewhat unique fracture-oriented conditions. Several postulated reactor accidents have been analyzed to discover flaw behavior trends; seven intermediate-scale thermal-shock experiments with steel cylinders have been conducted; and corresponding materials characterization studies have been performed. Flaw behavior trends and related fracture-mechanics data deduced from these studies are discussed

Scatter modelling of fracture toughness data for reactor pressure vessel structural integrity assessment

International Nuclear Information System (INIS)

Pesoz, M.

1997-01-01

In the last decade, there has been an increasing interest at EDF in developing and applying probabilistic methods for a variety of purposes. In the field of structural integrity and reliability they are used to evaluate the effect of deterioration due to ageing mechanisms, mainly on major passive structural components such as reactor pressure vessel, steam generator and piping in nuclear plants. Such approaches provide an attractive supplement to the more conventional deterministic method, based upon pessimistic assumptions, that give results too far from reality to support effective decisions. In addition to deterministic calculations, a Probabilistic Fracture Mechanics model has been developed in order to analyse the risk of brittle failure of the reactor pressure vessel and to perform sensitivity studies. The material fracture toughness (K IC ) uncertainty appears to be strongly influencing the probability of failure under accidental conditions. Up to now, this parameter is determined from the RCC-M code reference curve, which is the same as the ASME reference curve. But an important issue when performing probabilistic analysis is the correct statistical modelling of input parameters. That's why modelling works have been carried out using results of fracture toughness tests performed for demonstrating the validity of the reference curve. This paper presents the statistical treatments that have been performed to model the scatter of temperature dependent parameter (K IC (T). A specific data base containing a few hundreds of French and US results have been carried and Weibull models have been fitted, based on various master curve equations (K. Wallin (Senior Adviser at the Technical Research Centre of Finland) or RCC-M types). (author)

Fracture mechanics of thin wall cylindrical pressure vessels: an interim review

International Nuclear Information System (INIS)

Kurtz, R.J.; Olson, N.J.

1977-08-01

The report is a result of activities in the LMFBR Fuel Rod Transient Performance Program sponsored by the LMFBR Branch of the Division of Project Management, U.S. Nuclear Regulatory Commission. One of the objectives is to develop predictions relative to the length, direction, and rate of growth of cladding rips subsequent to (or concurrent with) the initial cladding breach during unprotected transients. To provide a basis for evaluation, Battelle, Pacific Northwest Laboratories has reviewed most available fracture mechanics assessments relative to thin-wall cylindrical pressure vessels. The purpose of the report is to review the various fracture mechanics models and to describe the pertinent fracture parameters. It is intended to provide a formal basis for assessing future analytical predictions of fracture behavior of materials exposed to transient LMFBR thermal and mechanical loading conditions. In addition, the report is expected to provide reference material for evaluating or developing experimental programs required to properly address the problem of predicting fracture behavior of materials during transient events

Verification of the analytical fracture assessments methods by a large scale pressure vessel test

Energy Technology Data Exchange (ETDEWEB)

Keinanen, H; Oberg, T; Rintamaa, R; Wallin, K

1988-12-31

This document deals with the use of fracture mechanics for the assessment of reactor pressure vessel. Tests have been carried out to verify the analytical fracture assessment methods. The analysis is focused on flaw dimensions and the scatter band of material characteristics. Results are provided and are compared to experimental ones. (TEC).

Probabilistic Fracture Mechanics of Reactor Pressure Vessels with Populations of Flaws

Energy Technology Data Exchange (ETDEWEB)

Spencer, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Backman, Marie [Univ. of Tennessee, Knoxville, TN (United States); Williams, Paul [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoffman, William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dickson, Terry [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bass, B. Richard [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Klasky, Hilda [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-09-01

This report documents recent progress in developing a tool that uses the Grizzly and RAVEN codes to perform probabilistic fracture mechanics analyses of reactor pressure vessels in light water reactor nuclear power plants. The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. Because of the central role of the reactor pressure vessel (RPV) in a nuclear power plant, particular emphasis is being placed on developing capabilities to model fracture in embrittled RPVs to aid in the process surrounding decision making relating to life extension of existing plants. A typical RPV contains a large population of pre-existing flaws introduced during the manufacturing process. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation at one or more of these flaws during a transient event. This report documents development and initial testing of a capability to perform probabilistic fracture mechanics of large populations of flaws in RPVs using reduced order models to compute fracture parameters. The work documented here builds on prior efforts to perform probabilistic analyses of a single flaw with uncertain parameters, as well as earlier work to develop deterministic capabilities to model the thermo-mechanical response of the RPV under transient events, and compute fracture mechanics parameters at locations of pre-defined flaws. The capabilities developed as part of this work provide a foundation for future work, which will develop a platform that provides the flexibility needed to consider scenarios that cannot be addressed with the tools used in current practice.

Proposed rule package on fracture toughness and thermal annealing requirements and guidance for light water reactor vessels

International Nuclear Information System (INIS)

Allen Hiser, J.R.

1993-01-01

In the framework of updating and clarification of the fracture toughness and thermal annealing requirements and guidance for light water reactor pressure vessels, proposed revisions concerning the pressurized thermal shock rule, fracture toughness requirements and reactor vessel material surveillance program requirements, are described. A new rule concerning thermal annealing requirements and a draft regulatory guide on 'Format and Content of Application for Approval for Thermal Annealing of RPV' are also proposed

Proposed rule package on fracture toughness and thermal annealing requirements and guidance for light water reactor vessels

Energy Technology Data Exchange (ETDEWEB)

Allen Hiser, J R [UKAEA Harwell Lab. (United Kingdom). Engineering Div.

1994-12-31

In the framework of updating and clarification of the fracture toughness and thermal annealing requirements and guidance for light water reactor pressure vessels, proposed revisions concerning the pressurized thermal shock rule, fracture toughness requirements and reactor vessel material surveillance program requirements, are described. A new rule concerning thermal annealing requirements and a draft regulatory guide on `Format and Content of Application for Approval for Thermal Annealing of RPV` are also proposed.

Elastic-plastic fracture mechanics for nuclear pressure vessels: a preliminary appraisal

International Nuclear Information System (INIS)

Hahn, G.T.; Broek, D.; Marschall, C.W.; Rosenfield, A.R.; Rybicki, E.F.; Schmueser, D.W.; Stonesifer, R.B.; Kanninen, M.F.

1978-01-01

A research program directed at assessing the margin of safety of flawed nuclear pressure vessels near and beyond general yielding is described. The program has the general objective of developing an elastic-plastic fracture mechanics methodology. The approach is based on the use of finite element models together with experimental results to identify criteria appropriate for the onset of crack extension and for stable crack growth. A number of criteria beyond the conventional LEFM R curve are being evaluated. These include the critical values of the J-integral, its derivative, the crack tip opening angle, the average crack opening angle, a generalized energy release rate, its components and a crack tip force. The optimum fracture criterion for nuclear vessels is being determined by systematic measurements of load extension curves, strain distribution, crack opening displacement, stable crack growth and instability on 'toughness scaled' model materials. Computations have been performed for center cracked panels of a model material (2219-T87 aluminium) for full shear failure. (author)

Development of a shallow-flaw fracture assessment methodology for nuclear reactor pressure vessels

International Nuclear Information System (INIS)

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

1996-01-01

Shallow-flaw fracture technology is being developed within the Heavy-Section Steel Technology (HSST) Program for application to the safety assessment of radiation-embrittled nuclear reactor pressure vessels (RPVs) containing postulated shallow flaws. Cleavage fracture in shallow-flaw cruciform beam specimens tested under biaxial loading at temperatures in the lower transition temperature range was shown to be strain-controlled. A strain-based dual-parameter fracture toughness correlation was developed and shown to be capable of predicting the effect of crack-tip constraint on fracture toughness for strain-controlled fracture. A probabilistic fracture mechanics (PFM) model that includes both the properties of the inner-surface stainless-steel cladding and a biaxial shallow-flaw fracture toughness correlation gave a reduction in probability of cleavage initiation of more than two orders of magnitude from an ASME-based reference case

Revision of the fracture models in steels for nuclear pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Darwish, F A.I. [Pontificia Univ. Catolica do Rio de Janeiro (Brazil). Dept. de Ciencia dos Materiais e Metalurgia

1981-01-01

The variation of toughness with the temperature of steels used in the fabrication of nuclear pressure vessels is presented and discuted by mathematical models aiming to reach a critical value of stress or deformation at the moment of the fracture. The mathematical model considered are compatible with the fracture micromechanisms in action and they are capable of foreseeing the variations in the toughness from the mechanical properties evaluated in the tension test. The neutron irradiation effects in the toughness as well as in the variation of this toughness with the operating temperature are still described.

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.

Probabilistic fracture mechanics analysis of reactor vessel for pressurized thermal shock: the effect of residual stress and fracture toughness

International Nuclear Information System (INIS)

Jung, Sung Gyu; Jin, Tae Eun; Jhung, Myung Jo; Choi, Young Hwan

2003-01-01

The structural integrity of the reactor vessel with the approaching end of life must be assured for pressurized thermal shock. The regulation specifies the screening criteria for this and requires that specific analysis be performed for the reactor vessel which is anticipated to exceed the screening criteria at the end of plant life. In case the screening criteria is exceeded by the deterministic analysis, probabilistic analysis must be performed to show that failure probability is within the limit. In this study, probabilistic fracture mechanics analysis of the reactor vessel for pressurized thermal shock is performed and the effects of residual stress and master curve on the failure probability are investigated

Heavy-Section Steel Technology program fracture issues

International Nuclear Information System (INIS)

Pennell, W.E.

1989-10-01

Large scale fracture mechanics tests have resulted in the identification of a number of fracture technology issues. Identification of additional issues has come from the reactor vessel materials irradiation test program and from reactor operating experience. This paper provides a review of fracture issues with an emphasis on their potential impact on a reactor vessel pressurized thermal shock (PTS) analysis. Mixed mode crack propagation emerges as a major issue, due in large measure to the poor performance of existing models for the prediction of ductile tearing. Rectification of ductile tearing technology deficiencies may require extending the technology to include a more complete treatment of stress state and loading history effects. The effect of cladding on vessel fracture remains uncertain to the point that it is not possible to determine at this time if the net effect will be positive or negative. Enhanced fracture toughness for shallow flaws has been demonstrated for low strength structural steels. Demonstration of a similar effect in reactor pressure vessel steels could have a significant beneficial effect on the probabilistic analysis of reactor vessel fracture. Further development of existing fracture mechanics models and concepts is required to meet the special requirements for fracture evaluation of circumferential flaws in the welds of ring forged vessels. Fracture technology advances required to address the issues discussed in this paper are the major objective for the ongoing Heavy Section Steel Technology (HSST) program at ORNL. 24 refs., 18 figs

Heavy-section steel technology program: Fracture issues

International Nuclear Information System (INIS)

Pennell, W.E.

1992-01-01

Large-scale fracture mechanics tests have resulted in the identification of a number of fracture technology issues. Identification of additional issues has come from the reactor vessel materials irradiation test program and from reactor operating experience. This paper provides a review of fracture issues with an emphasis on their potential impact on a reactor vessel pressurized thermal shock (PTS) analysis. Mixed mode crack propagation emerges as a major issue, due in large measure to the poor performance of existing models for the prediction of ductile tearing. Rectification of ductile tearing technology deficiencies may require extending the technology to include a more complete treatment of stress state and loading history effects. The effect of cladding on vessel fracture remains uncertain to the point that it is not possible to determine at this time if the net effect will be positive or negative. Enhanced fracture toughness for shallow flaws has been demonstrated for low-strength structural steels. Demonstration of a similar effect in reactor pressure vessel steels could have a significant beneficial effect on the probabilistic analysis of reactor vessel fracture. Further development of existing fracture mechanics models and concepts is required to meet the special requirements for fracture evaluation of circumferential flaws in the welds of ring-forged vessels. Fracture technology advances required to address the issues discussed in this paper are the major objective for the ongoing Heavy Section Steel Technology (HSST) program at ORNL

A ductile fracture mechanics methodology for predicting pressure vessel and piping failure

International Nuclear Information System (INIS)

Landes, J.D.; Zhou, Z.

1991-01-01

This paper reports on a ductile fracture methodology based on one used more generally for the prediction of fracture behavior that was applied to the prediction of fracture behavior in pressure vessel and piping components. The model uses the load versus displacement record from a fracture toughness test to develop inputs for predicting the behavior of the structural component. The principle of load separation is used to convert the test record into two pieces of information, calibration functions which describe the structural deformation behavior and fracture toughness which describes the response of a crack-like flaw to the loading. These calibration functions and fracture toughness values which relate to the test specimen are then transformed to those appropriate to the structure. Often in this step computation procedures could be used but are not always necessary. The calibration functions and fracture for the structure are recombined to predict a load versus displacement behavior for the structure. The input for the model was generated from tests of compact specimen geometries; this geometry is often used for fracture toughness testing. The predictions were done for five model structures

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-08-15

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

Effect of high-temperature water and hydrogen on the fracture behavior of a low-alloy reactor pressure vessel steel

International Nuclear Information System (INIS)

Roychowdhury, S.; Seifert, H.-P.; Spätig, P.; Que, Z.

2016-01-01

Structural integrity of reactor pressure vessels (RPV) is critical for safety and lifetime. Possible degradation of fracture resistance of RPV steel due to exposure to coolant and hydrogen is a concern. In this study tensile and elastic-plastic fracture mechanics (EPFM) tests in air (hydrogen pre-charged) and EFPM tests in hydrogenated/oxygenated high-temperature water (HTW) was done, using a low-alloy RPV steel. 2–5 wppm hydrogen caused embrittlement in air tensile tests at room temperature (25 °C) and at 288 °C, effects being more significant at 25 °C and in simulated weld coarse grain heat affected zone material. Embrittlement at 288 °C is strain rate dependent and is due to localized plastic deformation. Hydrogen pre-charging/HTW exposure did not deteriorate the fracture resistance at 288 °C in base metal, for investigated loading rate range. Clear change in fracture morphology and deformation structures was observed, similar to that after air tests with hydrogen. - Highlights: • Hydrogen content, microstructure of LAS, and strain rate affects tensile properties at 288 °C. • Strength affects hydrogen embrittlement susceptibility to a greater extent than grain size. • Hydrogen in LAS leads to strain localization and restricts cross-slip at 288 °C. • Possible hydrogen pickup due to exposure to 288 °C water alters fracture surface appearance without affecting fracture toughness in bainitic base material. • Simulated weld heat affected zone microstructure shows unstable crack propagation in 288 °C water.

Effect of high-temperature water and hydrogen on the fracture behavior of a low-alloy reactor pressure vessel steel

Energy Technology Data Exchange (ETDEWEB)

Roychowdhury, S., E-mail: sroy27@gmail.com [Paul Scherrer Institut, Nuclear Energy and Safety Research Department, Laboratory for Nuclear Materials, 5232 Villigen, PSI (Switzerland); Materials Processing & Corrosion Engineering Division, Mod-Lab, D-Block, Bhabha Atomic Research Centre, Mumbai 400085 (India); Seifert, H.-P.; Spätig, P.; Que, Z. [Paul Scherrer Institut, Nuclear Energy and Safety Research Department, Laboratory for Nuclear Materials, 5232 Villigen, PSI (Switzerland)

2016-09-15

Structural integrity of reactor pressure vessels (RPV) is critical for safety and lifetime. Possible degradation of fracture resistance of RPV steel due to exposure to coolant and hydrogen is a concern. In this study tensile and elastic-plastic fracture mechanics (EPFM) tests in air (hydrogen pre-charged) and EFPM tests in hydrogenated/oxygenated high-temperature water (HTW) was done, using a low-alloy RPV steel. 2–5 wppm hydrogen caused embrittlement in air tensile tests at room temperature (25 °C) and at 288 °C, effects being more significant at 25 °C and in simulated weld coarse grain heat affected zone material. Embrittlement at 288 °C is strain rate dependent and is due to localized plastic deformation. Hydrogen pre-charging/HTW exposure did not deteriorate the fracture resistance at 288 °C in base metal, for investigated loading rate range. Clear change in fracture morphology and deformation structures was observed, similar to that after air tests with hydrogen. - Highlights: • Hydrogen content, microstructure of LAS, and strain rate affects tensile properties at 288 °C. • Strength affects hydrogen embrittlement susceptibility to a greater extent than grain size. • Hydrogen in LAS leads to strain localization and restricts cross-slip at 288 °C. • Possible hydrogen pickup due to exposure to 288 °C water alters fracture surface appearance without affecting fracture toughness in bainitic base material. • Simulated weld heat affected zone microstructure shows unstable crack propagation in 288 °C water.

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

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

International Nuclear Information System (INIS)

Francisco, Alexandre S.; Duran, Jorge Alberto R.

2013-01-01

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

Probabilistic application of fracture mechanics

International Nuclear Information System (INIS)

Dufresne, J.

1981-04-01

The different methods used to evaluate the rupture probability of a pressure vessel are reviewed. Data collection and processing of all parameters necessary for fracture mechanics evaluation are presented with particular attention to the size distribution of defects in actual vessels. Physical process is followed during crack growth and unstable propagation, using LEFM (Linear Elastic Fracture Mechanism) and plastic instability. Results show that the final failure probability for a PWR pressure vessel is 3.5 10 -8 , and is due essentially to LOCAs for any break size. The weakest point is the internal side of the belt line

Uncertainty Characterization of Reactor Vessel Fracture Toughness

International Nuclear Information System (INIS)

Li, Fei; Modarres, Mohammad

2002-01-01

To perform fracture mechanics analysis of reactor vessel, fracture toughness (K Ic ) at various temperatures would be necessary. In a best estimate approach, K Ic uncertainties resulting from both lack of sufficient knowledge and randomness in some of the variables of K Ic must be characterized. Although it may be argued that there is only one type of uncertainty, which is lack of perfect knowledge about the subject under study, as a matter of practice K Ic uncertainties can be divided into two types: aleatory and epistemic. Aleatory uncertainty is related to uncertainty that is very difficult to reduce, if not impossible; epistemic uncertainty, on the other hand, can be practically reduced. Distinction between aleatory and epistemic uncertainties facilitates decision-making under uncertainty and allows for proper propagation of uncertainties in the computation process. Typically, epistemic uncertainties representing, for example, parameters of a model are sampled (to generate a 'snapshot', single-value of the parameters), but the totality of aleatory uncertainties is carried through the calculation as available. In this paper a description of an approach to account for these two types of uncertainties associated with K Ic has been provided. (authors)

Fracture toughness behavior and its analysis on nuclear pressure vessel steels

Energy Technology Data Exchange (ETDEWEB)

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

1983-02-01

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

Dynamic fracture characterization of a pressure vessel steel

International Nuclear Information System (INIS)

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

1991-01-01

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

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Habil; Ziliukas, A.

1997-04-01

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

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

International Nuclear Information System (INIS)

Bass, B.R.; Bryan, R.H.; Bryson, J.W.; Merkle, J.G.

1983-01-01

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

Biaxial loading effects on fracture toughness of reactor pressure vessel steel

International Nuclear Information System (INIS)

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

1995-03-01

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

Proceedings of the 1985 pressure vessels and piping conference. Volume PVP-98-8. Fracture, fatigue and advanced mechanics

International Nuclear Information System (INIS)

Short, W.E.; Zamrik, S.Y.

1985-01-01

State-of-the-art engineering practices in pressure vessel and piping technology are the result of continual efforts in the evaluation of problems which have been experienced and the development of appropriate design and analysis methods for those applications. The resulting advances in technology benefit industry with properly engineered, safe, cost-effective pressure vessels and piping systems. To this end, advanced study continues in specialized areas of mechanical engineering such as fracture mechanics, experimental stress analysis, high pressure applications and related material considerations, as well as advanced techniques for evaluation of commonly encountered design problems. This volume is comprised of current technical papers on various aspects of fracture, fatigue and advanced mechanics as related to the design and analysis of pressure vessels and piping

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

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

International Nuclear Information System (INIS)

Potirniche, Gabriel; Barlow, Fred D.; Charit, Indrajit; Rink, Karl

2013-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-11-26

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

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

International Nuclear Information System (INIS)

Bass, B.R.; Bryan, R.H.; Bryson, J.W.; Merkle, J.G.

1985-01-01

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

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

International Nuclear Information System (INIS)

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

1979-01-01

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

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

The method of life extension for the High Flux Isotope Reactor vessel

International Nuclear Information System (INIS)

Chang, Shib-Jung.

1995-01-01

The state of the vessel steel embrittlement as a result of neutron irradiation can be measured by its increase in the nil ductility temperature (NDT). This temperature is sometimes referred to as the brittle-ductile transition temperature (DBT) for fracture. The life extension of the High Flux Isotope Reactor (HFIR) vessel is calculated by using the method of fracture mechanics. A hydrostatic pressure test (hydrotest) is performed in order to determine a safe vessel static pressure. It is then followed by using fracture mechanics to project the reactor life from the safe hydrostatic pressure. The life extension calculation provides the following information on the remaining life of the reactor as a function of the nil ductility temperature increase: the probability of vessel fracture due to hydrotest vs vessel life at several hydrotest pressures; the hydrotest time interval vs the uncertainty of the nil ductility temperature increase rate; and the hydrotest pressure vs the uncertainty of the nil ductility temperature increase rate. It is understood that the use of a complete range of uncertainties of the nil ductility temperature increase is equivalent to the entire range of radiation damage that can be experienced by the vessel steel. From the numerical values for the probabilities of the vessel fracture as a result of hydrotest, it is estimated that the reactor vessel life can be extended up to 50 EFPY (100 MW) with the minimum vessel operating temperature equal to 85 degree F

Spinal compression fractures due to pregnancy-associated osteoporosis

Directory of Open Access Journals (Sweden)

R Krishnakumar

2016-01-01

Conclusion: Vertebral fractures due to PAO should be considered as a differential diagnosis in patients with back pain who are in the third trimester of pregnancy or in postpartum. Early recognition and appropriate conservative management would be necessary to prevent complications such as new vertebral fractures and chronic back pain.

Reactor pressure vessel structural integrity research

International Nuclear Information System (INIS)

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

1994-01-01

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

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

International Nuclear Information System (INIS)

Yagawa, Genki; Yoshimura, Shinobu

1997-01-01

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

Ductile fracture estimation of reactor pressure vessel under thermal shock

International Nuclear Information System (INIS)

Takahashi, Jun; Sakai, Shinsuke; Okamura, Hiroyuki

1990-01-01

This paper presents a new scheme for the estimation of unstable ductile fracture of a reactor pressure vessel under thermal shock conditions. First, it is shown that the bending moment applied to the cracked section can be evaluated by considering the plastic deformation of the cracked section and the thermal deformation of the shell. As the contribution of the local thermal stress to the J-value is negligible, the J-value under thermal shock can be easily evaluated by using fully plastic solutions for the cracked part. Next, the phenomena of ductile fracture under thermal shock are expressed on the load-versus-displacement diagram which enables us to grasp the transient phenomena visually. In addition, several parametrical surveys are performed on the above diagram concerning the variation of (1) thermal shock conditions, (2) initial crack length, and (3) J-resistance curve (i.e. embrittlement by neutron irradiation). (author)

Applicability of the fracture toughness master curve to irradiated reactor pressure vessel steels

International Nuclear Information System (INIS)

Sokolov, M.A.; McCabe, D.E.; Alexander, D.J.; Nanstad, R.K.

1997-01-01

The current methodology for determination of fracture toughness of irradiated reactor pressure vessel (RPV) steels is based on the upward temperature shift of the American Society of Mechanical Engineers (ASME) K Ic curve from either measurement of Charpy impact surveillance specimens or predictive calculations based on a database of Charpy impact tests from RPV surveillance programs. Currently, the provisions for determination of the upward temperature shift of the curve due to irradiation are based on the Charpy V-notch (CVN) 41-J shift, and the shape of the fracture toughness curve is assumed to not change as a consequence or irradiation. The ASME curve is a function of test temperature (T) normalized to a reference nit-ductility temperature, RT NDT , namely, T-RT NDT . That curve was constructed as the lower boundary to the available K Ic database and, therefore, does not consider probability matters. Moreover, to achieve valid fracture toughness data in the temperature range where the rate of fracture toughness increase with temperature is rapidly increasing, very large test specimens were needed to maintain plain-strain, linear-elastic conditions. Such large specimens are impractical for fracture toughness testing of each RPV steel, but the evolution of elastic-plastic fracture mechanics has led to the use of relatively small test specimens to achieve acceptable cleavage fracture toughness measurements, K Jc , in the transition temperature range. Accompanying this evolution is the employment of the Weibull distribution function to model the scatter of fracture toughness values in the transition range. Thus, a probabilistic-based bound for a given data population can be made. Further, it has been demonstrated by Wallin that the probabilistic-based estimates of median fracture toughness of ferritic steels tend to form transition curves of the same shape, the so-called ''master curve'', normalized to one common specimen size, namely the 1T [i.e., 1.0-in

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

Influence of steel-making process and heat-treatment temperature on the fatigue and fracture properties of pressure vessel steels

International Nuclear Information System (INIS)

Koh, S. K.; Na, E. G.; Baek, T. H.; Won, S. Y.; Park, S. J.; Lee, S. W.

2001-01-01

In this paper, high strength pressure vessel steels having the same chemical compositions were manufactured by the two different steel-making processes, such as Vacuum Degassing(VD) and Electro-Slag Remelting(ESR) methods. After the steel-making process, they were normalized at 955 deg. C, quenched at 843 .deg. C, and finally tempered at 550 .deg. C or 450 deg. C, resulting in tempered martensitic microstructures with different yielding strengths depending on the tempering conditions. Low-Cycle Fatigue(LCF) tests, Fatigue Crack Growth Rate(FCGR) tests, and fracture toughness tests were performed to investigate the fatigue and fracture behaviors of the pressure vessel steels. In contrast to very similar monotonic, LCF, and FCGR behaviors between VD and ESR steels, a quite difference was noticed in the fracture toughness. Fracture toughness of ESR steel was higher than that of VD steel, being attributed to the removal of impurities in steel-making process

Probabilistic study of PWR reactor pressure vessel fracture

International Nuclear Information System (INIS)

Dufresne, J.; Lucia, A.C.; Grandemange, J.; Pellissier-Tanon, A.

1983-01-01

Different methods are used to evaluate the rupture probability of a nuclear pressure vessel. On of them extrapolates to nuclear pressure vessels, data of failure found in conventional pressure vessels. The disadvantage of such an approach is that the effects of systematic changes in key parameters cannot be taken into account. For example, the influence of irradiation and the use of quality assurance programs encompassing design, fabrication and materials cannot be considered. But the most important disadvantage of this method is the limited size of the representative population and consequently the high value of the upper bound failure rate corresponding to a requested confidence level. The method used in the present work involves the development of physical models based on an understanding of the failure modes and expressing the conventional concepts of fracture mechanics in a probabilistic form; the fatigue crack growth rate, calculated for conditions of cyclic loading, the initiation of unstable crack propagation, and the possibility of crack arrest. The analysis therefore requires the statistical expression of the factors and parameters which appear in the expressions of the law of crack growth and of toughness, and also those which are used in the calculation of the stress intensity factor K 1 . All input data are entered in COVASTOL code in histogram form. This code takes into account the degree of correlation between the flaw size and the Paris' law coefficients. It computes the propagation of a given defect in a given position, and the corresponding failure probability during accidental loading

Reactor pressure vessel structural integrity research in the US Nuclear Regulatory Commission HSST and HSSI Programs

International Nuclear Information System (INIS)

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

1994-01-01

This report discusses development on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels containing flaws. Fracture mechanics tests on reactor pressure vessel steel have shown that local brittle zones do not significantly degrade the material fracture toughness, constraint relaxation at the crack tip of shallow surface flaws results in increased fracture toughness, and biaxial loading reduces but does not eliminate the shallow-flaw fracture toughness elevation. Experimental irradiation investigations have shown that the irradiation-induced shift in Charpy V-notch versus temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement and the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties

Critical cleavage fracture stress characterization of A508 nuclear pressure vessel steels

International Nuclear Information System (INIS)

Wu, Sujun; Jin, Huijin; Sun, Yanbin; Cao, Luowei

2014-01-01

The critical cleavage fracture stress of SA508 Gr.4N and SA508 Gr.3 low alloy reactor pressure vessel (RPV) steels was studied through the combination of experiments and finite element method (FEM) analysis. The results showed that the value of the local cleavage fracture stress, σ F , of SA508 Gr.4N steel was significantly higher than that of SA508 Gr.3 steel. Detailed microstructural analysis was carried out using FEGSEM which revealed much smaller grains, finer and more homogenous carbide particles formed in SA508 Gr.4N steel. Compared with the SA508 Gr.3 steel currently used in the nuclear industry, the SA508 Gr.4N steel possesses higher strength and notch toughness as well as improved cleavage fracture behavior, and is considered a better candidate RPV steel for the next generation nuclear reactors. - Highlights: • Critical cleavage fracture stress was calculated through experiments and FEM. • Effects of both grain and carbide particle sizes on σ F were discussed. • The SA508 Gr.4N steel is a better candidate for the next generation nuclear reactors

Comparison of ASME pressure–temperature limits on the fracture probability for a pressurized water reactor pressure vessel

International Nuclear Information System (INIS)

Chou, Hsoung-Wei; Huang, Chin-Cheng

2017-01-01

Highlights: • P-T limits based on ASME K_I_a curve, K_I_C curve and RI method are presented. • Probabilistic and deterministic methods are used to evaluate P-T limits on RPV. • The feasibility of substituting P-T curves with more operational is demonstrated. • Warm-prestressing effect is critical in determining the fracture probability. - Abstract: The ASME Code Section XI-Appendix G defines the normal reactor startup (heat-up) and shut-down (cool-down) operation limits according to the fracture toughness requirement of reactor pressure vessel (RPV) materials. This paper investigates the effects of different pressure-temperature limit operations on structural integrity of a Taiwan domestic pressurized water reactor (PWR) pressure vessel. Three kinds of pressure-temperature limits based on different fracture toughness requirements – the K_I_a fracture toughness curve of ASME Section XI-Appendix G before 1998 editions, the K_I_C fracture toughness curve of ASME Section XI-Appendix G after 2001 editions, and the risk-informed revision method supplemented in ASME Section XI-Appendix G after 2013 editions, respectively, are established as the loading conditions. A series of probabilistic fracture mechanics analyses for the RPV are conducted employing ORNL’s FAVOR code considering various radiation embrittlement levels under these pressure-temperature limit conditions. It is found that the pressure-temperature operation limits which provide more operational flexibility may lead to higher fracture risks to the RPV. The cladding-induced shallow surface breaking flaws are the most critical and dominate the fracture probability of the RPV under pressure-temperature limit transients. Present study provides a risk-informed reference for the operation safety and regulation viewpoint of PWRs in Taiwan.

Cracking at nozzle corners in the nuclear pressure vessel industry

International Nuclear Information System (INIS)

Smith, C.W.

1986-01-01

Cracks in nozzle corners at the pressure boundary of nuclear reactors have been frequently observed in service. These cracks tend to form with radial orientations with respect to the nozzle central axis and are believed to be initiated by thermal shock. However, their growth is believed to be primarily due to a steady plus a fluctuating internal pressure. Due to the impracticality of fracture testing of full-scale models, the Oak Ridge National Laboratory instituted the use of an intermediate test vessel (ITV) for use in fracture testing which had the same wall thickness and nozzle size as the prototype but significantly reduced overall length and diameter. In order to determine whether or not these ITVs could provide realistic data for full-scale reactor vessels, laboratory models of full-scale boiling water reactors and ITVs were constructed and tested. After briefly reviewing the laboratory testing and correlating results with service experience, results obtained will be used to draw some general conclusions regarding the stable growth of nonplanar cracks with curved crack fronts which are the most common precursors to fracture of pressure vessel components near junctures. Use of linear elastic fracture mechanics is made in determining stress-intensity distribution along the crack fronts

FRACTURE MECHANICS UNCERTAINTY ANALYSIS IN THE RELIABILITY ASSESSMENT OF THE REACTOR PRESSURE VESSEL: (2D SUBJECTED TO INTERNAL PRESSURE

Directory of Open Access Journals (Sweden)

Entin Hartini

2016-06-01

Full Text Available ABSTRACT FRACTURE MECHANICS UNCERTAINTY ANALYSIS IN THE RELIABILITY ASSESSMENT OF THE REACTOR PRESSURE VESSEL: (2D SUBJECTED TO INTERNAL PRESSURE. The reactor pressure vessel (RPV is a pressure boundary in the PWR type reactor which serves to confine radioactive material during chain reaction process. The integrity of the RPV must be guaranteed either  in a normal operation or accident conditions. In analyzing the integrity of RPV, especially related to the crack behavior which can introduce break to the reactor pressure vessel, a fracture mechanic approach should be taken for this assessment. The uncertainty of input used in the assessment, such as mechanical properties and physical environment, becomes a reason that the assessment is not sufficient if it is perfomed only by deterministic approach. Therefore, the uncertainty approach should be applied. The aim of this study is to analize the uncertainty of fracture mechanics calculations in evaluating the reliability of PWR`s reactor pressure vessel. Random character of input quantity was generated using probabilistic principles and theories. Fracture mechanics analysis is solved by Finite Element Method (FEM with  MSC MARC software, while uncertainty input analysis is done based on probability density function with Latin Hypercube Sampling (LHS using python script. The output of MSC MARC is a J-integral value, which is converted into stress intensity factor for evaluating the reliability of RPV’s 2D. From the result of the calculation, it can be concluded that the SIF from  probabilistic method, reached the limit value of  fracture toughness earlier than SIF from  deterministic method.  The SIF generated by the probabilistic method is 105.240 MPa m0.5. Meanwhile, the SIF generated by deterministic method is 100.876 MPa m0.5. Keywords: Uncertainty analysis, fracture mechanics, LHS, FEM, reactor pressure vessels   ABSTRAK ANALISIS KETIDAKPASTIAN FRACTURE MECHANIC PADA EVALUASI KEANDALAN

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

Structural failure analysis of reactor vessels due to molten core debris

International Nuclear Information System (INIS)

Pfeiffer, P.A.

1993-01-01

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

[Epidemiology of maxillofacial fractures due to traffic accidents in Medellin (Colombia)].

Science.gov (United States)

Agudelo-Suárez, Andrés A; Duque-Serna, Francisco Levi; Restrepo-Molina, Lucas; Martínez-Herrera, Eliana

2015-09-01

To characterize maxillofacial fractures due to traffic accidents in patients attending the Hospital Universitario San Vicente Fundación (Medellin-Colombia) from 1998 to 2010. A descriptive study (n =1609) was carried out with information from the medical records of patients meeting the inclusion criteria established by the general objective of the study. The variables consisted of sex, age, year, type and number of fractures, and type of vehicle. A descriptive analysis of the variables was performed and the frequency of fractures due to traffic accidents was calculated according to year and sex. Crude and adjusted odds ratios (aOR) were estimated to establish associations among age, type of vehicle, and the presence of two or more fractures with stratification by sex. The frequency of maxillofacial fractures due to traffic accidents increased in 2007 (men: n=198, women: n=35) and decreased from 2008 to 2010 in both sexes. Fractures were more frequent in persons aged <35 years (80%) and in men (82%). The highest frequency of fractures was observed in motorists. Male users of motorcycles (aOR=1.41; confidence interval 95% [95%CI]: 1.02- 1.94) and bicycles (aOR=1.61; 95%CI: 1.01- 2.56) were more likely to report two or more fractures compared with pedestrians, after adjustment for other variables. Most maxillofacial fractures occurred in men and in motorists. Future studies should analyze other determinants affecting the epidemiology of maxillofacial fractures. Strategies should be designed to improve the use of protective elements and drivers' knowledge and practices. Copyright © 2014 SESPAS. Published by Elsevier Espana. All rights reserved.

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

Energy Technology Data Exchange (ETDEWEB)

Grounes, M

1966-03-15

After a short review of the part of the Swedish nuclear energy program that is of interest in this context the Swedish reactor pressure vessels and the reasoning behind the choice of materials are surveyed. Problems and desirable aims for future reactors are discussed. Much work is now being done on new types of pressure vessel steels with high strength, low transition temperature and good corrosion resistance. These steels are of the martensitic austenitic type Bofors 2RMO (13 % Cr, 6 % Ni, 1. 5 % Mo) and of the ferritic martensitic austenitic type Avesta 248 SV (16 % Cr, 5 % Ni, 1 % Mo). An applied philosophy for estimating the brittle-fracture tendency of pressure vessels is described. As a criterion of this tendency we use the crack-propagation transition temperature, e. g. as measured by the Robertson isothermal crack-arrest test. An estimate of this transition temperature at the end of the reactor' s lifetime must take increases due to fabrication, welding, geometry, ageing and irradiation into account. The transition temperature vs. stress curve moves towards higher temperatures during the reactor' s lifetime. As long as this curve does not cross the reactor vessel stress vs. temperature curve the vessel is considered safe. The magnitude of the different factors influencing the final transition temperature are discussed and data for the Marviken reactor's pressure vessel are presented. At the end of the reactor's lifetime the estimated transition temperature is 115 deg C, which is below the maximum permissible value. A program for the study of strain ageing has been initiated owing to the uncertainty as to the extent of strain ageing at low strains. A study of a simple crack-arrest test, developed in Sweden, is in progress. An extensive irradiation-effects program on several steels is in progress. Results from tests on the Swedish carbon-manganese steels 2103/R3, SIS 142103 and SIS 142102, the low-alloy steels Degerfors DE-631A, Bofors NO 345 and Fortiweld

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

International Nuclear Information System (INIS)

Grounes, M.

1966-03-01

After a short review of the part of the Swedish nuclear energy program that is of interest in this context the Swedish reactor pressure vessels and the reasoning behind the choice of materials are surveyed. Problems and desirable aims for future reactors are discussed. Much work is now being done on new types of pressure vessel steels with high strength, low transition temperature and good corrosion resistance. These steels are of the martensitic austenitic type Bofors 2RMO (13 % Cr, 6 % Ni, 1. 5 % Mo) and of the ferritic martensitic austenitic type Avesta 248 SV (16 % Cr, 5 % Ni, 1 % Mo). An applied philosophy for estimating the brittle-fracture tendency of pressure vessels is described. As a criterion of this tendency we use the crack-propagation transition temperature, e. g. as measured by the Robertson isothermal crack-arrest test. An estimate of this transition temperature at the end of the reactor' s lifetime must take increases due to fabrication, welding, geometry, ageing and irradiation into account. The transition temperature vs. stress curve moves towards higher temperatures during the reactor' s lifetime. As long as this curve does not cross the reactor vessel stress vs. temperature curve the vessel is considered safe. The magnitude of the different factors influencing the final transition temperature are discussed and data for the Marviken reactor's pressure vessel are presented. At the end of the reactor's lifetime the estimated transition temperature is 115 deg C, which is below the maximum permissible value. A program for the study of strain ageing has been initiated owing to the uncertainty as to the extent of strain ageing at low strains. A study of a simple crack-arrest test, developed in Sweden, is in progress. An extensive irradiation-effects program on several steels is in progress. Results from tests on the Swedish carbon-manganese steels 2103/R3, SIS 142103 and SIS 142102, the low-alloy steels Degerfors DE-631A, Bofors NO 345 and Fortiweld

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-01

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

OCA-P, a deterministic and probabilistic fracture-mechanics code for application to pressure vessels

International Nuclear Information System (INIS)

Cheverton, R.D.; Ball, D.G.

1984-05-01

The OCA-P code is a probabilistic fracture-mechanics code that was prepared specifically for evaluating the integrity of pressurized-water reactor vessels when subjected to overcooling-accident loading conditions. The code has two-dimensional- and some three-dimensional-flaw capability; it is based on linear-elastic fracture mechanics; and it can treat cladding as a discrete region. Both deterministic and probabilistic analyses can be performed. For the former analysis, it is possible to conduct a search for critical values of the fluence and the nil-ductility reference temperature corresponding to incipient initiation of the initial flaw. The probabilistic portion of OCA-P is based on Monte Carlo techniques, and simulated parameters include fluence, flaw depth, fracture toughness, nil-ductility reference temperature, and concentrations of copper, nickel, and phosphorous. Plotting capabilities include the construction of critical-crack-depth diagrams (deterministic analysis) and various histograms (probabilistic analysis)

Laryngeal fracture due to blunt trauma presenting with pneumothorax and pneumomediastinum.

Science.gov (United States)

Narcı, Adnan; Embleton, Didem Baskın; Ayçiçek, Abdullah; Yücedağ, Fatih; Cetinkurşun, Salih

2011-01-01

Injuries due to traffic accidents are frequent in childhood, and they have high mortality and morbidity. Laryngeal injury due to a traffic accident is a rare pathology and might be missed if not suspected. Here we present a laryngeal fracture in a child after a blunt chest trauma during a traffic accident that presented with pneumomediastinum and pneumothorax. A 14-year-old girl was referred for pneumomediastinum. Her physical examination was normal except subcutaneous emphysema, edema and tenderness in the cervical area, hoarseness, facial and extremity abrasions and ecchymoses. Chest tomography revealed pneumothorax and pneumomediastinum, and cranial tomography revealed maxillofacial fractures. Upper airway damage was suspected, flexible endoscopy revealed right vocal cord paralysis and cervical tomography revealed thyroid cartilage fracture. The fracture was repaired and tracheotomy was performed. She was discharged on postoperative day 6. Facial fractures were repaired in another center. Tracheotomy was removed on postoperative day 20. Her hoarseness, although decreased, still persists. Pneumomediastinum is a rare result of a laryngeal fracture and if not suspected, the fracture can easily be missed. It should be kept in mind after blunt cervical trauma with pneumomediastinum and/or pneumothorax. Direct endoscopy and cervical tomography may be necessary for the differential diagnosis. Copyright © 2011 S. Karger AG, Basel.

Fracture probability evaluation of a LWR pressure vessel

International Nuclear Information System (INIS)

Grandemange, J.; Pellissier-Tanon, A.; Quero, J.; Carnino, A.; Dufresne, J.

1978-01-01

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

Fracture Analysis of CNG High Pressure Container using Fractography and Measurement of Property

Directory of Open Access Journals (Sweden)

Kim Eui-Soo

2017-01-01

Full Text Available Bursting accidents of pressure containers due to design and manufacturing defects are frequently occurring. Due to high-pressure gas or harmful substances, when this vessel is fractured, it can lead to catastrophic disasters. Especially, in the event of bursting accident of composite pressure vessel for CNG bus, many unspecified people can be damaged. Most of the accidents were caused by problems in the manufacturing process. The manufacturing process for TYPE2 pressure vessel is very complicated such as three drawing processes, two ironing processes and one spinning process. In the middle of process, various heat treatments are performed for imparting toughness and removing residual stresses. It should cause a serious problem such as bursting and fragmentation of the pressure container due to defects of this process. In this research, the fracture cause of CNG vessel is evaluated through fractography and measuring material property using IIT and analysis of chemical composition.

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

International Nuclear Information System (INIS)

Rao, S.

1980-09-01

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

Fractures due to insufficient pelvic girdle

International Nuclear Information System (INIS)

Garcia Aguayo, F.J.; Martinez Almagro, A.

1995-01-01

Eleven cases are presented of postmenopausal women with a total of 37 fractures due to insufficient pelvic girdle: 15 located in sacrum, ten in the pubic rami, four in ilium proximal to the sacroiliac joint, three in iliac fossa, two in iliac tuberosity and three in the public body. Eight of the patients were diagnosed over a period of six years when seeking medical attention for bone pain. The other three were diagnosed retrospectively among a group of 33 cancer patients (the majority having having breast cancer) who presented positive pelvic radionuclide bone scan. CT was superior to conventional radiology in detecting fractures of this type, especially those of sacrum and ilium. Radionuclide bone scan was highly sensitive but its specificity was low, requiring back-up radiology and above all CT to establish the differential diagnosis with respect to other types of lesions, especially metastases. (Author) 14 refs

Development of Deterministic and Probabilistic Fracture Mechanics Analysis Code PROFAS-RV for Reactor Pressure Vessel - Progress of the Work

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong Min; Lee, Bong Sang [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In this study, a deterministic/probabilistic fracture mechanics analysis program for reactor pressure vessel, PROFAS-RV, is developed. This program can evaluate failure probability of RPV using recent radiation embrittlement model of 10CFR50.61a and stress intensity factor calculation method of RCC-MRx code as well as the required basic functions of PFM program. Applications of some new radiation embrittlement model, material database, calculation method of stress intensity factors, and others which can improve fracture mechanics assessment of RPV are introduced. The purpose of this study is to develop a probabilistic fracture mechanics (PFM) analysis program for RPV considering above modification and application of newly developed models and calculation methods. In this paper, it deals with the development progress of the PFM analysis program for RPV, PROFAS-RV. The PROFAS-RV is being tested with other codes, and it is expected to revise and upgrade by reflecting the latest model and calculation method continuously. These efforts can minimize the uncertainty of the integrity evaluation for the reactor pressure vessel.

Triple –E Vessels: Tonnage Measurement and Suez Canal Dues Assessment

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Elsayed Hussein Galall

2015-08-01

Full Text Available Container is growing faster than GDP, Shipping lines always attempt to augment efficiencyby reducing cost and by attracting larger volumes of containers. As a result rising containerfreight rates the lines have been driven to increase economic of scale, by building mega shipsand fewer mere efficient port calls. In 2011 Maersk line ordered up to 20 new “Triple- E “Class of container vessels deliversbetween 2013- 2015. These class of mega container vessels have its way through Suez Canal,other companies CMA, CGM also ordered this type of mega container vessels, in order toreach higher profits due to the achieved economics of scale It is believed that 20000 TEUcould be the next target size. Present mega container fleet and any future feasible potential vessel capacity expansionmore than 18000 TEU put Suez Canal route in strong competitive position. MeanwhilePanama Canal will not be able to handle vessels larger than 12600 TEU even after itsexpansion in 2015.

Analysis of size effect applicable to evaluation of fracture toughness of base metal for PWR vessel

International Nuclear Information System (INIS)

Benhamou, C.; Joly, P.; Andrieu, A.; Parrot, A.; Vidard, S.

2015-01-01

The objective of the present paper is to review the specimen size effect (also called crack front length effect) on Fracture Toughness of PWR Reactor Pressure Vessel Steel base metal. The analysis of the reality and amplitude of this effect is conducted in a first step on a database (the so-called GKSS database) including fracture toughness test results on a single representative material using specimens of different thicknesses, tested in the same temperature range. A realistic analytical form for describing the size effect observed in this data set is thus derived from statistical analyses and proposed for engineering application. In a second step, this size effect formulation is then applied to a large number of fracture toughness data, obtained in Irradiation Surveillance Programs, and also to the numerous data used for the definition of the ASME (and RCC-M) fracture toughness reference curves. This analysis allows normalizing all the available fracture toughness data with a single specimen width of 100 mm and defining the fracture toughness reference curve as the lower bound of this normalized set of data points. It is thus demonstrated that the fracture toughness reference curve is associated with a reference crack length of 100 mm, and can be used in RPV integrity analyses for other crack front length in association with the crack front length correction formula defined in the first step. (authors)

Residual stresses due to weld repairs, cladding and electron beam welds and effect of residual stresses on fracture behavior. Annual report, September 1, 1977--November 30, 1978

International Nuclear Information System (INIS)

Rybicki, E.F.

1978-11-01

The study is divided into three tasks. Task I is concerned with predicting and understanding the effects of residual stresses due to weld repairs of pressure vessels. Task II examines residual stresses due to an electron beam weld. Task III addresses the problem of residual stresses produced by weld cladding at a nozzle vessel intersection. The objective of Task I is to develop a computational model for predicting residual stress states due to a weld repair of pressure vessel and thereby gain an understanding of the mechanisms involved in the creation of the residual stresses. Experimental data from the Heavy Section Steel Technology (HSST) program at Oak Ridge National Laboratories (ORNL) is used to validate the computational model. In Task II, the residual stress model is applied to the case of an electron beam weld of a compact tension freacture specimen. The results in the form of residual stresses near the weld are then used to explain unexpected fracture behavior which is observed in the testing of the specimen. For Task III, the residual stress model is applied to the cladding process used in nozzle regions of nuclear pressure vessels. The residual stresses obtained from this analysis are evaluated to determine their effect on the phenomena of under-clad cracking

User's manual and analysis methodology of probabilistic fracture mechanics analysis code PASCAL Ver.2 for reactor pressure vessel (Contract research)

International Nuclear Information System (INIS)

Osakabe, Kazuya; Onizawa, Kunio; Shibata, Katsuyuki; Kato, Daisuke

2006-09-01

As a part of the aging structural integrity research for LWR components, the probabilistic fracture mechanics (PFM) analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed in JAEA. This code evaluates the conditional probabilities of crack initiation and fracture of a reactor pressure vessel (RPV) under transient conditions such as pressurized thermal shock (PTS). The development of the code has been aimed to improve the accuracy and reliability of analysis by introducing new analysis methodologies and algorithms considering the recent development in the fracture mechanics and computer performance. PASCAL Ver.1 has functions of optimized sampling in the stratified Monte Carlo simulation, elastic-plastic fracture criterion of the R6 method, crack growth analysis models for a semi-elliptical crack, recovery of fracture toughness due to thermal annealing and so on. Since then, under the contract between the Ministry of Economy, Trading and Industry of Japan and JAEA, we have continued to develop and introduce new functions into PASCAL Ver.2 such as the evaluation method for an embedded crack, K I database for a semi-elliptical crack considering stress discontinuity at the base/cladding interface, PTS transient database, and others. A generalized analysis method is proposed on the basis of the development of PASCAL Ver.2 and results of sensitivity analyses. Graphical user interface (GUI) including a generalized method as default values has been also developed for PASCAL Ver.2. This report provides the user's manual and theoretical background of PASCAL Ver.2. (author)

Numerical simulation of a Charpy test and correlation of fracture toughness with fracture energy. Vessel steel and duplex stainless steel of the primary loop

International Nuclear Information System (INIS)

Breban, P; Eripret, C.

1995-01-01

The analysis methods used to evaluate the harmlessness of defects in the components of the primary coolant circuit of pressurized water reactor are based on the knowledge of the failure properties of concerned materials. The toughness is used to be measured through tests performed on normalized samples. But in some cases, especially for the vessel steel submitted to irradiation effects or for cast components in duplex stainless steel sensitive to thermal ageing, these measurements are not available on the material aged in operation. Therefore, fracture resistance has been evaluated through Charpy tests. Toughness is thus obtained on the basis of an empirical correlation. To improve these predictions, a modeling of the Charpy test in the framework of the local approach to fracture has been performed, for both materials. For the vessel steel, a complete evaluation of toughness has been achieved on the basis of a bidimensional viscoplastic modeling under large strain assumptions and a post-treatment with a Weibull model (cleavage fracture). The main hypothesis (partition between plain stress and plain strain areas in the bidimensional modeling) was corrected after a three dimensional calculations with the finite element program Code-Aster. The fracture analysis put into evidence that damage considerations like cavity nucleation and growth have to be introduced in the model in order to improve the description of physical phenomena. Two ways of progress have been suggested and are in course of being investigated, one in the framework of local approach to failure, the other with the help of micro-macro relationship. With regard to the duplex steel, the description of a Charpy (U) test allowed to clearly discriminate between crack initiation and propagation phases. A modeling through an equivalent homogenous material with a damage law based on a modified Gurson potential enables to describe quantitatively both phases of fracture. It clearly appears that a reliable

Total Pancreatic Fracture Due to Blunt Trauma: Report of a Rare Case

Directory of Open Access Journals (Sweden)

Kamil Gulpinar

2016-05-01

Full Text Available A rare case of pancreatic fracture due to blunt trauma was presented. The patient was 70 year old male who had a motor vehicle collision and was suspected a pancreatic trauma due his examinations with ultrasound and computerized tomography. The diagnosis of splenic injury and pancreas body total fracture in the point where the portal vein crosses the pancreatic body was made with the help of magnetic resonance cholangiopancreatography. He was taken to emergency surgery where a splenectomy and a distal pancreatectomy were performed. We represented this infrequent case of pancreatic fracture and its complications after blunt abdominal trauma and discuss the diagnostic and management practices.

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

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

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

The fracture mechanical significance of cracks formed during stress-relief annealing of a submerged arc weldment in pressure vessel steel of type A508 class 2

International Nuclear Information System (INIS)

Liljestrand, L.-G.; Oestberg, G.

1978-01-01

In large weldments of type A508 C12 cracks can form in the heat-affected zone during stress-relief annealing. The significance of such cracks with respect to catastrophic fracture is of interest from the point of view of safety, in particular for nuclear pressure vessels. In this investigation the size of reheat cracks, as formed and after fatigue growth, has been compared with the critical size for fast fracture. The latter was assessed by determination of the toughness of the heat-affected zones. The fracture toughness of the heat-affected zones did not differ much from that of the parent material. The presence of microcracks reduced the fracture toughness (of a special type of simulated specimen) at 20 0 C by about 20%. The fracture mechanical evaluation indicates that the cracks formed during stress-relief annealing should not impair the safety of the vessel under normal conditions, except for particular geometries and when the cracks may rapidly link together during fatigue. (author)

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

Fracture mechanical analysis of relevant transients in the pressure vessel of Atucha I reactor

International Nuclear Information System (INIS)

Saavedra, Fernando M.

2001-01-01

The evolution of the applied stress intensity factor K I for 10 relevant transients of the nuclear power station Atucha I obtained from thermohydraulic data is analyzed according to the methodology proposed in Section XI of ASME Boiler and Pressure Vessel Code. Vast knowledge was thus obtained about basic concepts of fracture mechanics and its application to remanent life of nuclear components. Basic knowledge which commands the performance of nuclear power stations was also obtained, especially that related to the Atucha I utility [es

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

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

Considerations on the manner to account for fast fracture risk in the design of PWR vessels

International Nuclear Information System (INIS)

Pellisier-Tanon, A.; Grandemange, J.M.

1985-08-01

The way followed in France for analyzing fast fracture resistance of PWR primary components is the one of a deterministic analysis with safety coefficients imposed in the fracture criteria. The study of margins towards fast fracture of the 900 MWe program vessels undertaken in 1982 includes parametric evaluations of the influence of essential variables. It has stimulated further thoughts on the level of safety to fix in the analysis methodology, on the orientations for the choice of safety factors and on the manner to introduce them in the analysis. A first chapter tries to characterize the French approach in comparison to those of other countries. A second chapter examines the manner according to which safety factors can be introduced in the deterministic analysis. It presents the principle for a logical approach accounting for the interdependency of all factors and variables. It establishes criteria for the selection of defect kind and size for the computation

Flaw preparations for HSST program vessel fracture mechanics testing: mechanical-cyclic pumping and electron-beam weld-hydrogen-charge cracking schemes

International Nuclear Information System (INIS)

Holz, P.P.

1980-06-01

The purpose of the document is to present schemes for flaw preparations in heavy section steel. The ability of investigators to grow representative sharp cracks of known size, location, and orientation is basic to representative field testing to determine data for potential flaw propagation, fracture behavior, and margin against fracture for high-pressure-, high-temperature-service steel vessels subjected to increasing pressurization and/or thermal shock. Gaging for analytical stress and strain procedures and ultrasonic and acoustic emission instrumentation can then be applied to monitor the vessel during testing and to study crack growth. This report presents flaw preparations for HSST fracture mechanics testing. Cracks were grown by two techniques: (1) a mechanical method wherein a premachined notch was sharpened by pressurization and (2) a method combining electron-beam welds and hydrogen charging to crack the chill zone of a rapidly placed autogenous weld. The mechanical method produces a naturally occurring growth shape controlled primarily by the shape of the machined notch; the welding-electrochemical method produces flaws of uniform depth from the surface of a wall or machined notch. Theories, details, discussions, and procedures are covered for both of the flaw-growing schemes

Early Vessels Exploration of Pink Pulseless Hand in Gartland III Supracondylar Fracture Humerus in Children: Facts and Controversies

Directory of Open Access Journals (Sweden)

Tunku-Naziha TZ

2017-03-01

Full Text Available The management of pink pulseless limbs in supracondylar fractures has remained controversial, especially with regards to the indication for exploration in a clinically well-perfused hand. We reviewed a series of seven patients who underwent surgical exploration of the brachial artery following supracondylar fracture. All patients had a non-palpable radial artery, which was confirmed by Doppler ultrasound. CT angiography revealed complete blockage of the artery with good collateral and distal run-off. Two patients were more complicated with peripheral nerve injuries, one median nerve and one ulnar nerve. Only one patient had persistent arterial constriction which required reverse saphenous graft. The brachial arteries were found to be compressed by fracture fragments, but were in continuity. The vessels were patent after the release of obstruction and the stabilization of the fracture. There was no transection of major nerves. The radial pulse was persistently present after 12 weeks, and the nerve activity returned to full function.

Open Fracture of the Forearm Bones due to Horse Bite

OpenAIRE

Santoshi, John Ashutosh; Leshem, Lall

2014-01-01

Introduction: Fractures have been described mainly following falling accidents in horse-related injuries. Horse bites are uncommon accidents. We present a case of open fracture of the forearm due to horse bite. Case Report: A 35-year-old male farm-worker presented to the emergency room with alleged history of horse bite to the right forearm about 2 hours prior to presentation while feeding the horse. There was deformity of the forearm with multiple puncture wounds, deep abrasions and small...

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)

Prevention against fragile fracture in PWR pressure vessel in the presence of pressurized thermal shock

International Nuclear Information System (INIS)

Carmo, E.G.D. do; Oliveira, L.F.S. de; Roberty, N.C.

1984-01-01

A method for the determination of operational limit curves (primary pressure versus temperature) for PWR is presented. Such curves give the operators indications related to the safety status of the plant concerning the possibility of a pressurized thermal shock. The method begins by a thermal analysis for several postulated transients, followed by the determination of the thermomechanical stresses in the vessel and finally it makes use of the linear elasticity fracture mechanics. Curves are shown for a typical PWR. (Author) [pt

A critical review on the application of elastic-plastic fracture mechanics to nuclear pressure vessel and piping systems

International Nuclear Information System (INIS)

Scarth, D.A.; Kim, Y.J.; Vanderglas, M.L.

1985-10-01

A comprehensive literature survey on the application of Elastic-Plastic Fracture Mechanics to the assessment of the structural integrity of nuclear pressure vessels and piping is presented. In particular, the J-integral/Tearing Modulus (J/T) approach and the Failure Assessment Diagram (FAD) are covered in detail because of their general suitability for use in Ontario Hydro. (25 refs.)

Ductile fracture prediction of an axially cracked pressure vessel under pressurized thermal shock

International Nuclear Information System (INIS)

Takahashi, Jun; Okamura, Hiroyuki

1991-01-01

In this paper, the J-value of an axially cracked cylinder under several PTS conditions are evaluated using a simple estimation scheme which we proposed. Results obtained are summerized as follow: (1) Under any PTS conditions, the effect of internal pressure is so predominant upon the J-value and dJ/da that it is very important to grasp the transient of internal pressure under any imaginable accident from the viewpoint of structural integrity. (2) Under any IP, TS, and PTS conditions, J - a/W relation shows that the J-value reaches its maximum at a certain crack depth, then drops to zero at a/W ≅ 0.9. Though the effect of inertia is not taken into account, this fact may explain the phenomena of crack arrest qualitatively. (3) The compliance of a cylindrical shell plays an important role in the fracture prediction of a pressure vessel. (4) Under typical PTS conditions, the region at the crack tip dominated by the Hutchinson-Rice-Rosengren singularity is substantially large enough to apply the J-based criterion to predict unstable ductile fracture. (author)

Fracture toughness and crack growth resistance of pressure vessel plate and weld metal steels

International Nuclear Information System (INIS)

Moskovic, R.

1988-01-01

Compact tension specimens were used to measure the initiation fracture toughness and crack growth resistance of pressure vessel steel plates and submerged arc weld metal. Plate test specimens were manufactured from four different casts of steel comprising: aluminium killed C-Mn-Mo-Cu and C-Mn steel and two silicon killed C-Mn steels. Unionmelt No. 2 weld metal test specimens were extracted from welds of double V butt geometry having either the C-Mn-Mo-Cu steel (three weld joints) or one particular silicon killed C-Mn steel (two weld joints) as parent plate. A multiple specimen test technique was used to obtain crack growth data which were analysed by simple linear regression to determine the crack growth resistance lines and to derive the initiation fracture toughness values for each test temperature. These regression lines were highly scattered with respect to temperature and it was very difficult to determine precisely the temperature dependence of the initiation fracture toughness and crack growth resistance. The data were re-analysed, using a multiple linear regression method, to obtain a relationship between the materials' crack growth resistance and toughness, and the principal independent variables (temperature, crack growth, weld joint code and strain ageing). (author)

On the dynamic fracture toughness and crack tip strain behavior of nuclear pressure vessel steel: Application of electromagnetic force

International Nuclear Information System (INIS)

Yagawa, G.; Yoshimura, S.

1986-01-01

This paper is concerned with the application of the electromagnetic force to the determination of the dynamic fracture toughness of materials. Taken is an edge-cracked specimen which carries a transient electric current and is simply supported in a steady magnetic field. As a result of their interaction, the dynamic electromagnetic force occurs in the whole body of the specimen, which is then deformed to fracture in the opening mode of cracking. Using the electric potential and the J-R curve methods to determine the dynamic crack initiation point in the experiment, together with the finite element method to calculate the extended J-integral with the effects of the electromagnetic force and inertia, the dynamic fracture toughness values of nuclear pressure vessel steel A508 class 3 are evaluated over a wide temperature range from lower to upper shelves. The strain distribution near the crack tip in the dynamic process of fracture is also obtained by applying a computer picture processing. (orig.)

Application of probabilistic fracture mechanics to reactor pressure vessel safety assessment

International Nuclear Information System (INIS)

Venturini, V.; Pitner, P.

1995-06-01

Among all the components of a PWR (Pressurized Water Reactor) nuclear power plant, the reactor vessel is of major importance for safety. The integrity of this structure must be guaranteed in all circumstances, even in the case of the most severe accidents, and its mechanical state can be decisive for the lifetime of the plant. The brittle rupture would be the most important of all potential hazards if the irradiation effects were not consistent with predictions. The interest of having a reliable and precise method of evaluating the available safety margins and the integrity of this component led Electricite de France (EDF) to carry out a probabilistic fracture mechanics analysis. The probabilistic model developed by integration of the uncertainties in the usual fracture mechanics equations is presented. A special focus is made on the problem of coupling thermo-mechanical finite element calculations and reliability analysis. The use of a finite element code can be associated with prohibitive computation times when it is invoked numerous times during simulations sequences or complex iterative procedures. The response surface method is used. It provides an approximation of the response from a reduced number of original data. The global approach is illustrated on an example corresponding to a specific accidental transient. A validation of the obtained results is also carried out through the comparison with an equivalent model without coupling. (author)

Consideration on the Mechanism of Microwave Emission Due to Rock Fracture

Science.gov (United States)

Takano, Tadashi; Sugita, Seiji; Yoshida, Shingo; Maeda, Takashi

2010-05-01

Microwave emission due to rock fracture was found at 300 MHz, 2 GHz, and 22 GHz, and its power was calibrated in laboratory for the first time in the world. The observed waveform is impulsive, and contains correspondent frequency component inside the envelope at each frequency band. At such high frequencies, the electro-magnetic signal power can be calibrated as a radiating wave with high accuracy. Accordingly, it was verified that a substantial power is emitted. The microwave emission phenomena were also observed on occasions of hypervelocity impact, and esteemed as phenomena generally associated with material destruction. Earthquakes and volcanic activities are association with rock fractures so that the microwave is expected to be emitted. Actually, the e emission was confirmed by the data analysis of the brightness temperature obtained by a remote sensing satellite, which flew over great earthquakes of Wuenchan and Sumatra, and great volcanic eruptions of Reventador and Chanten. It is important to show the microwave emission during rock fracture in natural phenomena. Therefore, the field test to detect the microwave due to the collapse of a crater cliff was planned and persecuted at the volcano of Miyake-jima about 100 km south of Tokyo. Volcanic activity may be more convenient than an earthquake because of the known location and time. As a result, they observed the microwave emission which was strongly correlated with the cliff collapses. Despite of the above-mentioned phenomenological fruits, the reason of the microwave emission is not fixed yet. We have investigated the mechanism of the emission in consideration of the obtained data in rock fracture experiments so far and the study results on material destruction by hypervelocity impact. This paper presents the proposal of the hypothesis and resultant discussions. The microwave sensors may be useful to monitor natural hazards such as an earthquake or a volcanic eruption, because the microwave due to rock

Fracture toughness calculation using dynamic testing

International Nuclear Information System (INIS)

Perosanz, F. J.; Serrano, M.; Martinez, C.; Lapena, J.

1998-01-01

The most critical component of a Nuclear Power Station is the Reactor Pressure Vessel (RPV), due to safety and integrity requirements. The RPV is subjected to neutron radiation and this phenomenon lead to microstructural changes in the material and modifications in the mechanical properties. Due to this effects, it is necessary to assess the structural integrity of the RPV along the operational life through surveillance programs. The main objective of this surveillance programs is to determine the fracture toughness of the material. At present this objective is reached combining direct measures and prediction techniques. In this work, direct measures of fracture toughness using instrumented Charpy V impact testing are present using a CIEMAT development on analysis of results. (Author) 6 refs

Fracture analyses of WWER reactor pressure vessels

International Nuclear Information System (INIS)

Sievers, J.; Liu, X.

1997-01-01

In the paper first the methodology of fracture assessment based on finite element (FE) calculations is described and compared with simplified methods. The FE based methodology was verified by analyses of large scale thermal shock experiments in the framework of the international comparative study FALSIRE (Fracture Analyses of Large Scale Experiments) organized by GRS and ORNL. Furthermore, selected results from fracture analyses of different WWER type RPVs with postulated cracks under different loading transients are presented. 11 refs, 13 figs, 1 tab

Fracture analyses of WWER reactor pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Sievers, J; Liu, X [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany)

1997-09-01

In the paper first the methodology of fracture assessment based on finite element (FE) calculations is described and compared with simplified methods. The FE based methodology was verified by analyses of large scale thermal shock experiments in the framework of the international comparative study FALSIRE (Fracture Analyses of Large Scale Experiments) organized by GRS and ORNL. Furthermore, selected results from fracture analyses of different WWER type RPVs with postulated cracks under different loading transients are presented. 11 refs, 13 figs, 1 tab.

Bilateral recurrent anterior fracture dislocation of shoulder joint due to grand mal epileptic convulsions

Directory of Open Access Journals (Sweden)

Chandrashekara Chowdipalya Maliyappa

2013-01-01

Full Text Available Bilateral shoulder dislocation is very much common with convulsions of different etiology. Often, these dislocations are associated with fractures due to violent muscle contractions. The typical lesion is bilateral posterior dislocation or fracture dislocations. The recurrent shoulder dislocations are common with traumatic etiology. The lack of asymmetry of the shoulders is stressed as a potential pitfall in the clinical evaluation of patients with this condition. We present a rare case of bilateral recurrent anterior fracture dislocation of the shoulder sustained due to repetitive episodes of convulsive seizures. Patient was treated by close reductions and immobilization on each episode. In epilepsy although posterior dislocations are common, the rare possibility of bilateral anterior fracture dislocation should be kept in mind. Often these patients are vulnerable for recurrence, similar to traumatic cases.

Open Fracture of the Forearm Bones due to Horse Bite

Directory of Open Access Journals (Sweden)

John Ashutosh Santoshi

2014-01-01

Full Text Available Introduction: Fractures have been described mainly following falling accidents in horse-related injuries. Horse bites are uncommon accidents. We present a case of open fracture of the forearm due to horse bite. Case Report: A 35-year-old male farm-worker presented to the emergency room with alleged history of horse bite to the right forearm about 2 hours prior to presentation while feeding the horse. There was deformity of the forearm with multiple puncture wounds, deep abrasions and small lacerations on the distal-third of the forearm. Copious irrigation with normal saline was done and he was administered anti-tetanus and post-exposure rabies prophylaxis. Prophylactic antibiotic therapy was commenced. Radiographs revealed fracture of radius and ulna in the mid-shaft region. He underwent emergency wound debridement, and the ulna was stabilised with an intra-medullary square nail. Seventy-two hours later, he underwent re-debridement and conversion osteosynthesis. He had an uneventful recovery and at three-month follow-up, the fractures had healed radiographically in anatomic alignment. At two-year follow-up, he is doing well, is pain free and has a normal range of motion compared to the contralateral side. Conclusion: Horse bites behave as compound fractures however rabies prophylaxis will be needed and careful observation is needed. Early radical debridement, preliminary skeletal stabilisation, re-debridement and conversion osteosynthesis to plate, and antibiotic prophylaxis were the key to the successful management of our patient. Keywords: Horse; animal bite; forearm; open fracture

Probabilistic fracture mechanics analysis of boiling water reactor vessel for cool-down and low temperature over-pressurization transients

Energy Technology Data Exchange (ETDEWEB)

Park, Jeong Soon; Choi, Young Hwan; Jhung, Myung Jo [Safety Research Division, Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2016-04-15

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

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

Test of 6-in.-thick pressure vessels. Series 4: intermediate test vessels V-5 and V-9 with inside nozzle corner cracks

International Nuclear Information System (INIS)

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

1977-01-01

Failure testing is described for two 99-cm-diam (39-in.), 15.2-cm-thick (6-in.) steel pressure vessels, each containing one flawed nozzle. Vessel V-5 was tested at 88 0 C (190 0 F) and failed by leaking without fracturing after extensive stable crack growth. Vessel V-9 was tested at 25 0 C (75 0 F) and failed by fracturing. Material properties measured before the tests were used for pretest and posttest fracture analyses. Test results supported by analysis indicate that inside nozzle corner cracks are not subject to plane strain under pressure loading. The preparation of inside nozzle corner cracks is described in detail. Extensive experimental data are tabulated and plotted

Application of ductile fracture assessment methods for the assessment of pressure vessels from high strength steels (HSS)

International Nuclear Information System (INIS)

Eisele, U.; Schiedermaier, J.

2003-01-01

The economical and safe design of pressure vessels requires, besides others, also a detailed knowledge of the vessel failure behaviour in the case of existing imperfections or cracks. The behaviour of a cracked component under a given loading situation depends on material toughness. For ferritic steels, the material toughness is varying with temperature. At low temperature dominantly brittle fracture behaviour is observed, at high temperature the failure mode is dominantly ductile fracture. The transition between these two extremes is floating. In the case of existing or postulated cracks, the safety analysis has to be performed using fracture mechanics methods. In the lower shelf of toughness, K iC as of ASTM E 399 is the characterising value for crack initiation and immediate unstable crack extension (cleavage). In the upper shelf level the characterising value is the ''actual crack initiation toughness'' J i acc. to ISO 12135, characterising the onset of slow stable crack extension. For the transition regime in ASTM E 1921 the instability values K JC are defined, characterising cleavage failure after more or less extended ductile crack growth. The safety analysis of a component operated in the upper shelf of the material toughness, has to consider initiation as well as stable crack extension following initiation. The inclusion of any crack extension into this consideration needs to consider the influence of the constraint in front of a crack tip, leading to multiaxial stress conditions and decreasing the material crack resistance significantly. Thus, the exclusion of crack initiation needs to be proven in a first step of each safety analysis. Assessing the component in a uniform way over the relevant temperature range is possible by using initiation characteristics, which also have the advantage of transferability. A change of criterion considering initiation at the lower shelf, instability in the transition range and again initiation in the upper shelf can be

Assessment of the effects of neutron fluence on Swedish nuclear pressure vessels

International Nuclear Information System (INIS)

Rao, S.

1980-11-01

Nuclear pressure vessels are subject to neutron irradiation during service causing embrittlement. This is one important factor in the overall problem of reactor vessel integrity. At present the irradiation effects are mainly assessed by the Charpy V-notch test. Two measures of embrittlement are defined: the increase of the ductile/brittle transition temperature and the decrease in the upper-shelf energy. The object of the present work is to assess these changes for the Swedish nuclear pressure vessels. On the basis of data from irradiations carried out in other countries and Swedish surveillance programmes, the expected end of life embrittlement is estimated for Swedish vessels. The results show that the embrittlement of most reactor vessels is expected to be quite small. Oskarshamn 1 and PWR-vessels, however, will probably show moderate changes, the former due to the higher copper content, and the latter due to the high end of life fluences. Some of the vessel materials which exhibit marginal properties in the upper-shelf energy, as measured by the Charpy V-notch impact test, are identified. It is recommended that fracture mechanics analyses be applied in these cases. (author)

Rib fixation for severe chest deformity due to multiple rib fractures.

Science.gov (United States)

Igai, Hitoshi; Kamiyoshihara, Mitsuhiro; Nagashima, Toshiteru; Ohtaki, Yoichi

2012-01-01

The operative indications for rib fracture repair have been a matter of debate. However, several reports have suggested that flail chest, pain on respiration, and chest deformity/defect are potential conditions for rib fracture repair. We describe our experience of rib fixation in a patient with severe chest deformity due to multiple rib fractures. A 70-year-old woman was admitted with right-sided multiple rib fractures (2nd to 7th) and marked chest wall deformity without flailing caused by an automobile accident. Collapse of the chest wall was observed along the middle anterior axillary line. At 11 days after the injury, surgery was performed to repair the chest deformity, as it was considered to pose a risk of restrictive impairment of pulmonary function or chronic intercostal pain in the future. Operative findings revealed marked displacement of the superior 4 ribs, from the 2nd to the 5th, and collapse of the osseous chest wall towards the thoracic cavity. After exposure of the fracture regions, ribs fixations were performed using rib staplers. The total operation time was 90 minutes, and the collapsed portion of the chest wall along the middle anterior axillary line was reconstructed successfully.

Considerations of the manner of accounting for fast fracture risk in the design of PWR vessels

International Nuclear Information System (INIS)

Pellissier-Tanon, A.; Grandemange, J.M.

1986-01-01

The French approach to the prevention of fast fracture in PWR vessels is to consider it as a whole and to choose the most convenient way to meet this general goal from an economic and technical point of view. According to this approach, there are no specific limits imposed on such factors as end of life RTsub(NDT) or neutron fluence, which are taken as criteria in other countries. The RCCM design and construction code specifications on chemical content and RTsub(NDT) for beltline and non-irradiated parts establish a sound basis for safety. However, for the most critical parts, the existence of large margins with respect to fast fracture is demonstrated by analysis for all second, third and fourth category design transients. To this aim, the RCCM code needs to demonstrate specified safety margins, depending on the transient category, for reference defects defined in kind and size, in order to bound realistically any defects which have a chance of occurring in the part during manufacture. This approach, which enables the disclosure of the influence of all significant design factors on fracture risk, ensures the most consistent way to improve design safety. (author)

Considerations of the manner of accounting for fast fracture risk in the design of PWR vessels

Energy Technology Data Exchange (ETDEWEB)

1986-01-01

The French approach to the prevention of fast fracture in PWR vessels is to consider it as a whole and to choose the most convenient way to meet this general goal from an economic and technical point of view. According to this approach, there are no specific limits imposed on such factors as end of life RTsub(NDT) or neutron fluence, which are taken as criteria in other countries. The RCCM design and construction code specifications on chemical content and RTsub(NDT) for beltline and non-irradiated parts establish a sound basis for safety. However, for the most critical parts, the existence of large margins with respect to fast fracture is demonstrated by analysis for all second, third and fourth category design transients. To this aim, the RCCM code needs to demonstrate specified safety margins, depending on the transient category, for reference defects defined in kind and size, in order to bound realistically any defects which have a chance of occurring in the part during manufacture. This approach, which enables the disclosure of the influence of all significant design factors on fracture risk, ensures the most consistent way to improve design safety.

Phenomenological vessel burst investigations

International Nuclear Information System (INIS)

Hippelein, K.W.; Julisch, P.; Muz, J.; Schiedermaier, J.

1985-07-01

Fourteen burst experiments have been carried out using vessels with circumferential and longitudinal flaws, for investigation of the fracture behaviour, i.e. the time-related fracture opening. The vessels had dimensions (outer diameter x wall thickness = 800 x 47 mm) which correspond to the dimensions of the main coolant piping of a 1300 MW e PWR. The test specimens had been made of the base-safe material 20 MnMoNi 55 and of a special, 22 NiMoCr 37 base alloy. The experimental conditions with regard to pressure and temperature have been chosen so as to correspond to normal operating conditions of a PWR (p∝17.5 MPa, T∝300 0 C), i.e. the flaws have been so dimensioned that failure was to be expected at a pressure of p∝17.5 MPa. As a rule, water has been used as the pressure medium, or in some cases air, in order to influence the time-dependent pressure decrease. Fluid and structural dynamics calculations have also been made. In order to determine the impact of a fast propagating crack on the leak-to-fracture curve, which normally is defined by quasistationary experiments, suitable tests have been made with large-volume, cylindrical vessels (outer diameter x wall thickness x length = 3000 x 21 x 14000 mm) made of the material WSt E 43. The leak-before-fracture criterion has been confirmed. (orig./HP) [de

Documentation of probabilistic fracture mechanics codes used for reactor pressure vessels subjected to pressurized thermal shock loading: Parts 1 and 2. Final report

International Nuclear Information System (INIS)

Balkey, K.; Witt, F.J.; Bishop, B.A.

1995-06-01

Significant attention has been focused on the issue of reactor vessel pressurized thermal shock (PTS) for many years. Pressurized thermal shock transient events are characterized by a rapid cooldown at potentially high pressure levels that could lead to a reactor vessel integrity concern for some pressurized water reactors. As a result of regulatory and industry efforts in the early 1980's, a probabilistic risk assessment methodology has been established to address this concern. Probabilistic fracture mechanics analyses are performed as part of this methodology to determine conditional probability of significant flaw extension for given pressurized thermal shock events. While recent industry efforts are underway to benchmark probabilistic fracture mechanics computer codes that are currently used by the nuclear industry, Part I of this report describes the comparison of two independent computer codes used at the time of the development of the original U.S. Nuclear Regulatory Commission (NRC) pressurized thermal shock rule. The work that was originally performed in 1982 and 1983 to compare the U.S. NRC - VISA and Westinghouse (W) - PFM computer codes has been documented and is provided in Part I of this report. Part II of this report describes the results of more recent industry efforts to benchmark PFM computer codes used by the nuclear industry. This study was conducted as part of the USNRC-EPRI Coordinated Research Program for reviewing the technical basis for pressurized thermal shock (PTS) analyses of the reactor pressure vessel. The work focused on the probabilistic fracture mechanics (PFM) analysis codes and methods used to perform the PTS calculations. An in-depth review of the methodologies was performed to verify the accuracy and adequacy of the various different codes. The review was structured around a series of benchmark sample problems to provide a specific context for discussion and examination of the fracture mechanics methodology

Methods for estimation and enhancing of resistance of pressure vessel materials to fracture at different stages of service taking into account actual dimensions of the construction

International Nuclear Information System (INIS)

Pokrovsky, V.V.; Ivanchenko, A.G.

1998-01-01

In the present report a method is proposed for assessment of cracked materials fracture toughness over a wide range of temperatures taking into account the size-effect of structural elements. The procedure proposed was evaluated on specimens of different thicknesses (25... 150 mm) and geometries from the parent metal and welded joint metal of the WWER-Type nuclear reactor pressure vessels of different classes of strength. The method of enhancing of fracture resistance of pressure vessel materials has been develop which is based on warm prestressing of materials with cracks. The stability of the favourable effect of the warm prestressing has been, investigated and shown for the above steels after their long term (to 24000 hours) keeping under static loading and temperature of 350 deg C, under different conditions of cyclic loading, corrosive action. A model and calculation procedure are proposed for predicting the influence of thermomechanical loading conditions on the resistance of reactor steels to brittle fracture. (authors)

Application of Master Curve fracture toughness for reactor pressure vessel integrity assessment in the USA

International Nuclear Information System (INIS)

Server, William; Rosinski, Stan; Lott, Randy; Kim, Charles; Weakland, Dennis

2002-01-01

The Master Curve fracture toughness approach has been used in the USA for better defining the transition temperature fracture toughness of irradiated reactor pressure vessel (RPV) steels for end-of-life (EOL) and EOL extension (EOLE) time periods. The first application was for the Kewaunee plant in which the life-limiting material was a circumferential weld metal. Fracture toughness testing of this weld metal corresponding to EOL and beyond EOLE was used to reassess the PTS screening value, RT PTS , and to develop new operating pressure-temperature curves. The NRC has approved this application using a shift-based methodology and higher safety margins than those proposed by the utility and its contractors. Beaver Valley Unit 1, a First Energy nuclear plant, has performed similar fracture toughness testing, but none of the testing has been conducted at EOL or EOLE at this time. Therefore, extrapolation of the life-limiting plate data to higher fluences is necessary, and the projections will be checked in the next decade by Master Curve fracture toughness testing of all of the Beaver Valley Unit 1 beltline materials (three plates and three welds) at fluences near or greater than EOLE. A supplemental surveillance capsule has been installed in the sister plant, Beaver Valley Unit 2, which has the capability of achieving a higher lead factor while operating under essentially the same environment. The Beaver Valley Unit 1 evaluation has been submitted to the NRC. This paper reviews the shift-based approach taken for the Beaver Valley Unit 1 RPV and presents the use of the RT T 0 methodology (which evolved out of the Master Curve testing and endorsed through two ASME Code Cases). The applied margin accounts for uncertainties in the various material parameters. Discussion of a direct measurement of RT T 0 approach, as originally submitted for the Kewaunee case, is also presented

Fractures and Fanconi syndrome due to prolonged sodium valproate use.

Science.gov (United States)

Dhillon, N; Högler, W

2011-06-01

Sodium valproate (VPA) is commonly used to treat epilepsy in children. Renal dysfunction is a rare side eff ect but can present as tubulopathy such as Fanconi syndrome. We report on an 8-year-old disabled girl with myoclonic epilepsy who was referred for investigation of recurrent low impact fractures of the distal femur which were initially thought to be caused by her severe immobility. However, she was subsequently found to have hypophosphataemia secondary to Fanconi syndrome due to prolonged VPA use. After VPA withdrawal renal function and serum phosphate levels normalised and X-rays improved dramatically. The possibility of drug-induced osteoporosis and fractures should always be considered in disabled children, even in the presence of severe immobility.

Multilayer Pressure Vessel Materials Testing and Analysis Phase 2

Science.gov (United States)

Popelar, Carl F.; Cardinal, Joseph W.

2014-01-01

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

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

International Nuclear Information System (INIS)

Sattari-Far, I.

2001-03-01

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

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.)

Stress criteria for nuclear vessel concrete

International Nuclear Information System (INIS)

Costes, D.

1975-01-01

Concrete nuclear vessels are submitted to prestressing forces which limit tensile stresses in concrete when the vessel is under pressure with thermal gradients. Hence, the most severe conditions for concrete appear when the vessel is prestressed and not submitted to internal pressure. The triaxial states of stress in the concrete may be computed postulating elastic or other behavior and compared with safe limits obtained from rupture tests and fatigue tests. The first part of the paper, recalls experimental rupture results and the acceptability procedures currently used. Criteria founded on the lemniscoid surfaces are proposed, parameters for which are obtained by various tests and safety considerations. In the second part, rupture tests are reported on small, thick, cylindrical vessels submitted to external hydraulic pressure simulating prestressing forces. Materials used are plain concrete, microconcrete, marble and graphite. The strengths obtained are much higher than those which could be elastically computed, triaxial rupture states being provided by previous experiments. Such results may be due to a plastic stress redistribution before fracture and to stabilizing effects of stress gradients around the more stressed areas. Fatigue tests by external hydraulic loading are reported [fr

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

International Nuclear Information System (INIS)

Pennell, W.E.

1992-01-01

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

Effects of temperature on corrosion fatigue crack growth of pressure vessel steels in PWR coolant

International Nuclear Information System (INIS)

Tice, D.R.; Bramwell, I.L.; Fairbrother, H.; Worswick, D.

1994-01-01

This paper presents experimental results concerning crack propagation rates in A508-III pressure vessel steel (medium sulphur content) exposed to PWR primary water at temperatures between 130 and 290 C. The results indicate that the greatest increase in corrosion fatigue crack growth rate occurs at temperatures in the range 150 to 200 C. Under these conditions, there was a marked change in the appearance of the fracture surface, with extensive micro-branching of the crack front and occasional bifurcation of the whole crack path. In contrast, at 290 C, the fracture surface is smoother, similar to that due to inert fatigue. The implication of these observations for assessment of the pressure vessel integrity, is examined. 14 refs., 15 figs., 3 tabs

Overview of input parameters for calculation of the probability of a brittle fracture of the reactor pressure vessel

International Nuclear Information System (INIS)

Horacek, L.

1994-12-01

The parameters are summarized for a calculation of the probability of brittle fracture of the WWER-440 reactor pressure vessel (RPV). The parameters were selected for 2 basic approaches, viz., one based on the Monte Carlo method and the other on the FORM and SORM methods (First and Second Order Reliability Methods). The approaches were represented by US computer codes VISA-II and OCA-P and by the German ZERBERUS code. The philosophy of the deterministic and probabilistic aspects of the VISA-II code is outlined, and the differences between the US and Czech PWR's are discussed in this context. Briefly described is the partial approach to the evaluation of the WWER type RPV's based on the assessment of their resistance to brittle fracture by fracture mechanics tools and by using the FORM and SORM methods. Attention is paid to the input data for the WWER modification of the VISA-II code. The data are categorized with respect to randomness, i.e. to the stochastic or deterministic nature of their behavior. 18 tabs., 14 refs

Age Related Macular Degeneration and Total Hip Replacement Due to Osteoarthritis or Fracture: Melbourne Collaborative Cohort Study.

Directory of Open Access Journals (Sweden)

Elaine W Chong

Full Text Available Osteoarthritis is the leading cause of total hip replacement, accounting for more than 80% of all total hip replacements. Emerging evidence suggests that osteoarthritis has a chronic inflammatory component to its pathogenesis similar to age-related macular degeneration. We evaluated the association between age-related macular degeneration and total hip replacement as proxy for severe osteoarthritis or fractured neck of femur in the Melbourne Collaborative Cohort Study. 20,744 participants had complete data on both age-related macular degeneration assessed from colour fundus photographs taken during 2003-2007 and total hip replacement. Total hip replacements due to hip osteoarthritis and fractured neck of femur during 2001-2011 were identified by linking the cohort records to the Australian Orthopedic Association National Joint Replacement Registry. Logistic regression was used to examine the association between age-related macular degeneration and risk of total hip replacement due to osteoarthritis and fracture separately, adjusted for confounders. There were 791 cases of total hip replacement for osteoarthritis and 102 cases of total hip replacement due to fractured neck of femur. After adjustment for age, sex, body mass index, smoking, and grouped country of birth, intermediate age-related macular degeneration was directly associated with total hip replacement for osteoarthritis (odds ratio 1.22, 95% CI 1.00-1.49. Late age-related macular degeneration was directly associated with total hip replacement due to fractured neck of femur (odds ratio 5.21, 95% CI2.25-12.02. The association between intermediate age-related macular degeneration and an increased 10-year incidence of total hip replacement due to osteoarthritis suggests the possibility of similar inflammatory processes underlying both chronic diseases. The association of late age-related macular degeneration with an increased 10-year incidence of total hip replacement due to fractured

Fracture Analysis of Vessels. Oak Ridge FAVOR, v06.1, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations

Energy Technology Data Exchange (ETDEWEB)

Williams, P. T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dickson, T. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yin, S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2007-12-01

The current regulations to insure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models developed in the early-to-mid 1980s. Since that time, advancements and refinements in relevant technologies that impact RPV integrity assessment have led to an effort by the NRC to re-evaluate its PTS regulations. Updated computational methodologies have been developed through interactions between experts in the relevant disciplines of thermal hydraulics, probabilistic risk assessment, materials embrittlement, fracture mechanics, and inspection (flaw characterization). Contributors to the development of these methodologies include the NRC staff, their contractors, and representatives from the nuclear industry. These updated methodologies have been integrated into the Fracture Analysis of Vessels -- Oak Ridge (FAVOR, v06.1) computer code developed for the NRC by the Heavy Section Steel Technology (HSST) program at Oak Ridge National Laboratory (ORNL). The FAVOR, v04.1, code represents the baseline NRC-selected applications tool for re-assessing the current PTS regulations. This report is intended to document the technical bases for the assumptions, algorithms, methods, and correlations employed in the development of the FAVOR, v06.1, code.

Rib Fracture Diagnosis in the Panscan Era.

Science.gov (United States)

Murphy, Charles E; Raja, Ali S; Baumann, Brigitte M; Medak, Anthony J; Langdorf, Mark I; Nishijima, Daniel K; Hendey, Gregory W; Mower, William R; Rodriguez, Robert M

2017-12-01

With increased use of chest computed tomography (CT) in trauma evaluation, traditional teachings in regard to rib fracture morbidity and mortality may no longer be accurate. We seek to determine rates of rib fracture observed on chest CT only; admission and mortality of patients with isolated rib fractures, rib fractures observed on CT only, and first or second rib fractures; and first or second rib fracture-associated great vessel injury. We conducted a planned secondary analysis of 2 prospectively enrolled cohorts of the National Emergency X-Radiography Utilization Study chest studies, which evaluated patients with blunt trauma who were older than 14 years and received chest imaging in the emergency department. We defined rib fractures and other thoracic injuries according to CT reports and followed patients through their hospital course to determine outcomes. Of 8,661 patients who had both chest radiograph and chest CT, 2,071 (23.9%) had rib fractures, and rib fractures were observed on chest CT only in 1,368 cases (66.1%). Rib fracture patients had higher admission rates (88.7% versus 45.8%; mean difference 42.9%; 95% confidence interval [CI] 41.4% to 44.4%) and mortality (5.6% versus 2.7%; mean difference 2.9%; 95% CI 1.8% to 4.0%) than patients without rib fracture. The mortality of patients with rib fracture observed on chest CT only was not statistically significantly different from that of patients with fractures also observed on chest radiograph (4.8% versus 5.7%; mean difference -0.9%; 95% CI -3.1% to 1.1%). Patients with first or second rib fractures had significantly higher mortality (7.4% versus 4.1%; mean difference 3.3%; 95% CI 0.2% to 7.1%) and prevalence of concomitant great vessel injury (2.8% versus 0.6%; mean difference 2.2%; 95% CI 0.6% to 4.9%) than patients with fractures of ribs 3 to 12, and the odds ratio of great vessel injury with first or second rib fracture was 4.4 (95% CI 1.8 to 10.4). Under trauma imaging protocols that commonly

Fracture flow due to hydrothermally induced quartz growth

Science.gov (United States)

Kling, Tobias; Schwarz, Jens-Oliver; Wendler, Frank; Enzmann, Frieder; Blum, Philipp

2017-09-01

Mineral precipitations are a common feature and limitation of initially open, permeable rock fractures by forming sealing structures or secondary roughness in open voids. Hence, the objective of this numerical study is the evaluation of hydraulic properties of fractures sealed by hydrothermally induced needle and compact quartz growth. Phase-field models of progressive syntaxial and idiomorphic quartz growth are implemented into a fluid flow simulation solving the Navier-Stokes equation. Flow simulations for both quartz types indicate an obvious correlation between changes in permeability, fracture properties (e.g. aperture, relative roughness and porosity) and crystal growth behavior, which also forms distinct flow paths. Thus, at lower sealing stages initial fracture permeability significantly drops down for the 'needle fracture' forming highly tortuous flow paths, while the 'compact fracture' records a considerably smaller loss. Fluid flow in both sealing fractures most widely is governed by a ;parallel plate;-like cubic law behavior. However, the 'needle fracture' also reveals flow characteristics of a porous media. A semi-theoretical equation is introduced that links geometrical (am) with hydraulically effective apertures (ah) and the relative fracture roughness. For this purpose, a geometry factor α is introduced being α = 2.5 for needle quartz and α = 1.0 for compact quartz growth. In contrast to most common ah-am-relationships this novel formulation not only reveals more precise predictions for the needle (RMSE = 1.5) and the compact fractures (RMSE = 3.2), but also exhibit a larger range of validity concerning the roughness of the 'needle' (σ/am = 0-2.4) and the 'compact fractures' (σ/am = 0-1.8).

Repeat LISS treatment for femoral shaft fractures due to hardware failure: a retrospective analysis of eleven cases.

Science.gov (United States)

Li, Xu; Xu, Xian; Liu, Lin; Shao, Qin; Wu, Wei

2013-10-01

To evaluate the effectiveness of a replating technique having a less-invasive stabilization system (LISS) for femoral shaft fractures due to LISS failure in adults. There were 11 patients with hardware failure of LISS for femoral shaft fractures, on an average of 50 days after the primary operation. The failed implants were removed, and the fractures were replated with a LISS following the rationale of biological osteosynthesis. Radiological fracture union and incidence of postoperative complications were employed to evaluate the effectiveness of this replating technique for femoral shaft fractures. Operative duration including removing failed hardware and replating fractures averaged 81.5 min, with an average blood loss of 330 ml. Patients had an average follow-up of 25.7 months. Radiological evaluation indicated that fracture union occurred in an average of 4.4 months in all patients. The length and alignment of the affected limb were satisfactory, and hardware failure did not recur. The replating technique with LISS for femoral shaft fractures due to hardware failure of LISS can obtain satisfactory results when the appropriate rationale of biological osteosynthesis and functional exercise is followed.

Assessment of the TRINO reactor pressure vessel integrity: theoretical analysis and NDE

Energy Technology Data Exchange (ETDEWEB)

Milella, P P; Pini, A [ENEA, Rome (Italy)

1988-12-31

This document presents the method used for the capability assessment of the Trino reactor pressure vessel. The vessel integrity assessment is divided into the following parts: transients evaluation and selection, fluence estimate for the projected end of life of the vessel, characterization of unirradiated and irradiated materials, thermal and stress analysis, fracture mechanics analysis and eventually fracture input to Non Destructive Examination (NDE). For each part, results are provided. (TEC).

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-03-01

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

Death due to fracture of thin calvarial bones after a fall: A forensic approach

Directory of Open Access Journals (Sweden)

Georgios Sioutas

2017-06-01

Full Text Available A 45-year-old male was autopsied. He had fallen backwards from a two-stairs height to the ground and passed away. A skull fracture was detected in the left occipital area, extending up to the left side of the skull base. The patient's death occurred due to the very low thickness of the calvarial bones, which led to the aforementioned fracture, and in turn resulted in subarachnoid hemorrhage and death. The cortical thickness was measured and compared with average values at standardized points. Uniform bone thinning was confirmed rather than localized. Calvarial thinning may result from various conditions. In the present case study, however, the exact mechanism which led to the low thickness of the calvarial bones of the patient is undetermined. Death due to the susceptible structure and fracture of calvarial bones has rarely been reported throughout relevant literature.

Post-traumatic cerebellar infarction due to vertebral artery foramina fracture: case report

Directory of Open Access Journals (Sweden)

Moscote-Salazar Luis Rafael

2016-03-01

Full Text Available Posttraumatic cerebral infarction is an uncommon cause of morbidity and mortality and many studies have highlighted that trauma needs to considered as causative factor for cerebellar infarction. We present a case of cerebellar infarction in a 35 year old young patient secondary to vertebral fracture involving the vertebral foramen and vertebral artery injury. CT scan cervical spine showed C2-3 fracture on left side with fracture extending into the left vertebral foramen. A CT scan angiogram could not be performed because of poor neurological status. Possibly the infarction was due to left vertebral artery injury. Without surgical intervention prognosis of these patients remain poor. Prognosis of patients with traumatic cerebellar infarction depends on the neurological status of the patient, intrinsic parenchymal damage and more importantly extrinsic compression of the brainstem by the edematous cerebellar hemispheres.

Applicability of JIS SPV 50 steel to primary containment vessels of nuclear power stations

International Nuclear Information System (INIS)

Iida, K.; Ishikawa, K.; Satoh, M.; Soya, I.

1980-01-01

The fracture toughness of JIS SPV 50 steel and its weldment has been examined in order to verify the applicability of these materials to primary containment vessels of nuclear power stations. Test results were evaluated using elastic plastic fracture mechanics through the COD and the J integral concepts for non ductile fracture initiation characteristics. Linear fracture mechanics was employed for propagation arrest characteristics. Results showed that the materials tested here have a sufficient fracture toughness to prevent nonductile fracture and that this steel is a suitable material for use in construction of primary containment vessels of nuclear power stations. (author)

Radiation embrittlement of Spanish nuclear reactor pressure vessel steels

International Nuclear Information System (INIS)

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

1993-01-01

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

Analytical modeling of the effect of crack depth, specimen size, and biaxial stress on the fracture toughness of reactor vessel steels

International Nuclear Information System (INIS)

Chao, Yuh-Jin

1995-01-01

Fracture, toughness values for A533-B reactor pressure vessel (RPV) steel obtained from test programs at Oak Ridge National Laboratory (ORNL) and University of Kansas (KU) are interpreted using the J-A 2 analytical model. The analytical model is based on the critical stress concept and takes into consideration the constraint effect using the second parameter A 2 in addition to the generally accepted first parameter J which represents the loading level. It is demonstrated that with the constraint level included in the model effects of crack depth (shallow vs deep), specimen size (small vs. large), and loading type (uniaxial vs biaxial) on the fracture toughness from the test programs can be interpreted and predicted

VISA-2, Reactor Vessel Failure Probability Under Thermal Shock

International Nuclear Information System (INIS)

Simonen, F.; Johnson, K.

1992-01-01

1 - Description of program or function: VISA2 (Vessel Integrity Simulation Analysis) was developed to estimate the failure probability of nuclear reactor pressure vessels under pressurized thermal shock conditions. The deterministic portion of the code performs heat transfer, stress, and fracture mechanics calculations for a vessel subjected to a user-specified temperature and pressure transient. The probabilistic analysis performs a Monte Carlo simulation to estimate the probability of vessel failure. Parameters such as initial crack size and position, copper and nickel content, fluence, and the fracture toughness values for crack initiation and arrest are treated as random variables. Linear elastic fracture mechanics methods are used to model crack initiation and growth. This includes cladding effects in the heat transfer, stress, and fracture mechanics calculations. The simulation procedure treats an entire vessel and recognizes that more than one flaw can exist in a given vessel. The flaw model allows random positioning of the flaw within the vessel wall thickness, and the user can specify either flaw length or length-to-depth aspect ratio for crack initiation and arrest predictions. The flaw size distribution can be adjust on the basis of different inservice inspection techniques and inspection conditions. The toughness simulation model includes a menu of alternative equations for predicting the shift in the reference temperature of the nil-ductility transition. 2 - Method of solution: The solution method uses closed form equations for temperatures, stresses, and stress intensity factors. A polynomial fitting procedure approximates the specified pressure and temperature transient. Failure probabilities are calculated by a Monte Carlo simulation. 3 - Restrictions on the complexity of the problem: Maxima of 30 welds. VISA2 models only the belt-line (cylindrical) region of a reactor vessel. The stresses are a function of the radial (through-wall) coordinate only

Numerical modelling of fracture displacements due to thermal load from a KBS-3 repository

Energy Technology Data Exchange (ETDEWEB)

Hakami, Eva; Olofsson, Stig-Olof [Itasca Geomekanik AB, Stockholm (Sweden)

2002-01-01

The objective of the project has been to estimate the largest shear displacements that could be expected on a pre-existing fracture located in the repository area, due to the heat release from the deposited waste. Two-dimensional numerical analyses using the 'Universal Distinct Element Code' (UDEC) have been performed. The UDEC models represent a vertical cross section of a KBS-3 type repository with a large planar fracture intersecting a deposition hole at the repository centre. The extension, dip and mechanical properties of the fracture were changed in different models to evaluate the influence of these parameters on fracture shear displacements. The fracture was modelled using a Coulomb slip criterion with no cohesion and no dilation. The rock mass surrounding the fracture was modelled as a homogeneous, isotropic and elastic material, with a Young's modulus of 40 GPa. The initial heat release per unit repository area was assumed to be 8W/m{sup 2} (total power/total repository area). The shear displacements occur due to the thermal expansion of the rock surrounding the heat generating canisters. The rock mass is almost free to expand vertically, but is constrained horizontally, which gives a temperature-induced addition of shear stresses in the plane of the fracture. The shear movement of the fracture therefore follows the temperature development in the surrounding rock and the maximum shear displacement develops about 200 years after the waste deposition. Altogether, twenty cases are analysed. The maximum shear displacement, which occurs at the fracture centre, amounts to 0.2-13.8 cm depending on the fracture parameters. Among the analysed cases, the largest shear values, about 13 cm, was calculated for the cases with about 700 m long fractures with a shear stiffness of 0.005 GPa/m. Also, for large fractures with a higher shear stiffness of 5 GPa/m, but with a low friction angle (15 deg), the shear displacement reaches similar magnitudes, about

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)

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

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.

Orbital fractures due to domestic violence: an epidemiologic study.

Science.gov (United States)

Goldberg, Stuart H.; McRill, Connie M.; Bruno, Christopher R.; Ten Have, Tom; Lehman, Erik

2000-09-01

Domestic violence is an important cause of orbital fractures in women. Physicians who treat patients with orbital fractures may not suspect this mechanism of injury. The purpose of this study was to assess the association between domestic violence and orbital fractures. A medical center-based case-control study with matching on age and site of admission was done. Medical center databases were searched using ICD-9 codes to identify all cases of orbital fractures encountered during a three-year period. Medical records of female patients age 13 and older were reviewed along with those of age, gender and site of admission matched controls. A stratified exact test was employed to test the association between domestic violence and orbital fracture. Among 41 adult female cases with orbital fractures treated at our medical center, three (7.3%) reported domestic violence compared to zero among the matched controls (p = 0.037). We believe that domestic violence may be under-reported in both orbital fracture cases and controls. This may result in an underestimate of the orbital fracture versus domestic violence association. Domestic violence is a serious women's health and societal problem. Domestic violence may have a variety of presentations, including illnesses and injuries. Orbital fracture is an identifiable manifestation of domestic violence. Domestic violence is more likely to be detected in adult female hospital patients with orbital fracture than in matched controls with any other diagnosis. Physicians who treat patients with orbital fractures should be familiar with this mechanism of injury.

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.)

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

Alzheimer's Disease Increases the Incidence of Hospitalization Due to Fall-related Bone Fracture in Elderly Chinese

Directory of Open Access Journals (Sweden)

Fang Li

2016-12-01

Conclusion: On the basis of our findings, we conclude that AD may increase the incidence of hospitalization due to falls and bone fracture. We also found that AD has no effect on fracture location, but larger studies are needed to confirm this finding. Physicians and family members should emphasize the possibility of falls and bone fracture in patients with AD. Our findings suggest that preventing falls in AD patients may reduce the number of hospitalized AD patients.

Research program plan: reactor vessels. Volume 1

International Nuclear Information System (INIS)

Vagins, M.; Taboada, A.

1985-07-01

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

Death due to fracture of thin calvarial bones after a fall: A forensic approach.

Science.gov (United States)

Sioutas, Georgios; Karakasi, Maria-Valeria; Kapetanakis, Stylianos; Pavlidis, Pavlos

2017-06-01

A 45-year-old male was autopsied. He had fallen backwards from a two-stairs height to the ground and passed away. A skull fracture was detected in the left occipital area, extending up to the left side of the skull base. The patient's death occurred due to the very low thickness of the calvarial bones, which led to the aforementioned fracture, and in turn resulted in subarachnoid hemorrhage and death. The cortical thickness was measured and compared with average values at standardized points. Uniform bone thinning was confirmed rather than localized. Calvarial thinning may result from various conditions. In the present case study, however, the exact mechanism which led to the low thickness of the calvarial bones of the patient is undetermined. Death due to the susceptible structure and fracture of calvarial bones has rarely been reported throughout relevant literature. Copyright © 2017 Daping Hospital and the Research Institute of Surgery of the Third Military Medical University. Production and hosting by Elsevier B.V. All rights reserved.

Estimation scheme for unstable ductile fracture of pressure vessel

International Nuclear Information System (INIS)

Takahashi, Jun; Okamura, Hiroyuki; Sakai, Shinsuke

1990-01-01

This paper presents a new scheme for the estimation of unstable ductile fracture using the J-integral. The proposed method uses a load-versus-displacement diagram which is generated using fully plastic solutions. By this method, the phenomena of the ductile fracture can be grasped visually. Thus, the parametrical survey can be executed far more easily than before. Then, using the proposed method, unstable ductile fracture is analyzed for single-edge cracked plates under both uniform tension and pure bending. In addition, several parametrical surveys are performed concerning (1) J-controlled crack growth, (2) compliance of the structure, (3) ductility of the material (i.e., J-resistance curve), and (4) scale of the structure (i.e., screening criterion). As a result, it is shown that the proposed method is especially effective for the paramtrical study of unstable ductile fracture. (author)

Effects of degradation on the mechanical properties and fracture toughness of a steel pressure-vessel weld metal

International Nuclear Information System (INIS)

Wu, S.J.; Knott, J.F.

2003-01-01

A degradation procedure has been devised to simulate the effect of neutron irradiation on the mechanical properties of a steel pressure-vessel weld metal. The procedure combines the application of cold prestrain together with an embrittling heat treatment to produce an increase in yield stress, a decrease in strain hardening rate, and an increased propensity for brittle intergranular fracture. Fracture tests were carried out using blunt-notch four-point-bend specimens in slow bend over a range of temperatures and the brittle/ductile transition was shown to increase by approximately 110 deg. C as a result of the degradation. Fractographic analysis of specimens broken at low temperatures showed about 30% intergranular failure in combination with transgranular cleavage. Predictions have been made of the ductile-brittle transition curves for the weld metal (sharp crack) fracture toughness in degraded and non-degraded states, based on the notched-bar test results and on finite element analyses of the stress distributions ahead of the notches and sharp cracks. The ductile-brittle transition temperature shift (ΔT=110 deg. C) between non-degraded and degraded weld metal at a notch opening displacement of 0.31 mm was combined with the Ritchie, Knott and Rice (RKR) model to predict an equivalent shift of 115 deg. C for sharp-crack specimens at a toughness level of 70 MN/m 3/2

[Distal clavicle fracture].

Science.gov (United States)

Seppel, G; Lenich, A; Imhoff, A B

2014-06-01

Reposition and fixation of unstable distal clavicle fractures with a low profile locking plate (Acumed, Hempshire, UK) in conjunction with a button/suture augmentation cerclage (DogBone/FibreTape, Arthrex, Naples, FL, USA). Unstable fractures of the distal clavicle (Jäger and Breitner IIA) in adults. Unstable fractures of the distal clavicle (Jäger and Breitner IV) in children. Distal clavicle fractures (Jäger and Breitner I, IIB or III) with marked dislocation, injury of nerves and vessels, or high functional demand. Patients in poor general condition. Fractures of the distal clavicle (Jäger and Breitner I, IIB or III) without marked dislocation or vertical instability. Local soft-tissue infection. Combination procedure: Initially the lateral part of the clavicle is exposed by a 4 cm skin incision. After reduction of the fracture, stabilization is performed with a low profile locking distal clavicle plate. Using a special guiding device, a transclavicular-transcoracoidal hole is drilled under arthroscopic view. Additional vertical stabilization is arthroscopically achieved by shuttling the DogBone/FibreTape cerclage from the lateral portal cranially through the clavicular plate. The two ends of the FibreTape cerclage are brought cranially via adjacent holes of the locking plate while the DogBone button is placed under the coracoid process. Thus, plate bridging is achieved. Finally reduction is performed and the cerclage is secured by surgical knotting. Use of an arm sling for 6 weeks. Due to the fact that the described technique is a relatively new procedure, long-term results are lacking. In the short term, patients postoperatively report high subjective satisfaction without persistent pain.

Fracture mechanics based life assessment in petrochemical plants

International Nuclear Information System (INIS)

Norasiah Ab Kasim; Abd Nassir Ibrahim; Ab Razak Hamzah; Shukri Mohd

2004-01-01

The increasing use of thick walled pressure vessels in petrochemical plants operating at high pressure under severe service conditions could lead to catastrophic failure. In the Malaysian Institute for Nuclear Technology Research (MINT), initial efforts are underway to apply fracture mechanics approach for assessment of significance of defects detected during periodic in service inspection (ISI) of industrial plants. This paper outlines the integrity management strategy based on fracture mechanics and proposes a new procedure for life assessment of petrochemical plants based on ASME Boiler and Pressure Vessel Code, Section XI, BSI PD 6493:1991, BSI 6539:1994, BSI Standard 7910:1999 and API 579:2000. Essential relevant data required for the assessment is listed. Several methods available for determination of fracture toughness are reviewed with limitations in their application to petrochemical plants. A new non destructive method for determination of fracture toughness based on hardness testing and normalized key roughness curve is given. Results of fracture mechanics based life assessment conducted for 100 mm thick ammonia converter of Ni r o steel and 70 mm thick plat forming reactor vessel of ASTM A 38 7 grade B steel in operational fertilizer and petroleum refining plants are presented. (Author)

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.

Wrist Fractures

Science.gov (United States)

... All Topics A-Z Videos Infographics Symptom Picker Anatomy Bones Joints Muscles Nerves Vessels Tendons About Hand Surgery What is a Hand Surgeon? What is a Hand Therapist? Media Find a Hand Surgeon Home Anatomy Wrist Fractures Email to a friend * required fields ...

Shoulder Fractures

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Preliminary test results from the HSST shallow-crack fracture toughness program

International Nuclear Information System (INIS)

Theiss, T.J.; Robinson, G.C.; Rolfe, S.T.

1991-01-01

The Heavy Section Steel Technology (HSST) Program under sponsorship of the Nuclear Regulatory Commission (NRC) is investigating the influence of crack depth on the fracture toughness of reactor pressure vessel steel. The ultimate goal of the investigation is the generation of a limited data base of elastic-plastic fracture toughness values appropriate for shallow flaws in a reactor pressure vessel and the application of this data to reactor vessel life assessments. It has been shown that shallow-flaws play a dominant role in the probabilistic fracture mechanics analysis of reactor pressure vessels during a pressurized-thermal-shock event. In addition, recent research has shown that the crack initiation toughness measured using specimens with shallow flaws is greater that the toughness determined with conventional, deeply notched specimens at temperatures within the transition region for non-nuclear steels. The influence of crack depth on the elastic-plastic fracture toughness for prototypic reactor material is being investigated. Preliminary results indicate a significant increase in the toughness associated with shallow-flaws which has the potential to significantly impact the conditional probability of vessel failure. 8 refs., 4 figs., 1 tab

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.)

Fracture mechanics characterisation of the WWER-440 reactor pressure vessel beltline welding seam of Greifswald unit 8

International Nuclear Information System (INIS)

Viehrig, Hans-Werner; Schuhknecht, Jan

2008-01-01

WWER-440 second generation (V-213) reactor pressure vessels (RPV) were produced by IZHORA in Russia and by SKODA in the former Czechoslovakia. The surveillance Charpy-V and fracture mechanics SE(B) specimens of both producers have different orientations. The main difference is the crack extension direction which is through the RPV thickness and circumferential for ISHORA and SKODA RPV, respectively. In particular for the investigation of weld metal from multilayer submerged welding seams the crack extension direction is of importance. Depending on the crack extension direction in the specimen there are different welding beads or a uniform structure along the crack front. The specimen orientation becomes more important when the fracture toughness of the weld metal is directly determined on surveillance specimens according to the Master Curve (MC) approach as standardised in the ASTM Standard Test Method E1921. This approach was applied on weld metal of the RPV beltline welding seam of Greifswald Unit 8 RPV. Charpy size SE(B) specimens from 13 locations equally spaced over the thickness of the welding seam were tested. The specimens are in TL and TS orientation. The fracture toughness values measured on the SE(B) specimens with both orientations follow the course of the MC. Nearly all values lie within the fracture toughness curves for 5% and 95% fracture probability. There is a strong variation of the reference temperature T 0 though the thickness of the welding seam, which can be explained with structural differences. The scatter is more pronounced for the TS SE(B) specimens. It can be shown that specimens with TS and TL orientation in the welding seam have a differentiating and integrating behaviour, respectively. The statistical assumptions behind the MC approach are valid for both specimen orientations even if the structure is not uniform along the crack front. By comparison crack extension, JR, curves measured on SE(B) specimens with TL and TS orientation show

Ultimate load analysis of prestressed concrete reactor pressure vessels considering a general material law

International Nuclear Information System (INIS)

Schimmelpfennig, K.

1975-01-01

A method of analysis is presented, by which progressive fracture processes in axisymmetric prestressed concrete pressure vessels during increasing internal pressure can be evaulated by means of a continuum calculation considering a general material law. Formulations used in the analysis concerning material behaviour are derived on one hand from appropriate results of testing small concrete specimens, and are on the other hand gained by parametric studies in order to solve questions still existing by recalulating fracture tests on concrete bodies with more complex states of stress. Due attention is focussed on investigating the behaviour of construction members subjected to high shear forces (end slabs.). (Auth.)

Application of local approach to quantitative prediction of degradation in fracture toughness of steels due to pre-straining and irradiation

International Nuclear Information System (INIS)

Miyata, T.; Tagawa, T.

1996-01-01

Degradation of cleavage fracture toughness for low carbon steels due to pre-straining and irradiation was investigated on the basis of the local fracture criterion approach. Formulation of cleavage fracture toughness through the statistical modelling proposed by BEREMIN has been simplified by the present authors to the expression involving yield stress and cleavage fracture stress of materials. A few percent pre-strain induced by cold rolling deteriorates significantly the cleavage fracture toughness. Ductile-brittle transition temperature is increased to more than 70 C higher by 8% straining in 500 MPa class high strength steel. Quantitative prediction of degradation has been successfully examined through the formulation of the cleavage fracture toughness. Analytical and experimental results indicate that degradation in toughness is caused by the increase of flow stress in pre-strained materials. Quantitative prediction of degradation of toughness due to irradiation has been also examined for the past experiments on the basis of the local fracture criterion approach. Analytical prediction from variance of yield stress by irradiation is well consistent with the experimental results. (orig.)

Generic analyses for evaluation of low Charpy upper-shelf energy effects on safety margins against fracture of reactor pressure vessel materials

International Nuclear Information System (INIS)

Dickson, T.L.

1993-07-01

Appendix G to 10 CFR Part 50 requires that reactor pressure vessel beltline material maintain Charpy upper-shelf energies of no less than 50 ft-lb during the plant operating life, unless it is demonstrated in a manner approved by the Nuclear Regulatory Commission (NRC), that lower values of Charpy upper-shelf energy provide margins of safety against fracture equivalent to those in Appendix G to Section XI of the ASME Code. Analyses based on acceptance criteria and analysis methods adopted in the ASME Code Case N-512 are described herein. Additional information on material properties was provided by the NRC, Office of Nuclear Regulatory Research, Materials Engineering Branch. These cases, specified by the NRC, represent generic applications to boiling water reactor and pressurized water reactor vessels. This report is designated as HSST Report No. 140

Humeral fractures due to low-energy trauma: an epidemiological survey in patients referred to a large emergency department in Northern Italy.

Science.gov (United States)

Pedrazzoni, M; Abbate, B; Verzicco, I; Pedrazzini, A; Benatti, M; Cervellin, G

2015-01-01

This survey describes the epidemiology of approximately 1800 low-energy humeral fractures seen in a large emergency department in Northern Italy over 7 years (2007-2013), highlighting the differences from previous Italian studies. The purpose of this study was to determine the incidence of humeral fractures due to low-energy trauma in patients 40 years of age or older referred to a large Emergency Department (Parma, Northern Italy) in a 7-year period (2007-2013). All humeral fractures referred to the emergency department of the Academic Hospital of Parma (the main hospital in the province with a catchment area of approximately 345,000) were retrieved from the hospital database using both ICD-9CM codes and text strings. The diagnosis of humeral fracture due to low-energy trauma was confirmed by medical records and X-ray reports, after exclusion of injuries due to a clear-cut high-energy trauma or cancer. The query identified 1843 humeral fractures (1809 first fractures), with a clear predominance in women (78 %). Fractures of the proximal humerus represented the large majority of humeral fractures (more than 85 %), with an incidence progressively increasing with age (more than 60-fold in women and 20-fold in men). Simultaneous fractures (hip in particular) were frequent especially after 85 years of age (1 out of 8 cases). When compared to other Italian studies, the incidence of humeral fractures was significantly lower than that derived from discharge data corrected for hospitalization rate (standardized rate ratio 0.74; p energy humeral fractures in Italy. Our results partly differ from previous Italian studies based on indirect estimations.

Fracture toughness of manet II steel

International Nuclear Information System (INIS)

Gboneim, M.M.; Munz, D.

1997-01-01

High fracture toughness was evaluated according to the astm and chromium (9-12) martensitic steels combine high strength and toughness with good corrosion and oxidation resistance in a range of environments, and also show relatively high creep strength at intermediate temperatures. They therefore find applications in, for example, the offshore oil and gas production and chemical industries i pipe work and reaction vessels, and in high temperature steam plant in power generation systems. Recently, the use of these materials in the nuclear field was considered. They are candidates as tubing materials for breeder reactor steam generators and as structural materials for the first wall and blanket in fusion reactors. The effect of ageing on the tensile properties and fracture toughness of a 12 Cr-1 Mo-Nb-v steel, MANET II, was investigated in the present work. Tensile specimens and compact tension (CT) specimens were aged at 550 degree C for 1000 h. The japanese standards. Both microstructure and fracture surface were examined using optical and scanning electron microscopy (SEM). The results showed that ageing did not affect the tensile properties. However, the fracture toughness K Ic and the tearing modules T were reduced due to the ageing treatment. The results were discussed in the light of the chemical composition and the fracture surface morphology. 9 figs., 3 tabs

Applicability of JIS SPV 50 steel to primary containment vessel of nuclear power station

International Nuclear Information System (INIS)

Iida, Kunihiro; Ishikawa, Koji; Sakai, Keiichi; Onozuka, Masakazu; Sato, Makoto.

1979-01-01

The space within reactor containment vessels must be expanded in order to improve the reliability of nuclear power plants, accordingly the adoption of large reactor containment vessels is investigated. SGV 42 and 49 steels in JIS G 3118 have been used for containment vessels so far, but stress relief annealing is required when the thickness exceeds 38 mm. The time has come when the use of thicker conventional plates without stress relieving or the use of high strength steel must be examined in detail. In this study, the tests of confirming material properties were carried out on SPV 50 in JIS G 3115, Steels for pressure vessels, aiming at the method of fabrication without stress relieving. The highest and lowest temperatures in use were set at 171 deg and -8 deg C, respectively. The chemical composition and the mechanical properties of the plates tested, the method of welding, the results of tensile test on the parent metal and the welds, the required lowest preheating temperature, the fracture toughness at low temperature and the brittle fracture causing test are reported. The parent metal and the welded joints of SPV 50 have the properties suitable to reactor containment vessels, namely the sufficient fracture toughness to guarantee the prevention of unstable fracture when the method of welding without stress relieving is adopted. (Kako, I.)

Fractures Due to Gunshot Wounds: Do Retained Bullet Fragments Affect Union?

Science.gov (United States)

Riehl, John T; Connolly, Keith; Haidukewych, George; Koval, Ken

2015-01-01

Many types of projectiles, including modern hollow point bullets, fragment into smaller pieces upon impact, particularly when striking bone. This study was performed to examine the effect on time to union with retained bullet material near a fracture site in cases of gunshot injury. All gunshot injuries operatively treated with internal fixation at a Level 1 Trauma Center between March 2008 and August 2011 were retrospectively reviewed. Retained bullet load near the fracture site was calculated based on percentage of material retained compared to the cortical diameter of the involved bone. Analyses were performed to assess the effect of the lead-cortical ratio and amount of comminution on time to fracture union. Thirty-two patients (34 fractures) met the inclusion criteria, with an equal number of comminuted (17) and non-comminuted fractures (17). Seventeen of 34 fractures (50%) united within 4 months, 16/34 (47%) developed a delayed union, and 1/34 (3%) developed a nonunion requiring revision surgery. Sixteen of 17 fractures (94%) that united by 4 months had a cumulative amount of bullet fragmentation retained near the fracture site of less than 20% of the cortical diameter. Nine out of 10 fractures (90%) with retained fragments near the fracture site was equal to or exceeding 20% of the cortical diameter had delayed or nonunion. Fracture comminution had no effect on time to union. The quantity of retained bullet material near the fracture site was more predictive of the rate of fracture union than was comminution. Fractures with bullet fragmentation equal to or exceeding 20% of the cortical width demonstrated a significantly higher rate of delayed union/nonunion compared to those fractures with less retained bullet material, which may indicate a local cytotoxic effect from lead on bone healing. These findings may influence decisions on timing of secondary surgeries. Level III.

Some advances in fracture studies under the heavy-section steel technology program

International Nuclear Information System (INIS)

Pugh, C.E.; Corwin, W.R.; Bryan, R.H.; Bass, B.R.

1985-01-01

Recent results are summarized from HSST studies in three major areas that relate to assessing nuclear reactor pressure vessel integrity under pressurized-thermal-shock (PTS) conditions: irradiation effects on the fracture properties of stainless steel cladding, crack run-arrest behavior under nonisothermal conditions, and fracture behavior of a thick-wall vessel under combined thermal and pressure loadings

Geriatric Trauma Patients With Cervical Spine Fractures due to Ground Level Fall: Five Years Experience in a Level One Trauma Center.

Science.gov (United States)

Wang, Hao; Coppola, Marco; Robinson, Richard D; Scribner, James T; Vithalani, Veer; de Moor, Carrie E; Gandhi, Raj R; Burton, Mandy; Delaney, Kathleen A

2013-04-01

It has been found that significantly different clinical outcomes occur in trauma patients with different mechanisms of injury. Ground level falls (GLF) are usually considered "minor trauma" with less injury occurred in general. However, it is not uncommon that geriatric trauma patients sustain cervical spine (C-spine) fractures with other associated injuries due to GLF or less. The aim of this study is to determine the injury patterns and the roles of clinical risk factors in these geriatric trauma patients. Data were reviewed from the institutional trauma registry of our local level 1 trauma center. All patients had sustained C-spine fracture(s). Basic clinical characteristics, the distribution of C-spine fracture(s), and mechanism of injury in geriatric patients (65 years or older) were compared with those less than 65 years old. Furthermore, different clinical variables including age, gender, Glasgow coma scale (GCS), blood alcohol level, and co-existing injuries were analyzed by multivariate logistic regression in geriatric trauma patients due to GLF and internally validated by random bootstrapping technique. From 2006 - 2010, a total of 12,805 trauma patients were included in trauma registry, of which 726 (5.67%) had sustained C-spine fracture(s). Among all C-spine fracture patients, 19.15% (139/726) were geriatric patients. Of these geriatric patients 27.34% (38/139) and 53.96% (75/139) had C1 and C2 fractures compared with 13.63% (80/587) and 21.98% (129/587) in young trauma patients (P geriatric trauma patients 13.67% (19/139) and 18.71% (26/139) had C6 and C7 fractures compared with 32.03% (188/587) and 41.40% (243/587) in younger ones separately (P geriatric patients had sustained C-spine fractures due to GLF with more upper C-spine fractures (C1 and C2). Only 3.2% of those had positive blood alcohol levels compared with 52.9% of younger patients (P geriatric patients due to GLF had intracranial pathology (ICP) which was one of the most common co

Fracture toughness of welded joints of ASTM A543 steel plate

International Nuclear Information System (INIS)

Susukida, H.; Uebayashi, T.; Yoshida, K.; Ando, Y.

1977-01-01

Fracture toughness and weldability tests have been performed on a high strength steel which is a modification of ASTM A543 Grade B Class 1 steel, with a view to using it for nuclear reactor containment vessels. The results showed that fracture toughness of welded joints of ASTM A543 modified high strength steel is superior and the steel is suitable for manufacturing the containment vessels

Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-1 and TSE-2

International Nuclear Information System (INIS)

Cheverton, R.D.

1976-09-01

The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and two thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. The PWR calculations indicated that under some circumstances crack propagation could be expected and that experiments should be conducted for cracks that would have the potential for propagation at least halfway through the wall

Planary bone scintigraphy of the foot in the diabetics with peripheral macroangiopathy, treated with Sulodexide (Vessel Due F)

International Nuclear Information System (INIS)

Klisarova, A.; Bihchelian, H.; Koeva, L.

2000-01-01

It is the purpose of the paper to assay the effect of Sulodexide (Vessel Due F) on some semi-quantitative parameters of bone scintigraphy of the foot in diabetics with peripheral vascular disease. Fifteen patients with diabet type II and peripheral macroangiopathy (6 women and 9 men, mean age 57 ±3.7 y, and mean body mass index 29.8 ±2.7 kg/m 3 ) are studied. The investigation is carried out after informed consent and in state of adequate glycemic control. Vessel Due F is administrated im for 10 days - 600 lypoproteinli pase releasing units (LSU), followed by 500 LSU per os for further 60 days. Before and after treatment, planar foot bone scintigraphy with 99m Tc - MDP is carried out. The fixation indices are measured through comparison of radionuclide accumulation in symmetrical zones of the right and left foot. Prior to treatment the fixation indices are increased, and after treatment they are significantly decreased (p< 0.05). The semiquantitative scintigraphic parameters adopted are used for dynamic measurement of the bone metabolism level in the feet of patients with diabet and peripheral macroangiopathy. After Sulodexide (Vessel Due F) treatment, a positive effect on radionuclide fixation indices is documented. (author)

L5 radiculopathy due to sacral stress fracture

International Nuclear Information System (INIS)

Aylwin, Anthony; Saifuddin, Asif; Tucker, Stuart

2003-01-01

We report the case of a 70-year-old man who presented with a history of left buttock pain with radiation into the left leg in an L5 distribution. MRI of the lumbar spine revealed a left sacral stress fracture with periosteal reaction involving the left L5 nerve root anterior to the sacral ala. With spontaneous healing of the fracture, the patient's symptoms resolved completely. (orig.)

Structural Properties of EB-Welded AlSi10Mg Thin-Walled Pressure Vessels Produced by AM-SLM Technology

Science.gov (United States)

Nahmany, Moshe; Stern, Adin; Aghion, Eli; Frage, Nachum

2017-10-01

Additive manufacturing of metals by selective laser melting (AM-SLM) is hampered by significant limitations in product size due to the limited dimensions of printing trays. Electron beam welding (EBW) is a well-established process that results in relatively minor metallurgical modifications in workpieces due to the ability of EBW to pass high-density energy to the related substance. The present study aims to evaluate structural properties of EB-welded AlSi10Mg thin-walled pressure vessels produced from components prepared by SLM technology. Following the EB welding process, leak and burst tests were conducted, as was fractography analysis. The welded vessels showed an acceptable holding pressure of 30 MPa, with a reasonable residual deformation up to 2.3% and a leak rate better than 1 × 10-8 std-cc s-1 helium. The failures that occurred under longitudinal stresses reflected the presence of two weak locations in the vessels, i.e., the welded joint region and the transition zone between the vessel base and wall. Fractographic analysis of the fracture surfaces of broken vessels displayed the ductile mode of the rupture, with dimples of various sizes, depending on the failure location.

A Rare Entity: Bilateral First Rib Fractures Accompanying Bilateral Scapular Fractures

OpenAIRE

Gulbahar, Gultekin; Kaplan, Tevfik; Turker, Hasan Bozkurt; Gundogdu, Ahmet Gokhan; Han, Serdar

2015-01-01

First rib fractures are scarce due to their well-protected anatomic locations. Bilateral first rib fractures accompanying bilateral scapular fractures are very rare, although they may be together with scapular and clavicular fractures. According to our knowledge, no case of bilateral first rib fractures accompanying bilateral scapular fractures has been reported, so we herein discussed the diagnosis, treatment, and complications of bone fractures due to thoracic trauma in bias of this rare en...

A Rare Entity: Bilateral First Rib Fractures Accompanying Bilateral Scapular Fractures.

Science.gov (United States)

Gulbahar, Gultekin; Kaplan, Tevfik; Turker, Hasan Bozkurt; Gundogdu, Ahmet Gokhan; Han, Serdar

2015-01-01

First rib fractures are scarce due to their well-protected anatomic locations. Bilateral first rib fractures accompanying bilateral scapular fractures are very rare, although they may be together with scapular and clavicular fractures. According to our knowledge, no case of bilateral first rib fractures accompanying bilateral scapular fractures has been reported, so we herein discussed the diagnosis, treatment, and complications of bone fractures due to thoracic trauma in bias of this rare entity.

A Rare Entity: Bilateral First Rib Fractures Accompanying Bilateral Scapular Fractures

Directory of Open Access Journals (Sweden)

Gultekin Gulbahar

2015-01-01

Full Text Available First rib fractures are scarce due to their well-protected anatomic locations. Bilateral first rib fractures accompanying bilateral scapular fractures are very rare, although they may be together with scapular and clavicular fractures. According to our knowledge, no case of bilateral first rib fractures accompanying bilateral scapular fractures has been reported, so we herein discussed the diagnosis, treatment, and complications of bone fractures due to thoracic trauma in bias of this rare entity.

Different approaches to estimation of reactor pressure vessel material embrittlement

Directory of Open Access Journals (Sweden)

V. M. Revka

2013-03-01

Full Text Available The surveillance test data for the nuclear power plant which is under operation in Ukraine have been used to estimate WWER-1000 reactor pressure vessel (RPV material embrittlement. The beltline materials (base and weld metal were characterized using Charpy impact and fracture toughness test methods. The fracture toughness test data were analyzed according to the standard ASTM 1921-05. The pre-cracked Charpy specimens were tested to estimate a shift of reference temperature T0 due to neutron irradiation. The maximum shift of reference temperature T0 is 84 °C. A radiation embrittlement rate AF for the RPV material was estimated using fracture toughness test data. In addition the AF factor based on the Charpy curve shift (ΔTF has been evaluated. A comparison of the AF values estimated according to different approaches has shown there is a good agreement between the radiation shift of Charpy impact and fracture toughness curves for weld metal with high nickel content (1,88 % wt. Therefore Charpy impact test data can be successfully applied to estimate the fracture toughness curve shift and therefore embrittlement rate. Furthermore it was revealed that radiation embrittlement rate for weld metal is higher than predicted by a design relationship. The enhanced embrittlement is most probably related to simultaneously high nickel and high manganese content in weld metal.

Low cycle fatigue behavior of a ferritic reactor pressure vessel steel

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-15

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

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

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

Safety vessels for explosive fusion reactor

International Nuclear Information System (INIS)

Mineev, V.

1994-01-01

The failure of several types of geometrically similar cylindrical and spherical steel and glass fibers vessels filled with water or air was investigated when an explosive charge of TNT was detonated in the center. Vessels had radius 50-1000 mm, thickness of walls 2-20%. The detonation on TNT imitated energy release. The parameter: K = M/mf is a measure of the strength of the vessel where M is the mass of the vessel, and mf is the mass of TNT for which the vessel fails. This demanded 2-4 destroyed and nondestroyed shots. It may be showed that: K=A/σ f where σ f is the fracture stress of the material vessel, and A = const = F(energy TNT, characteristic of elasticity of vessel material). The chief results are the following: (1) A similar increase in the geometrical dimensions of steel vessels by a factor of 10 leads to the increase of parameter K in about 5 times and to decrease of failure deformation in 7 times (scale effect). (2) For glass fibers, scale effect is absent. (3) This problem is solved in terms of theory energetic scale effect. (4) The concept of TNT equivalent explosive makes it possible to use these investigations to evaluate the response of safety vessels for explosive fusion reactor

Scintigraphic follow-up of fracture healing in animals

International Nuclear Information System (INIS)

Klug, W.; Franke, W.G.; Schulze, M.

1983-01-01

Secondary bone fracture heating was analysed by scintigraphic follow-up studies in rabbits using sup(99m)Tc-HEDP. 24 hours after fracture the activity ratio between the fractured and the non-fractured lower limb was 2,2. The maximal count density in the fracture region is found during the 14th and 28th day after fracture. Concomitantly there is a significant increase of bone marrow vessels and content of copper, magnesium, sodium and water in the callus. Although roentgenographic controls and static investigations with respect to consolidation reveal a complete heating already 126 days after fracture, the complete scintigraphic normalisation of the lower limb fracture of the rabbit is found not earlier than at the 203rd day after fracture. (orig.) [de

Scintigraphic follow-up of fracture healing in animals

Energy Technology Data Exchange (ETDEWEB)

Klug, W.; Franke, W.G.; Schulze, M.

1983-08-01

Secondary bone fracture healing was analysed by scintigraphic follow-up studies in rabbits using sup(99m)Tc-HEDP. 24 hours after fracture the activity ratio between the fractured and the non-fractured lower limb was 2,2. The maximal count density in the fracture region is found during the 14th and 28th day after fracture. Concomitantly there is a significant increase of bone marrow vessels and content of copper, magnesium, sodium and water in the callus. Although roentgenographic controls and static investigations with respect to consolidation reveal a complete healing already 126 days after fracture, the complete scintigraphic normalisation of the lower limb fracture of the rabbit is found not earlier than at the 203rd day after fracture.

Preliminary fracture analysis on the integrity of HSST intermediate test vessels

International Nuclear Information System (INIS)

Zahoor, A.; Paris, P.C.

1980-01-01

Whenever pressure in the vessels is such that the vessel has not yielded the indication is for stable crack growth. With higher pressures which cause full thickness yielding there is unstable crack growth

The irradiation embrittlement of two pressure vessel steels -Contribution of local approach

Energy Technology Data Exchange (ETDEWEB)

Soulat, P; Marini, B [CEA Centre d` Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France). Service de Recherches Metallurgiques Appliquees; Miannay, D; Horowitz, H [CEA Centre d` Etudes de Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire; Schill, R [CEA Centre d` Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie

1994-12-31

Within the IAEA Coordinated Research Programme on ``Optimizing the Reactor Pressure Vessel Surveillance Programmes and their Analyses``, the French participation has been focused on the contribution of the local approach to the determination of the sensitivity to radiation embrittlement of two different pressure vessel steels: a low sensitive French forging steel (FFA) and a high sensitive ``monitor`` Japanese plate steel (JRQ) were irradiated to a fluence of 3.10{sup 19} n/cm{sup 2} at 290 C. The irradiation embrittlement of the two steels measured by the shift of Charpy V transition curves is in good agreement with the estimated shifts given by theoretical prediction. The fracture toughness properties were examined at low temperature with brittle fracture, and at service temperature (290 C), with ductile tearing. The values of K{sub 1C} or K{sub JC} for the brittle fracture and J{sub 1C} for the ductile fracture are compared to predictions established using the local approach of cleavage fracture (Weibull analysis) and the critical rate of void growth respectively. 8 refs., 14 figs., 10 tabs.

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

Evaluation of fatigue damage of pressure vessel materials by observation of microstructures

International Nuclear Information System (INIS)

Yoshida, Kazuo

1994-01-01

As the important factor as the secular change mode of pressure vessel materials, there is fatigue damage. In USA, there is the move to use LWRs by extending their life, and it becomes necessary to show the soundness of the structures of machinery and equipment for long period. For exactly evaluating the soundness of the structures of machinery and equipment, it is important to clarify the degree of secular deterioration of the materials. In this report, by limiting to the fatigue damage of LWR pressure vessel steel, the method of grasping the change of microstructure and the method of estimating the degree of fatigue damage from the change of microstructure are shown. The change of microstructure arising in materials due to fatigue advances in the following steps, namely, the multiplication of dislocations, the tangling of dislocations, the formation of cell structure, the turning of cells, the formation of microcracks, the growth of cracks and fracture. In the case of pressure vessel steel, due to the quenching and tempering, the cell structure is formed from the beginning, and the advance of fatigue is recognized as the increase of the turning angle of cell structures. The detection of fatigue damage by microstructure is reported. (K.I.)

Fracture mechanics assessment of surface and sub-surface cracks in the RPV under non-symmetric PTS loading

Energy Technology Data Exchange (ETDEWEB)

Keim, E; Shoepper, A; Fricke, S [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

1997-09-01

One of the most severe loading conditions of a reactor pressure vessel (rpv) under operation is the loss of coolant accident (LOCA) condition. Cold water is injected through nozzles in the downcomer of the rpv, while the internal pressure may remain at a high level. Complex thermal hydraulic situations occur and the fluid and downcomer temperatures as well as the fluid to wall heat transfer coefficient at the inner surface are highly non-linear. Due to this non-symmetric conditions, the problem is investigated by three-dimensional non-linear finite element analyses, which allow for an accurate assessment of the postulated flaws. Transient heat transfer analyses are carried out to analyze the effect of non-symmetrical cooling of the inner surface of the pressure vessel. In a following uncoupled stress analysis the thermal shock effects for different types of defects, surface flaws and sub-surface flaws are investigated for linear elastic and elastic-plastic material behaviour. The obtained fracture parameters are calculated along the crack fronts. By a fast fracture analysis the fracture parameters at different positions along the crack front are compared to the material resistance. Safety margins are pointed out in an assessment diagram of the fracture parameters and the fracture resistance versus the transient temperature at the crack tip position. (author). 4 refs, 10 figs.

Prevention of catastrophic failure in pressure vessels and pipings

International Nuclear Information System (INIS)

Rintamaa, R.; Wallin, K.; Ikonen, K.; Toerroenen, K.; Talja, H.; Keinaenen, H.; Saarenheimo, A.; Nilsson, F.; Sarkimo, M.; Waestberg, S.; Debel, C.

1989-01-01

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

ORNL probabilistic fracture-mechanics code OCA-P

International Nuclear Information System (INIS)

Cheverton, R.D.; Ball, D.G.

1984-01-01

The computer code OCA-P was developed at the request of the USNRC for the purpose of helping to evaluate the integrity of PWR pressure vessels during overcooling accidents (OCA's). The code can be used for both deterministic and probabilistic fracture-mechanics calculations, and consists essentially of OCA-II and a Monte Carlo routine similar to that developed by Strosnider et al. In the probabilistic mode OCA-P generates a large number of vessels (10 6 more or less), each with a different combination of the various values of the different parameters involved in the analysis of flaw behavior. For each of these vessels a deterministic fracture-mechanics analysis is performed (calculation of K/sub I/, K/sub Ic/, K/sub Ia/) to determine whether vessel failure takes place. The conditional probability of failure is simply the number of vessels that fail divided by the number of vessels generated. OCA-II is used for the deterministic analysis. Basic input to OCA-II includes, among other things, the primry-system pressure transient and the temperature transient for the coolant in the reactor-vessel downcomer. With this and additional information available OCA-II performs a one-dimensional thermal analysis to obtain the temperature distribution in the wall as a function of time and then a one-dimensional linear-elastic stress analysis. OCA-P has been checked against similar codes and is presently being used in the Integrated Pressurized Thermal Shock Program for specific PWR plants

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

International Nuclear Information System (INIS)

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

1995-01-01

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

BILATERAL ANTERIOR DISLOCATION OF SHOULDER WITH GREATER TUBEROSITY FRACTURE DUE TO HYPONATREMIA : A RARE PRESENTATION

Directory of Open Access Journals (Sweden)

Sivananda

2015-01-01

Full Text Available We here report a rare presentation of bilateral anterior dislocation of shoulder with associated fracture of greater tuberosity in a 38 year old male due to minor trauma which he sustained secondary to hyponatremia induced irritability. There was no associ ated rotator cuff tear which is often associated with BADS which makes this presentation unique. Unilateral dislocation of shoulder is a common condition which is frequently encountered in emergency trauma department. Anterior dislocation is more common th an posterior dislocation. However, simultaneous bilateral shoulder dislocations are usually posterior. Bilateral anterior dislocations with fractures of the greater tuberosity are even rarer and are usually associated with trauma or seizures

First spinning cylinder test analysis by using local approach to fracture

International Nuclear Information System (INIS)

Eripret, C.; Rousselier, G.

1993-01-01

In recent years, several experimental programs on large scale specimens were organized to evaluate capabilities of the fracture mechanics concepts employed in structural integrity assessment of PWR pressure vessels. During the first spinning cylinder test, a geometry effect was experimentally pointed out and exhibited the problem of transferability of toughness data from small scale to large scale specimens. An original analysis of this test, by means of local approach to fracture is presented in this paper. Both compact tension specimen and spinning cylinder fracture behaviour were computed by using a continuum damage mechanics model developed at EDF. The authors confirmed by numerical analysis that the cylinder's resistance to ductile tearing was considerably larger than in small scale fracture mechanics specimens tests, about 50 percent. The final crack growth predicted by the model was close to the experimental value. Discrepancies in J-R curves seemed to be due to an effect of stress triaxiality and plastic zone evolution. The geometry effect inducing differences in resistance to ductile tearing of the material involved in the specimens can be investigated and explained by using local approach to fracture methodology. 14 refs., 9 figs., 2 tabs

Hydrogen induced plastic damage in pressure vessel steel of 2.25Cr-1Mo

International Nuclear Information System (INIS)

Han, G.W.; Song, Y.J.

1995-01-01

2.25Cr-1Mo steel is generally employed as a hydrogenation reaction vessel material used at elevated temperature and in a hydrogen containing environment. During service of the reaction vessel, a large number of hydrogen atoms would enter its wall. When the reaction vessel is shutdown and the temperature reduces to about ambient temperature, the hydrogen atoms remaining in the wall would induce plastic damage in the steel. The mechanism of hydrogen induced plastic damage is different for various materials with different microstructures. Investigations have demonstrated that the hydrogen induced plastic damage in carbide annealed carbon steels is caused by hydrogen accelerating the initiating and growing of microvoids from the carbide particles. However, SEM examination on the fracture surface of hydrogen charged tensile specimen of 2.25Cr-1Mo steel show that a large number of fisheyes appear on the fracture surface. This indicates that hydrogen induced plastic damage in 2.25Cr-1Mo steel is related to the occurrence of fisheye cracks during plastic deformation. By means of micro-fracture mechanics to analyze fisheye crack occurrence from the first generation microvoid, the mechanism of hydrogen induced plastic damage in the pressure vessel steel is investigated

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

Fracture mechanics as judgement criterion in reference publications

International Nuclear Information System (INIS)

Bartholome, G.

1976-01-01

Fracture mechanics is applied in particular in ship and aeroplane construction, in astronautics, and in nuclear engineering. Around 1950, the high quality demands in nuclear engineering led to the first regulation for brittle-fracture-safe operation of thick-walled nuclear pressure vessels. These regulations are based on the brittle-fracture-plan (NDT concept). For reactor engineering this plan is applied in a simplified way, the so-called modified PORSE-diagram. The permissible operational stresses must be out of the range of brittle fracture margin which is defined by the NDT temperature extension limit. (RW) [de

Development of Mini-Compact Tension Test Method for Determining Fracture Toughness Master Curves for Reactor Pressure Vessel Steels

Energy Technology Data Exchange (ETDEWEB)

Sokolov, Mikhail A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-05-01

Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Validation of the mini compact tension specimen (mini-CT) geometry has been performed on previously well characterized Midland beltline Linde 80 (WF-70) weld in the unirradiated condition. It was shown that the fracture toughness transition temperature, To, measured by these Mini-CT specimens is almost the same as To value that was derived from various larger fracture toughness specimens. Moreover, an International collaborative program has been established to extend the assessment and validation efforts to irradiated Linde 80 weld metal. The program is underway and involves the Oak Ridge National Laboratory (ORNL), Central Research Institute for Electrical Power Industry (CRIEPI), and Electric Power Research Institute (EPRI). The irradiated Mini-CT specimens from broken halves of previously tested Charpy

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

Intergranular fracture stress and phosphorus grain boundary segregation of a Mn-Ni-Mo steel

International Nuclear Information System (INIS)

Naudin, C.; Frund, J.M.; Pineau, A.

1999-01-01

Nuclear Reactor Pressure Vessel (RPV) steel A508 class 3 which is a low alloyed steel is not usually sensitive to reversible temper embrittlement when properly heat treated. However heterogeneous zones may be present in particular near the inner side of the vessel. These zones result from the segregation of the alloying elements (C, Mn, Ni, Mo) and impurities (S, P) taking place during solidification of the material. They are called segregated zones (or ghost lines). They can reach 2 mm thick along the radius and 30 mm long through the circumferential direction. Their susceptibility to reversible temper embrittlement is mainly due to grain boundary phosphorus segregation triggering brittle intergranular fracture when the material is tested at low temperature. In this material like in other steels the influence of some other alloying elements (Mo, Mn...) is clearly significant and should also be taken into account. But phosphorus effect has proved to be predominant. The aim of the present study is therefore to find out a quantitative relationship between grain boundary phosphorus segregation and critical intergranular fracture stress. A synthetic steel with a chemical composition representative of an average segregated zone was prepared for the present study. A number of heat treatments were applied to reach different embrittlement conditions. Then brittle fracture properties were obtained by performing cryogenic fracture tests on notched tensile specimens while the corresponding grain boundary phosphorus levels were measured by Auger electron spectroscopy. Systematic fractographic observations were carried out. Moreover an attempt to determine the influence of temperature on the critical intergranular fracture stress was made

Potential impact of enhanced fracture-toughness data on pressurized-thermal-shock analysis

International Nuclear Information System (INIS)

Dickson, T.L.; Theiss, T.J.

1990-01-01

The Heavy Section Steel Technology (HSST) Program is involved with the generation of ''enhanced'' fracture-initiation toughness and fracture-arrest toughness data of prototypic nuclear reactor vessel steels. These two sets of data are enhanced because they have distinguishing characteristics that could potentially impact PWR pressure vessel integrity assessments for the pressurized-thermal shock (PTS) loading condition which is a major plant-life extension issue to be confronted in the 1990's. Currently, the HSST Program is planning experiments to verify and quantify, for A533B steel, the distinguishing characteristic of elevated initiation-fracture toughness for shallow flaws which has been observed for other steels. Deterministic and probabilistic fracture-mechanics analyses were performed to examine the influence of the enhanced initiation and arrest fracture-toughness data on the cleavage fracture response of a nuclear reactor pressure vessel subjected to PTS loading. The results of the analyses indicated that application of the enhanced K Ia data does reduce the conditional probability of failure P(F|E); however, it does not appear to have the potential to significantly impact the results of PTS analyses. The application of enhanced fracture-initiation-toughness data for shallow flaws also reduces P(F|E), but it does appear to have a potential for significantly affecting the results of PTS analyses. The effect of including Type I warm prestress in probabilistic fracture-mechanics analyses is beneficial. The benefit is transient dependent and, in some cases, can be quite significant. 19 refs., 12 figs., 1 tab

Evaluation of HFIR [High Flux Isotope Reactor] pressure-vessel integrity considering radiation embrittlement

International Nuclear Information System (INIS)

Cheverton, R.D.; Merkle, J.G.; Nanstad, R.K.

1988-04-01

The High Flux Isotope Reactor (HFIR) pressure vessel has been in service for 20 years, and during this time, radiation damage was monitored with a vessel-material surveillance program. In mid-November 1986, data from this program indicated that the radiation-induced reduction in fracture toughness was greater than expected. As a result, a reevaluation of vessel integrity was undertaken. Updated methods of fracture-mechanics analysis were applied, and an accelerated irradiations program was conducted using the Oak Ridge Research Reactor. Results of these efforts indicate that (1) the vessel life can be extended 10 years if the reactor power level is reduced 15% and if the vessel is subjected to a hydrostatic proof test each year; (2) during the 10-year life extension, significant radiation damage will be limited to a rather small area around the beam tubes; and (3) the greater-than-expected damage rate is the result of the very low neutron flux in the HFIR vessel relative to that in samples of material irradiated in materials-testing reactors (a factor of ∼10 4 less), that is, a rate effect

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

Methodology for plastic fracture - a progress report

International Nuclear Information System (INIS)

Wilkinson, J.P.D.; Smith, R.E.E.

1977-01-01

This paper describes the progress of a study to develop a methodology for plastic fracture. Such a fracture mechanics methodology, having application in the plastic region, is required to assess the margin of safety inherent in nuclear reactor pressure vessels. The initiation and growth of flaws in pressure vessels under overload conditions is distinguished by a number of unique features, such as large scale yielding, three-dimensional structural and flaw configurations, and failure instabilities that may be controlled by either toughness or plastic flow. In order to develop a broadly applicable methodology of plastic fracture, these features require the following analytical and experimental studies: development of criteria for crack initiation and growth under large scale yielding; the use of the finite element method to describe elastic-plastic behaviour of both the structure and the crack tip region; and extensive experimental studies on laboratory scale and large scale specimens, which attempt to reproduce the pertinent plastic flow and crack growth phenomena. This discussion centers on progress to date on the selection, through analysis and laboratory experiments, of viable criteria for crack initiation and growth during plastic fracture. (Auth.)

Normalizing treatment influence on the forged steel SAE 8620 fracture properties

Directory of Open Access Journals (Sweden)

Paulo de Tarso Vida Gomes

2005-03-01

Full Text Available In a PWR nuclear power plant, the reactor pressure vessel (RPV contains the fuel assemblies and reactor vessels internals and keeps the coolant at high temperature and high pressure during normal operation. The RPV integrity must be assured all along its useful life to protect the general public against a significant radiation liberation damage. One of the critical issues relative to the VPR structural integrity refers to the pressurized thermal shock (PTS accident evaluation. To better understand the effects of this kind of event, a PTS experiment has been planned using an RPV prototype. The RPV material fracture behavior characterization in the ductile-brittle transition region represents one of the most important aspects of the structural assessment process of RPV's under PTS. This work presents the results of fracture toughness tests carried out to characterize the RPV prototype material behavior. The test data includes Charpy energy curves, T0 reference temperatures for definition of master curves, and fracture surfaces observed in electronic microscope. The results are given for the vessel steel in the "as received" and normalized conditions. This way, the influence of the normalizing treatment on the fracture properties of the steel could be evaluated.

Hydrogen fracture toughness tester completion

Energy Technology Data Exchange (ETDEWEB)

Morgan, Michael J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-09-30

The Hydrogen Fracture Toughness Tester (HFTT) is a mechanical testing machine designed for conducting fracture mechanics tests on materials in high-pressure hydrogen gas. The tester is needed for evaluating the effects of hydrogen on the cracking properties of tritium reservoir materials. It consists of an Instron Model 8862 Electromechanical Test Frame; an Autoclave Engineering Pressure Vessel, an Electric Potential Drop Crack Length Measurement System, associated computer control and data acquisition systems, and a high-pressure hydrogen gas manifold and handling system.

Fatigue and fracture behavior of low alloy ferritic forged steels

International Nuclear Information System (INIS)

Chaudhry, V.; Sharma, A.K.; Muktibodh, U.C.; Borwankar, Neeraj; Singh, D.K.; Srinivasan, K.N.; Kulkarni, R.G.

2016-01-01

Low alloy ferritic steels are widely used in nuclear industry for the construction of pressure vessels. Pressure vessel forged low alloy steels 20MnMoNi55 (modified) have been developed indigenously. Experiments have been carried out to study the Low Cycle Fatigue (LCF) and fracture behavior of these forged steels. Fully reversed strain controlled LCF testing at room temperature and at 350 °C has been carried out at a constant strain rate, and for different axial strain amplitude levels. LCF material behavior has been studied from cyclic stress-strain responses and the strain-life relationships. Fracture behavior of the steel has been studied based on tests carried out for crack growth rate and fracture toughness (J-R curve). Further, responses of fatigue crack growth rate tests have been compared with the rate evaluated from fatigue precracking carried out for fracture toughness (J-R) tests. Fractography of the samples have been carried out to reveal dominant damage mechanisms in crack propagation and fracture. The fatigue and fracture properties of indigenously developed low alloy steel 20MnMoNi55 (modified) steels are comparable with similar class of steels. (author)

Fracture mechanics analysis of reactor pressure vessel under pressurized thermal shock - The effect of elastic-plastic behavior and stainless steel cladding -

International Nuclear Information System (INIS)

Joo, Jae Hwang; Kang, Ki Ju; Jhung, Myung Jo

2002-01-01

Performed here is an assessment study for deterministic fracture mechanics analysis of a pressurized thermal shock (PTS). The PTS event means an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. The problems consisting of two transients and 10 cracks are solved and maximum stress intensity factors and maximum allowable nil-ductility reference temperatures are calculated. Their results are compared each other to address the general characteristics between transients, crack types and analysis methods. The effects of elastic-plastic material behavior and clad coating on the inner surface are explored

Effect of nickel content on mechanical properties and fracture toughness of weld metal of WWER-1000 reactor vessel welded joints

International Nuclear Information System (INIS)

Zubchenko, A.S.; Vasilchenko, G.S.; Starchenko, E.G.; Nosov, S.I.

2004-01-01

Welding of WWER-1000 reactor vessel of steel 15X2HMPHIA is performed using the C B -12X2H2MAA wire and PHI-16 or PHI-16A flux. Nickel content in the weld metal usually lays within the limits 1.2-1.9%. The experimental data is shown on the weld metal with the nickel contents 1.28-2.45% after irradiation with fluence up to 260.10 22 n/m 2 at energy more than 0.5 MEV. The embrittlement was measured by shift of critical brittleness temperature. Has appeared, that the weld metal with the low nickel content is the least responsive to irradiation embrittlement. The mechanical properties and fracture toughness of the weld metal with the contents of a nickel less than 1.3% are studied. Specimens CT-1T are tested, the 'master-curve', and its confidence bounds with probability of destruction 5 and 95% is built. 'Master-curve' in the specified confidence interval is affirmed by CT-4T specimens test data. Is shown, that the mechanical properties and fracture toughness of the weld metal with the contents of nickel less than 1.3% satisfy the normative requirements

Bilateral Avascular Necrosis and Pelvic Insufficiency Fractures Developing after Pelvic Radiotherapy in a Patient with Prostate Cancer: A Case Report

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Hidayet Sarı

2016-12-01

Full Text Available Prostate cancer is one of the most common malignancies in men. Pelvic radiotherapy is commonly used in both radical and palliative treatment for prostate cancer. Radiation-induced adverse effects might be seen on adjacent healthy tissues (such as vessels, bones and soft tissues with the exception of targeted area. Particularly several years after radiotherapy, low back and hip pain may occur due to bone edema, necrosis or fractures. In these cases, whether complaints due to the degenerative, metastatic or radiotherapy complications must be examined and appropriate treatment should be arranged. For this purpose, we present our elderly patient who received radiotherapy for prostate cancer, and thereafter, developed bilateral avascular hip necrosis and pelvic insufficiency fractures.

Fracture toughness evaluation in the transition region of reactor pressure vessel steel

International Nuclear Information System (INIS)

Onizawa, K.; Suzuki, M.

1995-01-01

The fracture toughness (K jc and Jc) values at the cleavage fracture initiation in the transition region of a RPV steel were investigated using mainly precracked Charpy specimens. A conventional statistical approach and a fractographic study were applied to analyze the scatter of the fracture toughness values from precracked Charpy specimens. The material used was an ASTM A533B class 1 steel, which was designated as an IAEA correlation monitor material, JRQ. A lower bound transition curve of the fracture toughness for unirradiated condition was determined by the 5% confidence limit from the Weibull and fractographic analyses. The lower bound transition curve after irradiation was evaluated based on the statistics of unirradiated specimens. The results indicated that the shift of the fracture toughness transition curbe were somewhat larger than the Charpy 41J transition temperature. The parameters to determine the lower bound toughness such as the Weibull slope and the amount of ductile crack growth are discussed. The results are also compared with a model based on weakest link theory. (author). 12 refs, 12 figs, 5 tabs

Stability-based classification for ankle fracture management and the syndesmosis injury in ankle fractures due to a supination external rotation mechanism of injury.

Science.gov (United States)

Pakarinen, Harri

2012-12-01

sensitivity and specificity of both clinical tests were calculated using the standard 7.5-Nm external rotation stress test as reference. Outcome was assessed after a minimum of one year of follow-up. Olerud-Molander (OM) scoring system, RAND 36-Item Health Survey, and VAS to measure pain and function were used as outcome measures in all studies. In study 1, 85 (53%) fractures were treated operatively using the stability based fracture classification. Non-operatively treated patients reported less pain and better OM (good or excellent 89% vs. 71%) and VAS functional scores compared to operatively treated patients although they experienced more displacement of the distal fibula (0 mm 30% vs. 69%; 0-2 mm 65% vs. 25%) after treatment. No non-operatively treated patients required operative fracture fixation during follow-up. In study 2, AITFL exploration and suture lead to equal functional outcome (OM mean, 77 vs. 73) to no exploration or fixation. In study 3, the hook test had a sensitivity of 0.25 and a specificity of 0.98. The external rotation stress test had a sensitivity of 0.58 and a specificity of 0.9. Both tests had excellent interobserver reliability; the agreement was 99% for the hook test and 98% for the stress test. There was no statistically significant difference in functional scores (OM mean, 79.6 vs. 83.6) or pain between syndesmosis transfixation and no fixation groups (Study 4). Our results suggest that a simple stability-based fracture classification is useful in choosing between non-operative and operative treatment of ankle fractures; approximately half of the ankle fractures can be treated non-operatively with success. Our observations also suggest that relevant syndesmosis injuries are rare in ankle fractures due to an SER mechanism of injury. According to our research, syndesmotic repair or fixation in SER ankle fracture has no influence on functional outcome or pain after minimum one year compared with no fixation.

Initial evaluation of ultrasonic attenuation measurements for estimating fracture toughness of RPV steels

Energy Technology Data Exchange (ETDEWEB)

Hiser, A.L. Jr.; Green, R.E. Jr. [Johns Hopkins Univ., Baltimore, MD (United States). Center for Nondestructive Evaluation

1999-08-01

Neutron bombardment of reactor pressure vessel (RPV) steels causes reductions in fracture toughness in these steels, termed neutron irradiation embrittlement. Currently, there are no accepted methods for nondestructive determination of the extent of the irradiation embrittlement nor the actual fracture toughness of the reactor pressure vessel. This paper provides initial results of an effort addressing the use of ultrasonic attenuation as a suitable parameter for nondestructive determination of irradiation embrittlement in RPV steels. (orig.)

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)

Dynamic fracture mechanics with electromagnetic force and its application to fracture toughness and testing

International Nuclear Information System (INIS)

Yagawa, G.; Yoshimura, S.

1986-01-01

This study is concerned with the application of the electromagnetic force to the determination of the dynamic fracture toughness of materials. Taken is an edge-cracked specimen which carries a transient electric current I and is simply supported in a uniform and steady magnetic field B. As a result of their interaction, the dynamic electromagnetic force occurs in the whole body of the specimen, which is then deformed to fracture in the opening mode of cracking. For the evaluation of dynamic fracture toughness, the extended J integral with the effects of the electromagnetic force and inertia is calculated using the dynamic finite-element method. To determine the dynamic crack-initiation point in the experiment, the electric potential method is used in the case of brittle fracture, and the electric potential and the J-R curve methods in the case of ductile fracture, respectively. Using these techniques, the dynamic fracture toughness values of nuclear pressure vessel steel A508 class 3 are evaluated over a wide temperature range. (author)

The reinitiation of fracture at the tip of an arrested crack in a reactor pressure vessel: The effect of ligaments on the reinitiation K value

International Nuclear Information System (INIS)

Smith, E.

1986-01-01

During a hypothetical thermal shock event involving a water-cooled nuclear reactor steel pressure vessel, it is possible for a crack to propagate deep into the reactor vessel thickness by a series of run-arrest-reinitiation events. Furthermore, within the transition temperature regime, crack propagation and arrest are associated with a combination of cleavage and ductile rupture processes, the latter being manifested by ligaments that are normal to the crack plane and parallel to the direction of crack propagation. Earlier work by the author has modelled the effect of ligaments on the reinitiation of fracture at the tip of an arrested crack. Proceeding from the basis that the ligaments fail by a ductile rupture process, reinitiation K values were calculated. These values were appreciably higher than the experimental reinitiation K values for cracks in model vessels subject to thermal shock; it was therefore argued that the ligaments, which are present at arrest, are unlikely to fail entirely by ductile rupture prior to the reinitiation of fracture at an arrested crack tip. Instead it was suggested that the ligaments fail by cleavage, and consequently do not markedly affect the reinitiation K value, which therefore correlates with Ksub(IC). This paper's theoretical analysis extends the earlier work by relaxing a key assumption in the earlier work that, when calculating the reinitiation K value on the basis that the ligaments fail by ductile rupture, they should disappear completely prior to reinitiation. The new results, however, show that the predicted reinitiation K values are still so much greater than the model test reinitiation K values, that it is unlikely that the ligaments fail solely by ductile rupture prior to reinitiation. The view that the ligaments can fail by cleavage is therefore reinforced. (orig.)

Use of Master Curve technology for assessing shallow flaws in a reactor pressure vessel material

International Nuclear Information System (INIS)

Bass, Bennett Richard; Taylor, Nigel

2006-01-01

In the NESC-IV project an experimental/analytical program was performed to develop validated analysis methods for transferring fracture toughness data to shallow flaws in reactor pressure vessels subject to biaxial loading in the lower-transition temperature region. Within this scope an extensive range of fracture tests was performed on material removed from a production-quality reactor pressure vessel. The Master Curve analysis of this data is reported and its application to the assessment of the project feature tests on large beam test pieces.

Embrittlement recovery due to annealing of reactor pressure vessel steels

International Nuclear Information System (INIS)

Eason, E.D.; Wright, J.E.; Nelson, E.E.; Odette, G.R.; Mader, E.V.

1998-01-01

The irradiation embrittlement of nuclear reactor pressure vessels (RPV) can be reduced by thermal annealing at temperatures higher than the normal operating conditions. The objective of this work was to analyze the pertinent data and develop quantitative models for estimating the recovery in 41 J (30 ft-lb) Charpy transition temperature (TT) and Charpy upper shelf energy (USE) due to annealing. An analysis data base was developed, reviewed for completeness and accuracy, and documented as part of this work. Models were developed based on a combination of statistical techniques, including pattern recognition and transformation analysis, and the current understanding of the mechanisms governing embrittlement and recovery. The quality of models fitted in this project was evaluated by considering both the Charpy annealing data used for fitting and a surrogate hardness data base. This work demonstrates that microhardness recovery is a good surrogate for shift recovery and that there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes. (orig.)

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.

Foal Fractures: Osteochondral Fragmentation, Proximal Sesamoid Bone Fractures/Sesamoiditis, and Distal Phalanx Fractures.

Science.gov (United States)

Reesink, Heidi L

2017-08-01

Foals are susceptible to many of the same types of fractures as adult horses, often secondary to external sources of trauma. In addition, some types of fractures are specific to foals and occur routinely in horses under 1 year of age. These foal-specific fractures may be due to the unique musculoskeletal properties of the developing animal and may present with distinct clinical signs. Treatment plans and prognoses are tailored specifically to young animals. Common fractures not affecting the long bones in foals are discussed in this article, including osteochondral fragmentation, proximal sesamoid bone fractures/sesamoiditis, and distal phalanx fractures. Copyright © 2017 Elsevier Inc. All rights reserved.

Fracture assessment of weld material from a full-thickness clad RPV shell segment

International Nuclear Information System (INIS)

Keeney, J.A.; Bass, B.R.; McAfee, W.J.

1996-01-01

Fracture analysis was applied to full-thickness clad beam specimens containing shallow cracks in material for which metallurgical conditions are prototypic of those found in reactor pressure vessels (RPV) at beginning of life. The beam specimens were fabricated from a section of an RPV wall (removed from a canceled nuclear plant) that includes weld, plate, and clad material. Metallurgical factors potentially influencing fracture toughness for shallow cracks in the beam specimens include gradients of material properties and residual stresses due to welding and cladding applications. Fracture toughness estimates were obtained from load vs load-line displacement and load vs crack-mouth-opening displacement data using finite-element methods and estimation schemes based on the η-factor method. One of the beams experienced a significant amount of precleavage stable ductile tearing. Effects of precleavage tearing on estimates of fracture toughness were investigated using continuum damage models. Fracture toughness results from the clad beam specimens were compared with other deep- and shallow-crack single-edge notch bend (SENB) data generated previously from A533 Grade B plate material. Range of scatter for the clad beam data is consistent with that from the laboratory-scale SENB specimens tested at the same temperature

Fracture toughness of fabrication welds investigated by metallographic methods

International Nuclear Information System (INIS)

Canonico, D.A.; Crouse, R.S.

1978-01-01

The intermediate scale test vessels (ITV) were fabricated to provide test specimens that have sufficient wall thickness and simulate light water reactor pressure vessels. They were fabricated from grades of steel that are similar to those used for nuclear pressure vessels, having a wall thickness of 150mm and the same welded construction. They are, however, considerably smaller in height and diameter than actual vessels. To date, ten vessels have been fabricated and eight have been tested. In preparation for testing the eighth vessel (ITV-8), an extensive investigation was conducted of the toughness properties of the fabrication weld. It was thoroughly characterized and the fracture specimens used in this metallographic investigation were taken from that weld metal

Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

CERN Document Server

American Society for Testing and Materials. Philadelphia

2003-01-01

1.1 This guide covers the general procedures to be considered for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1). 1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constrain...

Avascular Necrosis of Trochlea After Supracondylar Humerus Fractures in Children.

Science.gov (United States)

Etier, Brian E; Doyle, J Scott; Gilbert, Shawn R

2015-10-01

Avascular necrosis (AVN) is a rare but important complication after supracondylar humerus fractures. Posttraumatic humerus deformity was first reported in 1948 and sporadically thereafter. AVN deformity has been classified as type A (AVN of the lateral ossification center) and type B (AVN of the entire medial crista and a metaphyseal portion). In this article, we present 5 cases of AVN after supracondylar humerus fracture, discuss the importance of late clinical findings, and postulate a mechanism of AVN in nondisplaced fractures. Five cases of AVN after supracondylar humerus fracture were reviewed from the Children's of Alabama database. Four of the 5 patients were female. Four patients sustained a Gartland type III fracture, and 1 patient sustained a nondisplaced Gartland type I fracture. Age at time of injury ranged from 5 years to 10 years. All patients had an asymptomatic clinical period after treatment and re-presented 6 months to 7 years later with elbow pain or loss of motion. All patients were treated symptomatically. AVN of the trochlea has a late clinical presentation. The cause of this complication is interruption of the trochlea blood supply. In displaced fractures, the medial and/or lateral vessels are injured, leading to type A or type B deformity. In nondisplaced fractures, the lateral vessels are interrupted by tamponade because of encased fracture hematoma; this presents as a type A deformity. Both type A and type B deformities can be clinically significant. AVN of the trochlea should be considered in patients with late presentation of pain or loss of motion after treatment of supracondylar humerus fractures.

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

Bilateral first rib fractures: A case report

Directory of Open Access Journals (Sweden)

Dilip Amonkar

2014-01-01

Full Text Available From the time first rib fractures were first described in 1869, they have been a source of anxiety to attendant trauma surgeons working in the accident and emergency department of major hospitals. First rib fractures are associated with major thoracic trauma and may involve injury to subclavian vessels, brachial plexus, and mediastinal structures. But these complications are more often seen following unilateral first rib fractures. In contrast, bilateral first rib fractures may follow insignificant trauma, suggesting a different mechanism involved. Serious vascular injuries and brachial plexus injuries are rare and angiograms for evaluation of these patients aren′t routinely warranted. The case that we report illustrates this very point.

Material problems in accident analysis of prestressed concrete reactor vessels

International Nuclear Information System (INIS)

Bazant, Z.P.

1977-01-01

Due to their very high energy absorption capability, as well as their inherent safety advantages, prestressed concrete reactor vessels are presently being keenly studied as the basic barrier to contain hypothetical core disruptive accidents in a fast breeder reactor. One problem investigated is the nonlinear constitutive behavior and failure criteria for concrete. Previously, a comprehensive theory, called endochronic theory, has been shown to satisfy all basic currently known features of test data. Nevertheless uncertainty still exists with regard to non-proportional loading paths, for which good test data are lacking at present. An extension of the endochronic theory which correlates best with general experimental evidence and includes fracturing terms is given, and a comparison with vertex-type hardening in plasticity is made. A second problem which must be analysed in accident situations is the high temperature shock on the concrete walls (due to liquid sodium, up to 850 0 C). Refining a previous crude formulation, a rational model for calculating moisture and heat transfer and pore pressures in concrete subjected to thermal shock is presented. In conclusion, a new design concept, in which the concrete vessel is completely dehydrated and kept hot throughout its service life in order to substantially improve its response to thermal shock as well as liquid sodium contact, is described. (Auth.)

Synthesis of industrial applications of local approach to fracture models

International Nuclear Information System (INIS)

Eripret, C.

1993-03-01

This report gathers different applications of local approach to fracture models to various industrial configurations, such as nuclear pressure vessel steel, cast duplex stainless steels, or primary circuit welds such as bimetallic welds. As soon as models are developed on the basis of microstructural observations, damage mechanisms analyses, and fracture process, the local approach to fracture proves to solve problems where classical fracture mechanics concepts fail. Therefore, local approach appears to be a powerful tool, which completes the standard fracture criteria used in nuclear industry by exhibiting where and why those classical concepts become unvalid. (author). 1 tab., 18 figs., 25 refs

A finite element elastic-plastic analysis of residual stresses due to clad welding in reactor vessels

International Nuclear Information System (INIS)

Buchalet, C.; Riccardella, P.C.

1972-01-01

Residual stresses due to weld deposited cladding on the inside of a typical Westinghouse pressurized water reactor vessel are investigated using an axisymmetric finite element elastic-plastic analysis. At the beginning of the analysis, one head of the weld cladding is assumed to lie on the reactor vessel wall at melting temperature (2600degF), but in the solid phase, while the vessel remains at 300degF (preheat temperature). All material properties used in the calculations are taken as temperature-dependent. Temperature profiles are obtained in the cladding and base metal at several discrete time intervals. These temperatures profiles are used to obtain the stress distribution for the same time intervals. Residual hoop tensile stresses of approximately 25 ksi were found to exist in the cladding. Peak tensile stresses in the hoop direction occur in the base metal near the cladding interface and reach a value of 60 ksi at the end of the transient. The tensile stress decreases very rapidly through the thickness of the base metal and becomes insignificant at about two inches from the inside surface. In order to lower residual stresses, a post-weld heat treatment is performed by uniformly heating the vessel to 1100degF, holding at that temperature for a specified period of time and then cooling slowly. The analysis shows that after this treatment, the peak stresses in the base metal decrease from 60 ksi to 32 ksi, while the stress in the cladding does not change significantly. (author)

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

Primo vessel inside a lymph vessel emerging from a cancer tissue.

Science.gov (United States)

Lee, Sungwoo; Ryu, Yeonhee; Cha, Jinmyung; Lee, Jin-Kyu; Soh, Kwang-Sup; Kim, Sungchul; Lim, Jaekwan

2012-10-01

Primo vessels were observed inside the lymph vessels near the caudal vena cava of a rabbit and a rat and in the thoracic lymph duct of a mouse. In the current work we found a primo vessel inside the lymph vessel that came out from the tumor tissue of a mouse. A cancer model of a nude mouse was made with human lung cancer cell line NCI-H460. We injected fluorescent nanoparticles into the xenografted tumor tissue and studied their flow in blood, lymph, and primo vessels. Fluorescent nanoparticles flowed through the blood vessels quickly in few minutes, and but slowly in the lymph vessels. The bright fluorescent signals of nanoparticles disappeared within one hour in the blood vessels but remained much longer up to several hours in the case of lymph vessels. We found an exceptional case of lymph vessels that remained bright with fluorescence up to 24 hours. After detailed examination we found that the bright fluorescence was due to a putative primo vessel inside the lymph vessel. This rare observation is consistent with Bong-Han Kim's claim on the presence of a primo vascular system in lymph vessels. It provides a significant suggestion on the cancer metastasis through primo vessels and lymph vessels. Copyright © 2012. Published by Elsevier B.V.

Permeability evolution due to dissolution of natural shale fractures reactivated by fracking

Science.gov (United States)

Kwiatkowski, Kamil; Kwiatkowski, Tomasz; Szymczak, Piotr

2015-04-01

Investigation of cores drilled from gas-bearing shale formations reveals a relatively large number of calcite-cemented fractures. During fracking, some of these fractures will be reactivated [1-2] and may become important flow paths in the resulting fracture system. In this communication, we investigate numerically the effect of low-pH reactive fluid on such fractures. The low-pH fluids can either be pumped during the initial fracking stage (as suggested e.g. by Grieser et al., [3]) or injected later, as part of enhanced gas recovery (EGR) processes. In particular, it has been suggested that CO2 injection can be considered as a method of EGR [4], which is attractive as it can potentially be combined with simultaneous CO2 sequestration. However, when mixed with brine, CO2 becomes acidic and thus can be a dissolving agent for the carbonate cement in the fractures. The dissolution of the cement leads to the enhancement of permeability and interconnectivity of the fracture network and, as a result, increases the overall capacity of the reservoir. Importantly, we show that the dissolution of such fractures proceeds in a highly non-homogeneous manner - a positive feedback between fluid transport and mineral dissolution leads to the spontaneous formation of pronounced flow channels, frequently referred to as "wormholes". The wormholes carry the chemically active fluid deeper inside the system, which dramatically speeds up the overall permeability increase. If the low-pH fluids are used during fracking, then the non-uniform dissolution becomes important for retaining the fracture permeability, even in the absence of the proppant. Whereas a uniformly etched fracture will close tightly under the overburden once the fluid pressure is removed, the nonuniform etching will tend to maintain the permeability since the less dissolved regions will act as supports to keep more dissolved regions open. [1] Gale, J. F., Reed, R. M., Holder, J. (2007). Natural fractures in the Barnett

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Effects of irradiation on strength and toughness of commercial LWR vessel cladding

International Nuclear Information System (INIS)

Haggag, F.M.; Corwin, W.R.; Alexander, D.J.; Nanstad, R.K.

1987-01-01

The potential for stainless steel cladding to improve the fracture behavior of an operating nuclear reactor pressure vessel, particularly during certain overcooling transients, may depend greatly on the properties of the irradiated cladding. Therefore, weld overlay cladding irradiated at temperatures and to fluences relevant to power reactor operation was examined. The cladding was applied to a pressure vessel steel plate by the three-wire series-arc commercial method. Cladding was applied in three layers to provide adequate thickness for the fabrication of test specimens. The three-wire series-arc procedure, developed by Combustion Engineering, Inc., Chattanooga, Tennessee, produced a highly controlled weld chemistry, microstructure, and fracture properties in all three layers of the weld. Charpy V-notch and tensile specimens were irradiated at 288 0 C to fluence levels of 2 and 5 x 10 19 neutrons/cm 2 (>1 MeV). Postirradiation testing results show that, in the test temperature range from -125 to 288 0 C, the yield strength increased by 8 to 30%, ductility insignificantly increased, while there was almost no change in ultimate tensile strength. All cladding exhibited ductile-to-brittle transition behavior during Charpy impact testing, due to the dominance of delta-ferrite failures at low temperatures. On the upper shelf, energy was reduced, due to irradiation exposure, 15 and 20%, while the lateral expansion was reduced 43 and 41%, at 2 and 5 x 10 19 neutrons/cm 2 (>1 MeV), respectively. In addition, radiation damage resulted in 13 and 28 0 C shifts of the Charpy impact transition temperature at the 41-J level for the low and high fluences, respectively

Ageing Management Review of the reactor pressure vessels in Laguna Verde NPP

International Nuclear Information System (INIS)

Gris Cruz, Magdalena; Arganis, Carlos R.J.; Medina Almazan, A. Liliana

2012-01-01

In the present paper, for both units of Laguna Verde Nuclear Power Plant (LVNPP), the Ageing Management Review of the reactor pressure Vessel was carried out, including the identification of the intended functions, the materials and the environments. The evaluation of the ageing effect/mechanism and the Aging management programs currently implemented were prepared. The most important aging effects/ mechanisms are: loss of fracture toughness due to neutron irradiation embrittlement, fatigue, stress corrosion cracking (SCC), general corrosion and erosion-corrosion. The neutron irradiation embrittlement is managed by the reactor vessel materials surveillance program. The fatigue is a Time Limited Aging Analysis (TLAA), for which is necessary to calculate some fatigue usage factors. SCC is managed by, the In service inspections (ISI) program, but also by the Water Chemistry program, including, currently, On Line Noble Chem. The water chemistry program also manages General Corrosion and erosion-corrosion. The results were compared with the GALL report. (author)

Effect of nickel content on mechanical properties and fracture toughness of weld metal of WWER-1000 reactor vessel welded joints

Energy Technology Data Exchange (ETDEWEB)

Zubchenko, A.S.; Vasilchenko, G.S.; Starchenko, E.G.; Nosov, S.I

2004-08-01

Welding of WWER-1000 reactor vessel of steel 15X2HMPHIA is performed using the C{sub B}-12X2H2MAA wire and PHI-16 or PHI-16A flux. Nickel content in the weld metal usually lays within the limits 1.2-1.9%. The experimental data is shown on the weld metal with the nickel contents 1.28-2.45% after irradiation with fluence up to 260.10{sup 22}n/m{sup 2} at energy more than 0.5 MEV. The embrittlement was measured by shift of critical brittleness temperature. Has appeared, that the weld metal with the low nickel content is the least responsive to irradiation embrittlement. The mechanical properties and fracture toughness of the weld metal with the contents of a nickel less than 1.3% are studied. Specimens CT-1T are tested, the 'master-curve', and its confidence bounds with probability of destruction 5 and 95% is built. 'Master-curve' in the specified confidence interval is affirmed by CT-4T specimens test data. Is shown, that the mechanical properties and fracture toughness of the weld metal with the contents of nickel less than 1.3% satisfy the normative requirements.

Associations of early premenopausal fractures with subsequent fractures vary by sites and mechanisms of fractures.

Science.gov (United States)

Honkanen, R; Tuppurainen, M; Kroger, H; Alhava, E; Puntila, E

1997-04-01

In a retrospective population-based study we assessed whether and how self-reported former fractures sustained at the ages of 20-34 are associated with subsequent fractures sustained at the ages of 35-57. The 12,162 women who responded to fracture questions of the baseline postal enquiry (in 1989) of the Kuopio Osteoporosis Study, Finland formed the study population. They reported 589 former and 2092 subsequent fractures. The hazard ratio (HR), with 95% confidence interval (CI), of a subsequent fracture was 1.9 (1.6-2.3) in women with the history of a former fracture compared with women without such a history. A former low-energy wrist fracture was related to subsequent low-energy wrist [HR = 3.7 (2.0-6.8)] and high-energy nonwrist [HR = 2.4 (1.3-4.4)] fractures, whereas former high-energy nonwrist fractures were related only to subsequent high-energy nonwrist [HR = 2.8 (1.9-4.1)] but not to low-energy wrist [HR = 0.7 (0.3-1.8)] fractures. The analysis of bone mineral density (BMD) data of a subsample of premenopausal women who underwent dual x-ray absorptiometry (DXA) during 1989-91 revealed that those with a wrist fracture due to a fall on the same level at the age of 20-34 recorded 6.5% lower spinal (P = 0.140) and 10.5% lower femoral (P = 0.026) BMD than nonfractured women, whereas the corresponding differences for women with a former nonwrist fracture due to high-energy trauma were -1.8% (P = 0.721) and -2.4% (P = 0. 616), respectively. Our results suggest that an early premenopausal, low-energy wrist fracture is an indicator of low peak BMD which predisposes to subsequent fractures in general, whereas early high-energy fractures are mainly indicators of other and more specific extraskeletal factors which mainly predispose to same types of subsequent fractures only.

Tumor Blood Vessel Dynamics

Science.gov (United States)

Munn, Lance

2009-11-01

``Normalization'' of tumor blood vessels has shown promise to improve the efficacy of chemotherapeutics. In theory, anti-angiogenic drugs targeting endothelial VEGF signaling can improve vessel network structure and function, enhancing the transport of subsequent cytotoxic drugs to cancer cells. In practice, the effects are unpredictable, with varying levels of success. The predominant effects of anti-VEGF therapies are decreased vessel leakiness (hydraulic conductivity), decreased vessel diameters and pruning of the immature vessel network. It is thought that each of these can influence perfusion of the vessel network, inducing flow in regions that were previously sluggish or stagnant. Unfortunately, when anti-VEGF therapies affect vessel structure and function, the changes are dynamic and overlapping in time, and it has been difficult to identify a consistent and predictable normalization ``window'' during which perfusion and subsequent drug delivery is optimal. This is largely due to the non-linearity in the system, and the inability to distinguish the effects of decreased vessel leakiness from those due to network structural changes in clinical trials or animal studies. We have developed a mathematical model to calculate blood flow in complex tumor networks imaged by two-photon microscopy. The model incorporates the necessary and sufficient components for addressing the problem of normalization of tumor vasculature: i) lattice-Boltzmann calculations of the full flow field within the vasculature and within the tissue, ii) diffusion and convection of soluble species such as oxygen or drugs within vessels and the tissue domain, iii) distinct and spatially-resolved vessel hydraulic conductivities and permeabilities for each species, iv) erythrocyte particles advecting in the flow and delivering oxygen with real oxygen release kinetics, v) shear stress-mediated vascular remodeling. This model, guided by multi-parameter intravital imaging of tumor vessel structure

LNG cascading damage study. Volume I, fracture testing report.

Energy Technology Data Exchange (ETDEWEB)

Petti, Jason P.; Kalan, Robert J.

2011-12-01

As part of the liquefied natural gas (LNG) Cascading Damage Study, a series of structural tests were conducted to investigate the thermal induced fracture of steel plate structures. The thermal stresses were achieved by applying liquid nitrogen (LN{sub 2}) onto sections of each steel plate. In addition to inducing large thermal stresses, the lowering of the steel temperature simultaneously reduced the fracture toughness. Liquid nitrogen was used as a surrogate for LNG due to safety concerns and since the temperature of LN{sub 2} is similar (-190 C) to LNG (-161 C). The use of LN{sub 2} ensured that the tests could achieve cryogenic temperatures in the range an actual vessel would encounter during a LNG spill. There were four phases to this test series. Phase I was the initial exploratory stage, which was used to develop the testing process. In the Phase II series of tests, larger plates were used and tested until fracture. The plate sizes ranged from 4 ft square pieces to 6 ft square sections with thicknesses from 1/4 inches to 3/4 inches. This phase investigated the cooling rates on larger plates and the effect of different notch geometries (stress concentrations used to initiate brittle fracture). Phase II was divided into two sections, Phase II-A and Phase II-B. Phase II-A used standard A36 steel, while Phase II-B used marine grade steels. In Phase III, the test structures were significantly larger, in the range of 12 ft by 12 ft by 3 ft high. These structures were designed with more complex geometries to include features similar to those on LNG vessels. The final test phase, Phase IV, investigated differences in the heat transfer (cooling rates) between LNG and LN{sub 2}. All of the tests conducted in this study are used in subsequent parts of the LNG Cascading Damage Study, specifically the computational analyses.

Applications of energy-release-rate techniques to part-through cracks in experimental pressure vessels

International Nuclear Information System (INIS)

Bass, B.R.; Bryan, R.H.; Bryson, J.W.; Merkle, J.G.

1982-01-01

In nonlinear applications of computational fracture mechanics, energy release rate techniques are used increasingly for computing stress intensity parameters of crack configurations. Recently, deLorenzi used the virtual-crack-extension method to derive an analytical expression for the energy release rate that is better suited for three-dimensional calculations than the well-known J-integral. Certain studies of fracture phenomena, such as pressurized-thermal-shock of cracked structures, require that crack tip parameters be determined for combined thermal and mechanical loads. A method is proposed here that modifies the isothermal formulation of deLorenzi to account for thermal strains in cracked bodies. This combined thermo-mechanical formulation of the energy release rate is valid for general fracture, including nonplanar fracture, and applies to thermo-elastic as well as deformation plasticity material models. Two applications of the technique are described here. In the first, semi-elliptical surface cracks in an experimental test vessel are analyzed under elastic-plastic conditions using the finite element method. The second application is a thick-walled test vessel subjected to combined pressure and thermal shock loadings

An interim report on shallow-flaw fracture technology development

International Nuclear Information System (INIS)

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

1995-01-01

Shallow-flaw fracture technology is being developed for application to the safety assessment of radiation-embrittled nuclear reactor pressure vessels (RPVS) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite-element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack tip of shallow surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil ductility temperature (NDT) performs better than the reference temperature for nil ductility transition (RT NDT ) as a normalizing parameter for shallow-flaw fracture toughness data, (3) biaxial loading can reduce the shallow-flaw fracture toughness, (4) stress-based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on shallow-flaw fracture toughness because in-plane stresses at the crack tip are not influenced by biaxial loading, and (5) a strain-based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow-flaw fracture toughness

Intrapartum sacral stress fracture due to pregnancy-related osteoporosis: a case report.

Science.gov (United States)

Oztürk, Gülcan; Külcü, Duygu Geler; Aydoğ, Ece

2013-01-01

Low back pain (LBP) and hip pain frequently occur during pregnancy and postpartum period. Although pelvic and mechanic lesions of the soft tissues are most responsible for the etiology, sacral fracture is also one of the rare causes. A 32-year-old primigravid patient presented with LBP and right hip pain which started 3 days after vaginal delivery. Although direct radiographic examination was normal, magnetic resonance imaging of the sacrum revealed sacral stress fracture. Lumbar spine and femoral bone mineral density showed osteoporosis as a risk factor. There were no other risk factors such as trauma, excessive weight gain, and strenuous physical activity. It is considered that the patient had sacral fatigue and insufficiency fracture in intrapartum period. The patient's symptoms subsided in 3 months after physical therapy and rest. In conclusion, sacral fractures during pregnancy and postpartum period, especially resulting from childbirth, are very rare. To date, there are two cases in the literature. In cases who even do not have risk factors related to vaginal delivery such as high birth weight infant and the use of forceps, exc., sacral fracture should be considered in the differential diagnosis of LBP and hip pain started soon after child birth. Pregnancy-related osteoporosis may lead to fracture during vaginal delivery.

Development and application of an LWR reactor pressure vessel-specific flaw distribution

International Nuclear Information System (INIS)

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

1991-01-01

Previous efforts by the US Department of Energy have shown that the PWR reactor vessel integrity predictions performed through probabilistic fracture mechanics analysis for a pressurized thermal shock event are significantly sensitive to the overall flaw distribution input. It has also been shown that modern 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. The methodology helped provide original insight into several key inspection performance and vessel integrity prediction practice issues that will impact future vessel integrity evaluation. This paper briefly discusses the development and application of the methodology and the impact to future vessel integrity analyses

Tearing stability analysis of an axial surface flaw in thick-walled pressure vessels

International Nuclear Information System (INIS)

Zahoor, A.; Ghassemi, B.B.

1991-01-01

This paper presents two fracture mechanics models for evaluation of an axial surface flaw in pressure vessels. The surface flaw is located on the outside surface of the vessel. The first model assumes yielding of the remaining ligament directly ahead of the flaw. The second model assumes contained yielding ahead of the flaw and uses a linear elastic fracture mechanics solution. The former model is suitable for cases where the combination of material toughness, flaw size, and load is such that initiation of flaw growth follows ligament yielding. The latter model is suitable for low-toughness materials where initiation of crack growth and potential tearing instability may occur prior to the yielding of the ligament. Both models are suitable for thick-walled vessels. The paper discusses the applicability regime for both models. The models are then applied to a test vessel and the predicted failure pressure is compared against the pressure attained in the test. Results show that both models can be applied successfully. In particular, the contained yielding model when used with the plane-stress assumption can give reasonable predictions even for cases that involve yielding of the ligament. (orig.)

Tearing stability analysis of an axial surface flaw in thick-walled pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Ghassemi, B.B. (NOVETECH Corp., Rockville, MD (USA))

1991-04-01

This paper presents two fracture mechanics models for evaluation of an axial surface flaw in pressure vessels. The surface flaw is located on the outside surface of the vessel. The first model assumes yielding of the remaining ligament directly ahead of the flaw. The second model assumes contained yielding ahead of the flaw and uses a linear elastic fracture mechanics solution. The former model is suitable for cases where the combination of material toughness, flaw size, and load is such that initiation of flaw growth follows ligament yielding. The latter model is suitable for low-toughness materials where initiation of crack growth and potential tearing instability may occur prior to the yielding of the ligament. Both models are suitable for thick-walled vessels. The paper discusses the applicability regime for both models. The models are then applied to a test vessel and the predicted failure pressure is compared against the pressure attained in the test. Results show that both models can be applied successfully. In particular, the contained yielding model when used with the plane-stress assumption can give reasonable predictions even for cases that involve yielding of the ligament. (orig.).

Evaluation of creep damage due to stress relaxation in SA533 grade B class 1 and SA508 class 3 pressure vessel steels

International Nuclear Information System (INIS)

Hoffmann, C.L.; Urko, W.

1993-01-01

Creep damage can result from stress relaxation of residual stresses in components when exposed to high temperature thermal cycles. Pressure vessels, such as the reactor vessel of the modular high-temperature gas reactor (MHTGR), which normally operate at temperatures well below the creep range can develop relatively high residual stresses in high stress locations. During short term excursions to elevated-temperatures, creep damage can be produced by the loadings on the vessel. In addition, residual stresses will relax out, causing greater creep damage in the pressure vessel material than might otherwise be calculated. The evaluation described in this paper assesses the magnitude of the creep damage due to relaxation of residual stresses resulting from short term exposure of the pressure vessel material to temperatures in the creep range. Creep relaxation curves were generated for SA533 Grade B, Class 1 and SA508 Class 3 pressure vessel steels using finite element analysis of a simple uniaxial truss loaded under constant strain conditions to produce an initial axial stress equal to 1.25 times the material yield strength at temperature. The strain is held constant for 1000 hours at prescribed temperatures from 700 F to 1000 F. The material creep law is used to calculate the relaxed stress for each time increment. The calculated stress relaxation versus time curves are compared with stress relaxation test data. Creep damage fractions are calculated by integrating the stress relaxation versus time curves and performing a linear creep damage summation using the minimum stress to rupture curves at the respective relaxation temperatures. Cumulative creep damage due to stress relaxation as a function of time and temperature is derived from the linear damage summation

Comparison of fracture properties for two types of low alloy steels

International Nuclear Information System (INIS)

Nasreldin, A.M.

2004-01-01

The fracture properties of two types of low alloy steels used in the pressure vessel and boilers industry were determined. The first type was the steel A533-B which comprised a fully bainitic microstructure. The second one was the C-Mn steel which consisted of ferritic-pearlitic microstructure. The following fracture properties were determined using instrumented impact testing: the total fracture energy, the crack initiation and propagation energies, the brittleness transition temperature and the local fracture stress. The steel A533-B showed better fracture properties at high testing temperatures, while the C-Mn steel displayed higher resistance to brittle fracture at low testing temperatures. The results were discussed in relation to the difference in microstructure and fracture surface morphology for both steels

Execution of programme of post-service study of the condition of nuclear icebreaker Lenin reactor 1 pressure vessel metal and perspectives of application of results to increase service life of nuclear icebreakers reactor vessels

International Nuclear Information System (INIS)

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

2001-01-01

With the aim of determining the irradiation-induced embrittlement of a base metal and a weld metal in a pressure vessel of the nuclear icebreaker Lenin after 18 years operation the specimens cut out of a vessel wall are used to study the chemical composition and to carry out impact tests. From the test results the temperature dependences of fracture energy are built which define the irradiation embrittlement of a low alloy steel. It is noted that the annealing at 475 deg C for 100 h results in complete restoration of impact strength. Based on the results obtained the following conclusions are formulated: a reactor vessel base metal has high resistance to brittle fracture and high radiation resistance; a weld metal possesses rather high radiation resistance but unsatisfactory ductile-brittle transition temperature (∼ 63 deg C); for cladded vessels there is a potential reserve in the form of enhanced radiation resistance of an undercladding layer; in the final stage of operation the coolant temperature is recommended to be kept at the highest possible level [ru

Bilateral Incomplete Atypical Femoral Fracture due to Long-Term Bisphosphonate Use: A Case Report

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Sibel Başaran

2017-08-01

Full Text Available Although the overall safety profile of bisphosphonates (BP is favorable, adverse effects associated with long-term use have came up during recent years. In this report, a case of bilateral incomplete atypical femoral fracture (AFF due to prolonged BP use was presented. A 69-year-old patient, who has been in surgical menopause for 20 years and was started on BP following vertebral fracture almost 10 years ago, was admitted with thigh pain, which was increased two weeks ago. On physical examination, she had antalgic gait, increased thoracic kyphosis and tenderness to percussion over the thoracolumbar region. Lateral cortical thickness in the subtrochanteric region of both femurs and cortical radiolucency on the left femur were observed on plain radiography. Loss of height in L3 and L4 vertebrae was detected on vertebral radiography. Serum 25-hydroxy vitamin D [25(OHD], parathyroid hormone, alkaline phosphatase and calcium levels, along with osteoporosis markers were all within the normal ranges. As the patient was diagnosed with AFF, BP therapy was terminated and vitamin D-calcium supplementation was continued. Since she did not have severe pain, conservative management (limited weight bearing, using a walking stick was recommended for 3 months. Teriparatide therapy was started and she was discharged with recommendations. AFF, which is a rare disorder, should be kept in mind in patients on long-term BP treatment who are admitted with thigh pain and, necessary interventions should be tailored before the occurrence of complete fracture.

2014 New Trends in Fatigue and Fracture Conference

CERN Document Server

Milovic, Ljubica

2017-01-01

This book is a compilation of selected papers from the 2014 New Trends in Fatigue and Fracture (NT2F14) Conference, which was held in Belgrade, Serbia. This prestigious conference brought together delegates from around the globe to discuss how to characterize, predict and analyze the fatigue and fracture of engineering materials, components, and structures using theoretical, experimental, numerical and practical approaches. It highlights some important new trends in fracture mechanics presented at the conference, such as: • two-parameter fracture mechanics, arising from the coupling of fracture toughness and stress constraints • high-performance steel for gas and oil transportation and production (pressure vessels and boilers) • safety and reliability of welded joints This book includes 12 contributions from well-known international scientists and a special tribute dedicated to the scientific contributions of Stojan Sedmark, who passed away in 2014.

Analysis of the applicability of fracture mechanics on the basis of large scale specimen testing

International Nuclear Information System (INIS)

Brumovsky, M.; Polachova, H.; Sulc, J.; Anikovskij, V.; Dragunov, Y.; Rivkin, E.; Filatov, V.

1988-01-01

The verification is dealt with of fracture mechanics calculations for WWER reactor pressure vessels by large scale model testing performed on the large testing machine ZZ 8000 (maximum load of 80 MN) in the Skoda Concern. The results of testing a large set of large scale test specimens with surface crack-type defects are presented. The nominal thickness of the specimens was 150 mm with defect depths between 15 and 100 mm, the testing temperature varying between -30 and +80 degC (i.e., in the temperature interval of T ko ±50 degC). Specimens with a scale of 1:8 and 1:12 were also tested, as well as standard (CT and TPB) specimens. Comparisons of results of testing and calculations suggest some conservatism of calculations (especially for small defects) based on Linear Elastic Fracture Mechanics, according to the Nuclear Reactor Pressure Vessel Codes which use the fracture mechanics values from J IC testing. On the basis of large scale tests the ''Defect Analysis Diagram'' was constructed and recommended for brittle fracture assessment of reactor pressure vessels. (author). 7 figs., 2 tabs., 3 refs

Revisiting the reactor pressure vessel for long-time operation

International Nuclear Information System (INIS)

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

2013-01-01

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

Probabilistic retinal vessel segmentation

Science.gov (United States)

Wu, Chang-Hua; Agam, Gady

2007-03-01

Optic fundus assessment is widely used for diagnosing vascular and non-vascular pathology. Inspection of the retinal vasculature may reveal hypertension, diabetes, arteriosclerosis, cardiovascular disease and stroke. Due to various imaging conditions retinal images may be degraded. Consequently, the enhancement of such images and vessels in them is an important task with direct clinical applications. We propose a novel technique for vessel enhancement in retinal images that is capable of enhancing vessel junctions in addition to linear vessel segments. This is an extension of vessel filters we have previously developed for vessel enhancement in thoracic CT scans. The proposed approach is based on probabilistic models which can discern vessels and junctions. Evaluation shows the proposed filter is better than several known techniques and is comparable to the state of the art when evaluated on a standard dataset. A ridge-based vessel tracking process is applied on the enhanced image to demonstrate the effectiveness of the enhancement filter.

A case of bowel perforation due to traumatic hernia at a pelvic fracture site: a case report and review of the literature.

Science.gov (United States)

Tanaka, Ryota; Nagahara, Hisashi; Maeda, Kiyoshi; Ohtani, Hiroshi; Shibutani, Masatsune; Tamura, Tatsuro; Ikeya, Tetsuro; Sugano, Kenji; Iseki, Yasuhito; Sakurai, Katsunobu; Yamazoe, Sadaaki; Kimura, Kenjiro; Toyokawa, Takahiro; Amano, Ryosuke; Kubo, Naoshi; Tanaka, Hiroaki; Muguruma, Kazuya; Hirakawa, Kosei; Ohira, Masaichi

2017-07-12

Common complications of pelvic fractures include visceral injury, large-volume hemorrhage, genitourinary injury, rectal injury, and pulmonary embolism. On the other hand, traumatic hernia is a rare complication, especially in association with pelvic fractures. We report a case of bowel perforation due to traumatic hernia at a pelvic fracture site. A 65-year-old female was presented at our hospital for further examination and treatment of ileus. She was diagnosed with bowel perforation due to traumatic hernia at a pelvic fracture site, and an emergency operation was thus immediately performed. We performed segmental jejunum resection and constructed jejunostomy, and the iliac bone fracture was fixed with four pins. In the postoperative course, she received antibiotics and vasopressors for septic shock. However, there was no need for either a ventilator, dialysis or admission to the ICU. At seven days after the operation, a residual abscess was detected in the pouch of Douglas. We performed percutaneous drainage (Clavien-Dindo IIIa) and jejunostomy closedown 35 days after the first operation. The postoperative course was without complication, but she received rehabilitation until she was able to walk unaided. She was discharged 64 days after the first operation. The occurrence of traumatic hernia is rare, especially in association with pelvic fractures. Although its rarity, traumatic hernia follows a severe course. Thus, proper diagnosis and effective treatment are necessary. Surgeons treating patients with pelvic injuries should consider the possibility of any complications and perform a work-up examination in order to achieve an accurate diagnosis at an earlier time point.

Over-extending reduction combined with unilateral approach percutaneous vertebroplasty for the treatment of vertebral compression fractures due to osteoporosis

International Nuclear Information System (INIS)

Wei Xinjian; Ji Xianghui; Cao Fei; Zhang Fuhua

2012-01-01

Objective: To assess the clinical effect of over-extending reduction combined with percutaneous vertebroplasty (PVP) in treating vertebral compression fractures caused by osteoporosis. Methods: A total of 16 patients with vertebral compression fractures due to osteoporosis were treated with over-extending reduction by using traction on the operation table, and then PVP through trans-single-pedicular approach was performed on the fractured vertebra. The visual analogue scale (VAS) was used to evaluate the clinical effectiveness. The preoperative and postoperative heights of the fractured vertebral body were determined, and the vertebral height recovery ratio was calculated. Results: Technical success was achieved in 20 vertebrae of 16 cases. Bone cement leakage was observed in front of the vertebral body (n=5), in the side of vertebral body (n=20) and within the intervertebral (n=2). After the treatment VAS score decreased from preoperative 8.5±1.2 to postoperative 2.5±1.4. The vertebral height recovery ratio was (40.1±23.5)%. After the surgery, the VAS score and the vertebral height were significantly improved (P<0.05). Conclusion: The over-extending reduction combined with PVP through trans-single-pedicular approach is an effective treatment for vertebral compression fractures caused by osteoporosis. (authors)

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

International Nuclear Information System (INIS)

Jones, M.R.

1988-06-01

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

Relationships between Charpy impact shelf energies and upper shelf Ksub(IC) values for reactor pressure vessel steels

International Nuclear Information System (INIS)

Witt, F.J.

1983-01-01

Charpy shelf data and lower bound estimates of Ksub(IC) shelf data for the same steels and test temperatures are given. Included are some typical reactor pressure vessel steels as well as some less tough or degraded steels. The data were evaluated with shelf estimates of Ksub(IC) up to and exceeding 550 MPa√m. It is shown that the high shelf fracture toughness representative of tough reactor pressure vessel steels may be obtained from a knowledge of the Charpy shelf energies. The toughness transition may be obtained either by testing small fracture toughness specimens or by Charpy energy indexing. (U.K.)

An assessment of acoustic emission for nuclear pressure vessel monitoring

International Nuclear Information System (INIS)

Scruby, C.B.

1983-01-01

Recent research has greatly improved our understanding of the basic mechanisms of deformation and fracture that generate detectable acoustic emission signals in structural steels. A critical review of the application of acoustic emission (AE) to the fabrication, proof testing and in-service monitoring of nuclear pressure vessels is presented in the light of this improved understanding. The detectability of deformation and fracture processes in pressure vessel steels is discussed, and recommendations made for improving source location accuracy and the development of quantitative source assessment techniques. Published data suggest that AE can make an important contribution to fabrication monitoring, and to the detection of defects in lower toughness materials during vessel proof testing. In high toughness materials, however, the signals generated during ductile crack growth may frequently be too weak for reliable detection. The feasibility of AE for continuous monitoring has not yet been adequately demonstrated because of high background noise levels and uncertainty about AE signal strengths from the defect growth processes that occur in service. In-service leak detection by AE shows considerable promise. It is recommended that further tests are carried out with realistic defects, and under realistic conditions of loading (including thermal shock and fatigue) and of environment. (author)

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

Is new vertebral compression fractures after percutaneous vertebroplasty: due to the ongoing osteoporosis or complication?

International Nuclear Information System (INIS)

Shi Li'na; Wu Chungen; Li Wenbin; Gu Yifeng; Wang Jue; Cheng Yongde

2011-01-01

Objective: To clarify whether percutaneous vertebroplasty (PVP) for osteoporotic vertebral compression fractures will increase the risk of new vertebral fractures or not. Methods: A total of 197 vertebrae in 120 patients with new osteoporotic vertebral compression fractures, which were proved by CT, MRI and/or plain radiography, were enrolled in this study. Based on the therapeutic means, the patients were divided into two groups. Conservative therapy was employed in group A (n=60, 87 vertebrae), while PVP was carried out in group B (n=60, 100 vertebrae). All the patients were followed up for 1-5 years. Careful observations were carried out on the occurrence of new vertebral fractures. The location, distribution, the incidence of new fractures, the incidence of adjacent-level vertebral fractures (next to the treated vertebra), the time interval, etc. were documented. The results were compared between the two groups and the relative risks of new fracture for the two groups were assessed. Results: The incidence of new fracture and new adjacent-level vertebral fracture in group A was 27% and 56% (n=15), respectively, while in group B it was 38% and 52.5% (n=21), respectively. The difference in the incidence and the distribution of the location of new fractures was not statistically significant between the two groups (P>0.05). The relative risk of adjacent-level fracture versus nonadjacent vertebrae for group A and group B was 1.076 and 0.925, respectively. No higher fracture risk for adjacent-versus-nonadjacent vertebrae was found in both two groups. The mean time interval to the onset of new fracture for group A and group B was (12.9±8.5) months and (13.6±16.2) months, respectively, and the difference was not significant (Log-rank, P>0.05). Conclusion: Compared with conservative therapy, PVP does not increase the risk of inducing new vertebral fractures. PVP does not carry higher risk in inducing adjacent-level vertebral fractures when compared with that of distant

Is human fracture hematoma inherently angiogenic?

LENUS (Irish Health Repository)

Street, J

2012-02-03

This study attempts to explain the cellular events characterizing the changes seen in the medullary callus adjacent to the interfragmentary hematoma during the early stages of fracture healing. It also shows that human fracture hematoma contains the angiogenic cytokine vascular endothelial growth factor and has the inherent capability to induce angiogenesis and thus promote revascularization during bone repair. Patients undergoing emergency surgery for isolated bony injury were studied. Raised circulating levels of vascular endothelial growth factor were seen in all injured patients, whereas the fracture hematoma contained significantly higher levels of vascular endothelial growth factor than did plasma from these injured patients. However, incubation of endothelial cells in fracture hematoma supernatant significantly inhibited the in vitro angiogenic parameters of endothelial cell proliferation and microtubule formation. These phenomena are dependent on a local biochemical milieu that does not support cytokinesis. The hematoma potassium concentration is cytotoxic to endothelial cells and osteoblasts. Subcutaneous transplantation of the fracture hematoma into a murine wound model resulted in new blood vessel formation after hematoma resorption. This angiogenic effect is mediated by the significant concentrations of vascular endothelial growth factor found in the hematoma. This study identifies an angiogenic cytokine involved in human fracture healing and shows that fracture hematoma is inherently angiogenic. The differences between the in vitro and in vivo findings may explain the phenomenon of interfragmentary hematoma organization and resorption that precedes fracture revascularization.

Ankylosis Due Sequel Of Fracture Of The Mandibular Condyle: Case Report

Directory of Open Access Journals (Sweden)

Alina Alencar Ferreira Gomes

2017-08-01

Full Text Available The Ankylosis of the Temporomandibular joint (TMJ is a disorder of craniofacial complex that results in the merger between the condyle and the mandibular fossa, causing partial or complete immobilization of the mandible. The etiological factors are local and systemic inflammation, infection in the area of TMJ, rheumatic diseases and neoplasms, having the trauma as the main etiologic factor. The traumas are responsible for 31% to 98% of cases of ankylosis. The diagnosis is made from the anamnesis and imaging scans (computed tomography pointing to the union of joint components. The treatment of ankylosis is a big challenge due the high rate of recurrence that can be affected by factors such as type of ankylosis, surgical technique, age of the patient, post-operative physiotherapy and systematic follow-up of the patient. The various forms of treatment require careful analysis of type of ankylosis if it is intra or extra-articular, unilateral or bilateral and if it is bony or fibrous, There is no consensus in current literature regarding the best treatment. The aim of this work is to present through the report of a clinical case, a surgical treatment of Unilateral Temporomandibular joint Ankylosis, due to sequel of condylar fracture with re-establishment of the stomatognathic functions in postoperative follow-up.

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

International Nuclear Information System (INIS)

Silva Luna, O.

1982-01-01

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

Cerebral Venous Air Embolism due to a Hidden Skull Fracture Secondary to Head Trauma

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Ai Hosaka

2015-01-01

Full Text Available Cerebral venous air embolism is sometimes caused by head trauma. One of the paths of air entry is considered a skull fracture. We report a case of cerebral venous air embolism following head trauma. The patient was a 55-year-old man who fell and hit his head. A head computed tomography (CT scan showed the air in the superior sagittal sinus; however, no skull fractures were detected. Follow-up CT revealed a fracture line in the right temporal bone. Cerebral venous air embolism following head trauma might have occult skull fractures even if CT could not show the skull fractures.

Quality assurance of the reactor pressure vessel of nuclear power plants. Determination of the fracture toughness KIC above the ductile-brittle transition region on small test specimens by means of a conformal mapping

International Nuclear Information System (INIS)

Ullrich, G.; Krompholz, K.

1994-01-01

The ''surveillance-programs'' for the determination of the mechanical properties of reactor pressure vessel (RPV) materials, as a function of the neutron dose, include impact and tensile tests for the boiling water reactor; while for pressurized water reactors additional wedge opening load specimens (WOL), for the measurement of the fracture toughness K IC at low temperatures, are utilized. While the Charpy impact toughness gives the total magnitude of energy, which indicates the change of the material state, e.g. the state of embrittlement, the fracture toughness, I IC , gives a base for mechanical calculations. This is of importance for components in which cracks or flaws are assumed. The mechanical analysis, and its relevance to safety assessments, depends on the knowledge of different parameters such as geometry of the structure and flaws, and load history of the structure. Fracture mechanical methods play an important role, if the leak-before-fracture problem is considered. Within the frame work of fracture mechanical methods, only the influence of assumed macroscopic cracks on the structural behaviour can be handled. Flaw formation processes in flaw-free structures, as well as the treatment of short flaws, can not currently be included. In the regime of low and intermediate temperatures (for ferritic and austenitic materials, normally below 400 o C), the rules of linear elastic fracture mechanics (LEFM) and elasto-plastic fracture mechanics (EPFM) are applied, some of which are already part of the code cases. (author) 5 figs., 32 refs

Acetabular Fracture

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Chad Correa

2017-09-01

; they will show a sagittally oriented fracture line at the roof of the acetabulum on axial CT. Lastly, wall fractures should be evaluated with axial CT images. This is because wall fractures have an obliquely oriented fracture line on axial CT images at the roof of the acetabulum, as opposed to the coronal and sagittal fracture lines described with column and transverse fractures, respectively.2 Fractures are organized using the Letournel Classification based on whether the fracture site lies in the anterior or posterior walls and columns of bone. After diagnosis, early surgical intervention is critical in achieving good results.3 The majority of acetabular fractures are repaired by open reduction and internal fixation. Patients with significant osteopenia or communition benefit most from total hip arthroplasty. However, due to the complex nature of these fractures, there is potential for poor outcome regardless of the injury pattern due to contributing factors such as imperfect reduction, osteochondral defects in the acetabulum or femoral head, osteoarthritis, avascular necrosis of the femoral head, sciatic nerve injury and infection.4

A case report of an isolated fracture through the radial bicipital tuberosity

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Kanta Imao

Full Text Available Introduction: Generally, anatomical reduction of shaft fractures through operative treatment is necessary to restore the anatomical relationship of the forearm bones. However, a number of nerves and vessels are located in the proximal radius, which complicates surgery. In this study, we aimed to reduce postoperative complications by using a posterior approach. Presentation of case: We describe an isolated fracture through the radial bicipital tuberosity in a 69-year-old man caused by direct blunt force and our management of the fracture. The patient underwent an operation for the fracture under brachial plexus block. The injury was explored using the posterior approach, and plate fixation was performed after confirming the absence of obstacles to rotation on pronation and supination. One year later, the patient did not have any difficulties in activities of daily living. Discussion: Since an isolated fracture through the radial bicipital tuberosity is more distal than the radial head and neck and more proximal than a common radius diaphysis fracture, we had to consider a different operative approach. The nerve and blood vessels of the forearm, such as the radial nerve and artery, run in a complicated fashion around the proximal radius; thus, we chose the posterior approach because of its simpler surgical technique and lower complication risk, compared with the anterior approach. Conclusion: Surgeons can obtain a favorable treatment result using the posterior approach to the fracture and reduce complications by ensuring with rigid fixation using a locking plate. Keywords: Radial bicipital tuberosity, Posterior approach, Posterior interosseous nerve, Shaft fracture

Structural analysis of the KSTAR vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

In, Sang Ryul; Yoon, Byeong Joo [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-10-01

Structure analysis of the vacuum vessel for the KSTAR tokamak which, is in the end phase of the conceptual design have been performed. Mechanical stresses and deformations of the vessel produced by constant forces due to atmospheric pressure, dead weight, fluid pressure, etc and various transient electromagnetic forces induced during tokamak operations were calculated as well as modal characteristics and buckling properties were investigated. Influences of the temperature gradient and the constraint condition of the support on the thermal stress and deformation of the vessel were analyzed. The thermal stress due to the temperature distribution on the vessel as supplying the N{sub 2} gas of 400 deg C through poloidal channels according to the recent baking concept were calculated. No severe problem in the robustness of the vessel was found when applying the constant pressures on the vessel. However the mechanical stress due to the EM force induced by halo currents flowing on the vessel and the plasma facing components (PFCs) far exceeded the allowable limit. Some reinforcing components should be added on the boundary of the PFC support and the vessel, and that of the vessel support and the vessel. A steep temperature gradient in the vicinity of the inlet and oulet of the heating gas produced a thermal stress much higher than allowable. It is necessary to make the temperature of the vessel as uniform as possible and to develop a new support concept which is flexible enough to accommodate a thermal expansion of a few cm while sufficiently strong to resist mechanical impacts. (author). 5 refs., 41 figs., 9 tabs.

Unstable ductile fracture conditions in upper shelf region

International Nuclear Information System (INIS)

Nakano, Yoshifumi; Kubo, Takahiro

1985-01-01

The phenomenon of unstability of ductile fracture in the upper shelf region of a forged steel for nuclear reactor pressure vessels A508 Cl. 3 was studied with a large compliance apparatus, whose spring constants were 100, 170 and 230 kgf/mm, at the test temperatures of 100, 200 and 300 0 C and at the loading rates of 2, 20 and 200 mm/min in the crosshead speed. The main results obtained are as follows: (1) The fracture modes of the specimens consisted of (a) stable fracture, (b) unstable fracture which leads to a complete fracture rapidly and (c) quasiunstable fracture which does not lead to a complete fracture though a rapid extension of ductile crack takes place. (2) Side groove, high temperature or small spring constant made a ductile crack more unstable. (3) High temperature or large spring constant made the occurrence of quasiunstable fracture easier. (4) Quasiunstable ductile fracture took place before the maximum load, that is, at the J integral value of about 10 kgf/mm. The initiation of a microscopic ductile crack, therefore, seems to lead to quasiunstable fracture. (5) The concept that unstable ductile fracture takes place when Tsub(app) exceeds Tsub(mat) seems applicable only to the case in which unstable ductile fracture takes place after the maximum load has been exceeded. (author)

Stable crack growth during over stressing or proof testing of pressure vessels

International Nuclear Information System (INIS)

Njo, D.H.

1985-07-01

In an earlier study several years ago an attempt has been made by a study group in Switzerland formed by representatives of utilities, manufacturers, inspection agency and the licensing authority to evaluate systematically the advantages and disadvantages of a proof test for a reactor pressure vessel (RPV) taking into account the different aspects involved. In this study, the present day practice in the requirements regarding proof testing in several countries with nuclear power plants were assessed. Using the method of fracture mechanics, taking into account the proof test conditions and material behaviour at hand, qualified quantitative statements were then sought for. One of the important findings of the study was that a sufficiently accurate determination of stable crack growth (SCG) during proof testing is a necessary precondition for a qualified quantitative statement. At that time this requirement could not be met. The aim of the present exercise of collecting available information on SCG was twofold: to get an overview of the present day requirements on pre-service and in-service proof testing of reactor pressure vessels in different countries with NPPs; to collect data on stable crack growth due to over stressing during proof tests to be used in the quantitative assessment, using fracture mechanics methods, and the potential benefit, if any, of proof tests. In the framework of the present study, the originally stipulated aims could not be achieved. It turned out to be a very complex and difficult problem, where various important aspects are little known and could not be assessed without further comprehensive theoretical and experimental investigations

Application of probabilistic fracutre mechanics to allocation of NDT for nuclear pressure vessels

International Nuclear Information System (INIS)

Bergman, M.; Brickstad, B.; Dillstroem, P.; Nilsson, F.

1991-08-01

In order to study whether there are considerable differences in fracture probability between different regions in a reactor pressure vessel a limited probabilistic fracture mechanics (PFM) study is carried out. Two different regions (crack geometries) are considered and the fracture and leakage probabilities are calculated for a number of load cases. The loading is assumed to be deterministic while most of the other quantities are assumed to be of random character. The fracture probabilities are very dependent of assumption made for the fracture toughness distribution, but the mutual order of the fracture probabilities of the two regions seemed to be relatively unaffected by this. Of the transients considered, the cold overpressurization event is by far the most dangerous even. A sensitivity analysis shows however that this result is heavily dependent of the transition temperature of the material. The leakage probabilities are in most cases much lower than the fracture probabilities indicating that consequence considerations are not very important for NDT allocation purpose. (au)

Fracture toughness evaluation of steels through master curve approach using Charpy impact specimens

International Nuclear Information System (INIS)

Chatterjee, S.; Sriharsha, H.K.; Shah, Priti Kotak

2007-01-01

The master curve approach can be used for the evaluation of fracture toughness of all steels which exhibit a transition between brittle to ductile mode of fracture with increasing temperature, and to monitor the extent of embrittlement caused by metallurgical damage mechanisms. This paper details the procedure followed to evaluate the fracture toughness of a typical ferritic steel used as material for pressure vessels. The potential of master curve approach to overcome the inherent limitations of the estimation of fracture toughness using ASME Code reference toughness is also illustrated. (author)

Heavy Section Steel Technology Program. Part II. Intermediate vessel testing

International Nuclear Information System (INIS)

Whitman, G.D.

1975-01-01

The testing of the intermediate pressure vessels is a major activity under the Heavy Section Steel Technology Program. A primary objective of these tests is to develop or verify methods of fracture prediction, through the testing of selected structures and materials, in order that a valid basis can be established for evaluating the serviceability and safety of light-water reactor pressure vessels. These vessel tests were planned with sufficiently specific objectives that substantial quantitative weight could be given to the results. Each set of testing conditions was chosen so as to provide specific data by which analytical methods of predicting flaw growth, and in some cases crack arrest, could be evaluated. Every practical effort was made to assure that results would be relevant to some aspect of real reactor pressure vessel performance through careful control of material properties, selection of test temperatures, and design of prepared flaws. 5 references

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)

Bilateral femoral neck fractures following pelvic irradiation

International Nuclear Information System (INIS)

Mitsuda, Kenji; Nishi, Hosei; Oba, Hiroshi

1977-01-01

Over 300 cases of femoral neck fractures following radiotherapy for intrapelvic malignant tumor have been reported in various countries since Baensch reported this disease in 1927. In Japan, 40 cases or so have been reported, and cases of bilateral femoral neck fractures have not reached to ten cases. The authors experienced a case of 75 year-old female who received radiotherapy for cancer of the uterus, and suffered from right femoral neck fracture 3 months after and left femoral neck fracture one year and half after. As clinical symptoms, she had not previous history of trauma in bilateral femurs, but she complained of a pain in a hip joint and of gait disturbance. The pain in left femoral neck continued for about one month before fracture was recognized with roentgenogram. As histopathological findings, increase of fat marrow, decrease of bone trabeculae, and its marked degeneration were recognized. Proliferation of some blood vessels was found out, but thickness of the internal membrane and thrombogenesis were not recognized. Treatment should be performed according to degree of displacement of fractures. In this case, artificial joint replacement surgery was performed to the side of fracture of this time, because this case was bilateral femoral neck fractures and the patient had received artificial head replacement surgery in the other side of fracture formerly. (Tsunoda, M.)

Development of a plastic fracture methodology. Final report

International Nuclear Information System (INIS)

Kanninen, M.F.; Hahn, G.T.; Broek, D.; Stonesifer, R.B.; Marschall, C.W.; Abou-Sayed, I.S.; Zahoor, A.

1981-03-01

A number of candidate fracture criteria were investigated to determine the basis for plastic fracture mechanics assessments of nuclear pressure vessels and other components exhibiting fully ductile behavior. The research was comprised of an integrated combination of stable crack growth experiments and elastic-plastic finite element analyses. The results demonstrated that many different fracture criteria can be used as the basis of a resistance curve approach to predicting stable crack growth and fracture instability. All have some disadvantages and none is completely unacceptable. On balance, the best criteria were found to be the J-integral for initiation and limited amounts of stable crack growth and the local crack tip opening angle for extended amounts of stable growth. A combination of the two, which may preserve the advantages of each while reducing their disadvantages, was also suggested by these results

Development of a plastic fracture methodology. Final report

Energy Technology Data Exchange (ETDEWEB)

Kanninen, M.F.; Hahn, G.T.; Broek, D.; Stonesifer, R.B.; Marschall, C.W.; Abou-Sayed, I.S.; Zahoor, A.

1981-03-01

A number of candidate fracture criteria were investigated to determine the basis for plastic fracture mechanics assessments of nuclear pressure vessels and other components exhibiting fully ductile behavior. The research was comprised of an integrated combination of stable crack growth experiments and elastic-plastic finite element analyses. The results demonstrated that many different fracture criteria can be used as the basis of a resistance curve approach to predicting stable crack growth and fracture instability. All have some disadvantages and none is completely unacceptable. On balance, the best criteria were found to be the J-integral for initiation and limited amounts of stable crack growth and the local crack tip opening angle for extended amounts of stable growth. A combination of the two, which may preserve the advantages of each while reducing their disadvantages, was also suggested by these results.

Advances in crack-arrest technology for reactor pressure vessels

International Nuclear Information System (INIS)

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

1988-01-01

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

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)

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 ...

Limit analysis and design of containment vessels

International Nuclear Information System (INIS)

Save, M.

1984-01-01

In the introduction, the theory of plastic analysis of shells is briefly recalled. Minimum-volume design for assigned load factor at plastic collapse is then considered and optimality criteria are derived for plates and shells of continuously varying or piecewise-constant thickness. In the first part, containers made of metal are examined. Analytical and numerical limit analysis solutions and corresponding experimental results are considered for various types of vessels, including intersecting shells. Attention is given to experimental post-yield behavior. Some tests up to fracture are discussed. New theoretical and experimental results of limit analysis of stiffened cylindrical vessels are presented, in which reinforcing rings are treated as discrete structural element (no smearing out) and due account is taken of their strong curvature. Cases of collapse by instability under internal pressure are pointed out. Minimum-volume design of circular plates and cylindrical shells is then formulated and various examples are presented of sandwich and solid metal structures. Containers of piecewise-constant thickness are given particular attention. Available experimental evidence on minimum-volume design of plates and shells is reviewed and commented upon. The second part deals with reinforced concrete vessels. Cylindrical containers are studied, from both points of view of limit analysis and of limit design with minimum volume of reinforcement. The practical use of the latter solutions is discussed. A third part reviews other loading cases (including cyclic and impact loads) and gives indications on corresponding theories, formulations and solution methods. The last part is devoted to a discussion of the limitations of the methods presented, within the frame of the 'limit states' design philosophy, which is first briefly recalled. Considerations on further research in the field conclude the paper. (orig.)

Elastodynamic fracture analyses of large crack-arrest experiments

International Nuclear Information System (INIS)

Bass, B.R.; Pugh, C.E.; Walker, J.K.

1985-01-01

Results obtained to date show that the essence of the run-arrest events, including dynamic behavior, is being modeled. Refined meshes and optimum solution algorithms are important parameters in elastodynamic analysis programs to give sufficient resolution to the geometric and time-dependent aspects of fracture analyses. Further refinements in quantitative representation of material parameters and the inclusion of rate dependence through viscoplastic modeling is expected to give an even more accurate basis for assessing the fracture behavior of reactor pressure vessels under PTS and other off-normal loading conditions

The analysis of reactor vessel surveillance program data

International Nuclear Information System (INIS)

Norris, E.B.

1979-01-01

Commercial nuclear power reactor vessel surveillance programs are provided by the reactor supplier and are designed to meet the requirements of ASTM Method E 185. (3). Each surveillance capsule contains sets of Charpy V-notch (Csub(v)) specimens representing selected materials from the vessel beltline region and some reference steel, tension test specimens machined from selected beltline materials, temperature monitors, and neutron flux dosimeters. Surveillance capsules may also contain fracture mechanics specimens machined from selected vessel beltline materials. The major steps in the conduct of a surveillance program include (1) the testing of the surveillance specimens to determine the exposure conditions at the capsule location and the resulting embrittlement of the vessel steel, (2) the extrapolation of the capsule results to the pressure vessel wall, and (3) the determination of the heatup and cooldown limits for normal, upset, and test operation. This paper will present data obtained from commercial light water reactor surveillance programs to illustrate the methods of analysis currently in use at Southwest Research Institute and to demonstrate some of the limitations imposed by the data available. Details concerning the procedures for testing the surveillance capsule specimens will not be included because they are considered to be outside of the scope of this paper

Evaluation of fracture mechanics analyses used in RPV integrity assessment regarding brittle fracture

International Nuclear Information System (INIS)

Moinereau, D.; Faidy, C.; Valeta, M.P.; Bhandari, S.; Guichard, D.

1997-01-01

Electricite de France has conducted during these last years some experimental and numerical research programmes in order to evaluate fracture mechanics analyses used in nuclear reactor pressure vessels structural integrity assessment, regarding the risk of brittle fracture. These programmes included cleavage fracture tests on large scale cladded specimens containing subclad flaws with their interpretations by 2D and 3D numerical computations, and validation of finite element codes for pressurized thermal shocks analyses. Four cladded specimens made of ferritic steel A508 C13 with stainless steel cladding, and containing shallow subclad flaws, have been tested in four point bending at very low temperature in order to obtain cleavage failure. The specimen failure was obtained in each case in base metal by cleavage fracture. These tests have been interpreted by two-dimensional and three-dimensional finite element computations using different fracture mechanics approaches (elastic analysis with specific plasticity corrections, elastic-plastic analysis, local approach to cleavage fracture). The failure of specimens are conservatively predicted by different analyses. The comparison between the elastic analyses and elastic-plastic analyses shows the conservatism of specific plasticity corrections used in French RPV elastic analyses. Numerous finite element calculations have also been performed between EDF, CEA and Framatome in order to compare and validate several fracture mechanics post processors implemented in finite element programmes used in pressurized thermal shock analyses. This work includes two-dimensional numerical computations on specimens with different geometries and loadings. The comparisons show a rather good agreement on main results, allowing to validate the finite element codes and their post-processors. (author). 11 refs, 24 figs, 3 tabs

Evaluation of fracture mechanics analyses used in RPV integrity assessment regarding brittle fracture

Energy Technology Data Exchange (ETDEWEB)

Moinereau, D [Electricite de France, Dept. MTC, Moret-sur-Loing (France); Faidy, C [Electricite de France, SEPTEN, Villeurbanne (France); Valeta, M P [Commisariat a l` Energie Atomique, Dept. DMT, Gif-sur-Yvette (France); Bhandari, S; Guichard, D [Societe Franco-Americaine de Constructions Atomiques (FRAMATOME), 92 - Paris-La-Defense (France)

1997-09-01

Electricite de France has conducted during these last years some experimental and numerical research programmes in order to evaluate fracture mechanics analyses used in nuclear reactor pressure vessels structural integrity assessment, regarding the risk of brittle fracture. These programmes included cleavage fracture tests on large scale cladded specimens containing subclad flaws with their interpretations by 2D and 3D numerical computations, and validation of finite element codes for pressurized thermal shocks analyses. Four cladded specimens made of ferritic steel A508 C13 with stainless steel cladding, and containing shallow subclad flaws, have been tested in four point bending at very low temperature in order to obtain cleavage failure. The specimen failure was obtained in each case in base metal by cleavage fracture. These tests have been interpreted by two-dimensional and three-dimensional finite element computations using different fracture mechanics approaches (elastic analysis with specific plasticity corrections, elastic-plastic analysis, local approach to cleavage fracture). The failure of specimens are conservatively predicted by different analyses. The comparison between the elastic analyses and elastic-plastic analyses shows the conservatism of specific plasticity corrections used in French RPV elastic analyses. Numerous finite element calculations have also been performed between EDF, CEA and Framatome in order to compare and validate several fracture mechanics post processors implemented in finite element programmes used in pressurized thermal shock analyses. This work includes two-dimensional numerical computations on specimens with different geometries and loadings. The comparisons show a rather good agreement on main results, allowing to validate the finite element codes and their post-processors. (author). 11 refs, 24 figs, 3 tabs.

Ultimate state substantiation and methods of the life extension for pwr pressure vessel

International Nuclear Information System (INIS)

Troshchenko, V.T.; Pokrovsky, V.V.; Yasniy, P.V.; Kaplunenko, V.G.; Fedorov, V.G.; Dragunov, Y.G.; Timofeev, B.T.

1991-01-01

A model of brittle fracture was developed for structural materials with cracks under cyclic loading. The authors proposed a method of lifetime evaluation for structural elements (including pressure vessels) using the brittle fracture criteria and taking into account the stage of unstable crack growth. The influence of preliminary mechanical overloading of the metal with a crack in plastic state (warm prestress) upon fracture toughness characteristics of the parent metal and that of the weld was studied. The tests were carried out on 25 to 150 mm thick compact and three-point bend specimens with a long through-the-thickness crack and a short semi-elliptical one. (author)

Evidence Report: Risk of Bone Fracture due to Spaceflight-Induced Changes to Bone

Science.gov (United States)

Sibonga, Jean D.; Evans, Harlan J.; Smith, Scott A.; Spector, Elisabeth R.; Yardley, Greg; Myer, Jerry

2017-01-01

Given that spaceflight may induce adverse changes in bone ultimate strength with respect to mechanical loads during and post-mission, there is a possibility a fracture may occur for activities otherwise unlikely to induce fracture prior to initiating spaceflight.

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

International Nuclear Information System (INIS)

Eichhorn, F.; Hirsch, P.; Langenbahn, H.W.; Wubbels, B.

1978-01-01

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

Spontaneous rib fractures.

Science.gov (United States)

Katrancioglu, Ozgur; Akkas, Yucel; Arslan, Sulhattin; Sahin, Ekber

2015-07-01

Other than trauma, rib fracture can occur spontaneously due to a severe cough or sneeze. In this study, patients with spontaneous rib fractures were analyzed according to age, sex, underlying pathology, treatment, and complications. Twelve patients who presented between February 2009 and February 2011 with spontaneous rib fracture were reviewed retrospectively. The patients' data were evaluated according to anamnesis, physical examination, and chest radiographs. The ages of the patients ranged from 34 to 77 years (mean 55.91 ± 12.20 years), and 7 (58.4%) were male. All patients had severe cough and chest pain. The fractures were most frequently between 4th and 9th ribs; multiple rib fractures were detected in 5 (41.7%) patients. Eight (66.7%) patients had chronic obstructive pulmonary disease, 2 (16.7%) had bronchial asthma, and 2 (16.7%) had osteoporosis. Bone densitometry revealed a high risk of bone fracture in all patients. Patients with chronic obstructive pulmonary disease or bronchial asthma had been treated with high-dose steroids for over a year. Spontaneous rib fracture due to severe cough may occur in patients with osteoporosis, chronic obstructive pulmonary disease, or bronchial asthma, receiving long-term steroid therapy. If these patients have severe chest pain, chest radiography should be performed to check for bone lesions. © The Author(s) 2015.

PWR reactor pressure vessel failure probabilities

International Nuclear Information System (INIS)

Dufresne, J.; Lanore, J.M.; Lucia, A.C.; Elbaz, J.; Brunnhuber, R.

1980-05-01

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

Determination of the failure probability in the weld region of ap-600 vessel for transient condition

International Nuclear Information System (INIS)

Wahyono, I.P.

1997-01-01

Failure probability in the weld region of AP-600 vessel was determined for transient condition scenario. The type of transient is increase of the heat removal from primary cooling system due to sudden opening of safety valves or steam relief valves on the secondary cooling system or the steam generator. Temperature and pressure in the vessel was considered as the base of deterministic calculation of the stress intensity factor. Calculation of film coefficient of the convective heat transfers is a function of the transient time and water parameter. Pressure, material temperature, flaw depth and transient time are variables for the stress intensity factor. Failure probability consideration was done by using the above information in regard with the flaw and probability distributions of Octavia II and Marshall. Calculation of the failure probability by probability fracture mechanic simulation is applied on the weld region. Failure of the vessel is assumed as a failure of the weld material with one crack which stress intensity factor applied is higher than the critical stress intensity factor. VISA II code (Vessel Integrity Simulation Analysis II) was used for deterministic calculation and simulation. Failure probability of the material is 1.E-5 for Octavia II distribution and 4E-6 for marshall distribution for each transient event postulated. The failure occurred at the 1.7th menit of the initial transient under 12.53 ksi of the pressure

Modified Risdon approach using periangular incision in surgical treatment of subcondylar mandibular fractures

Directory of Open Access Journals (Sweden)

Nikolić Živorad S.

2016-01-01

Full Text Available Introduction. No consensus has been reached yet on the surgical approach for treatment of condylar fractures. Objective. The aim of this study was to present modified Risdon approach (without facial nerve identification in the treatment of subcondylar mandibular fractures. Method. This is a retrospective study of a period 2005-2012. During this seven-year period, 25 condylar mandibular fractures in 22 men and three women (19-68 years old were treated by modified Risdon approach without identifying the facial nerve. The main inclusion criterion was subcondylar fracture according to Lindahl classification. Results. No additional morbidity related to postoperative complications, such as infection or salivary fistula, was observed in this series. Only two (8% patients developed temporary weakness of the marginal branch of the facial nerve, which resolved six weeks postoperatively. Each patient achieved good mouth opening postoperatively. Scar was camouflaged in the first cervical wrinkle. Two patients developed temporomandibular joint dysfunction. No patient had postoperative occlusal disturbance. In all of the patients good aesthetic result was achieved in a two-year follow-up. Conclusion. In comparison with techniques described in the literature, the main advantages of the modified Risdon approach are the following: no need for facial vessels identification; direct, fast, and safe approach to mandibular angle and subcondylar region; relatively simple surgical technique and good cosmetic result - due to aesthetically placed incision. This approach could be recommended for subcondylar fracture as a simplified and safe procedure. [Projekat Ministarstva nauke Republike Srbije, br. 175075

The behavior of shallow flaws in reactor pressure vessels

International Nuclear Information System (INIS)

Rolfe, S.T.

1991-11-01

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

Percutaneous vertebroplasty performed with an 18 G needle for the treatment of severe compression fracture of cervical vertebral body due to malignancy

International Nuclear Information System (INIS)

Chen Long; Ni Caifang; Wang Zhentang; Liu Yizhi; Jin Yonghai; Zhu Xiaoli; Zou Jianwei; Xiao Xiangsheng

2010-01-01

Objective: To investigate the clinical feasibility and efficacy of percutaneous vertebroplasty performed with an 18G needle for the treatment of severe compression fracture of cervical vertebral body due to malignancy. Methods: During the period of 2006-2010 percutaneous vertebroplasty was performed in 10 patients with severe compression fracture of cervical vertebral body due to metastatic lesions. A total of 12 diseased vertebral bodies were detected, which distributed in the C 4 (n = 3), C 5 (n = 3), C 6 (n = 4) and C 7 (n = 2) vertebral bodies. Under DSA guidance an 18G needle was punctured into the target vertebral body and then polymethylmethacrylate bone cement was injected in. A follow-up lasting for one month was conducted. Results: The technical success of both needle puncturing and bone cement injection was achieved in all patients. The mean amount of bone cement injected in each diseased vertebra was 2.2 ml(1.5-3.2)ml. Marked pain relief was quickly obtained in al1 10 patients. No major complications occurred in this series, except for asymptomatic bone cement leaking around vertebra which appeared in 4 vertebral bodies. Conclusion: Percutaneous vertebroplasty, which is performed with an 18G needle, is a safe and effective technique for the treatment of severe compression fracture of cervical vertebral body due to malignancy. (authors)

In-vessel core debris retention experiments. Final report

International Nuclear Information System (INIS)

1998-10-01

The in-vessel cooling experimental program (Phase 1 and 2) was motivated by the survivability of the TMI lower vessel head during the TMI-2 accident. During that accident, molten debris relocation into the water filled lower head resulted in a localized hot spot in the lower head, but no lower head failure occurred. A postulated set of mechanisms which could be involved in and responsible for the survivability of the TMI lower head were identified and experimentally investigated as part of this program. These mechanisms included: the formation of a gap (contact resistance) between the relocated and frozen debris and the vessel wall was a key aspect of the in-vessel cooling mechanism; wall heatup due to the relocated debris in the presence of wall stress due to a pressure gradient across the vessel wall; gap growth due to a lack of debris adherence to the vessel wall and material creep of the heated vessel wall; and the potential for enhanced wall cooling due to gap growth. Each of these postulated mechanisms was investigated in this experimental program. This report summarizes the several insights and conclusions that were obtained from this experimental program. This report documents the entire set of five experiments completed in Phase 2 of this experimental program. Results from the Phase 1 effort were used to plan and select the Phase 2 test matrix. Conclusions from the Phase 1 and 2 experiments are identified and recommendations for future work are provided

Understanding the evolution of channeling and fracturing in porous medium due to fluid flow.

Science.gov (United States)

Turkaya, Semih; Toussaint, Renaud; Kvalheim Eriksen, Fredrik; Daniel, Guillaume; Langliné, Olivier; Grude Flekkøy, Eirik; Jørgen Måløy, Knut

2017-04-01

Fluid induced brittle deformation of porous medium is a phenomenon commonly present in everyday life. From an espresso machine to volcanoes, from food industry to construction, it is possible to see traces of this phenomenon. In this work, analogue models are developed in a linear geometry, with confinement and at low porosity to study the instabilities that occur during fast motion of fluid in dense porous materials: fracturing, fingering, and channeling. We study these complex fluid/solid mechanical systems - in a rectangular Hele-Shaw cell with three closed boundaries and one semi-permeable boundary - using two monitoring techniques: optical imaging using a high speed camera (1000 fps), high frequency resolution accelerometers and piezoelectrical sensors. Additionally, we develop physical models rendering for the fluid mechanics in the channels and the propagation of microseismic waves around the fracture. We then compare a numerical resolution of this physical system with the observed experimental system. In the analysis phase, we compute the power spectrum of the acoustic signal in time windows of 5 ms, recorded by shock accelerometers Brüel & Kjaer 4374 (Frq. Range 1 Hz - 26 kHz) with 1 MHz sampling rate. The evolution of the power spectrum is compared with the optical recordings. These peaks on the spectrum are strongly influenced by the size and branching of the channels, compaction of the medium, vibration of air in the pores and the fundamental frequency of the plate. Furthermore, the number of these stick-slip events, similar to the data obtained in hydraulic fracturing operations, follows a Modified Omori Law decay with an exponent p value around 0.5. An analytical model of overpressure diffusion predicting p = 0.5 and two other free parameters of the Omori Law (prefactor and origin time) is developed. The spatial density of the seismic events, and the time of end of formation of the channels can also be predicted using this developed model. Different

High revision rate but good healing capacity of atypical femoral fractures. A comparison with common shaft fractures.

Science.gov (United States)

Schilcher, Jörg

2015-12-01

Healing of complete, atypical femoral fractures is thought to be impaired, but the evidence is weak and appears to be based on the delayed healing observed in patients with incomplete atypical fractures. Time until fracture healing is difficult to assess, therefore we compared the reoperation rates between women with complete atypical femoral fractures and common femoral shaft fractures. We searched the orthopaedic surgical registry in Östergötland County for patients with subtrochanteric and femoral shaft fractures (ICD-10 diagnosis codes S72.2, S72.3 and M84.3F) between January 1st 2007 and December 31st 2013. Out of 895 patients with surgically treated femoral shaft fractures, 511 were women 50 years of age or older. Among these we identified 24 women with atypical femoral shaft fractures, and 71 with common shaft fractures. Reoperations were performed in 6 and 5 patients, respectively, odds ratio 4.4 (95% CI 1.2 to 16.1). However, 5 reoperations in the atypical fracture group could not be ascribed to poor healing. In 3 patients the reoperation was due to a new fracture proximal to a standard intramedullary nail. In 2 patients the distal locking screws were removed due to callus formation that was deemed incomplete 5 months post-operatively. The one patient with poor healing showed faint callus formation at 5 months when the fracture was dynamised and callus remained sparse at 11 months. Among patients with common shaft fractures, 2 reoperations were performed to remove loose screws, 2 because of peri-implant fractures and 1 reoperation due to infection. Reoperation rates in patients with complete atypical femoral fractures are higher than in patients with common shaft fractures. The main reason for failure was peri-implant fragility fractures which might be prevented with the use of cephalomedullary nails at the index surgery. Fracture healing however, seems generally good. A watchful waiting approach is advocated in patients with fractures that appear to

Study on the application of thickened welds without post weld heat treatment for containment vessels

International Nuclear Information System (INIS)

Takeuchi, T.; Fukaya, T.; Sato, M.; Takano, G.

1978-01-01

As material for containment vessels, SGV49 steel plates are mainly used. However, those used for this purpose are limited in thickness to smaller than 38 mm. This is because the present standard requires welds thicker than 38 mm to be subjected to post weld heat treatment but operation on the site is practically difficult. In the case of 3-loop containment vessels of pressurized water type reactors, use of 38 mm SGV49 brings an increase in their height and this is disadvantageous from a seismic viewpoint. Therefore, use of 45 mm-thick steel material has become necessary in order to increase design internal pressure and reduce the height of the vessels. To investigate the propriety of the use of 45 mm-thick SGV49 for this purpose without post weld heat treatment we investigated the basic performances of base metal and welded joints. We also conducted large-scale embrittlement fracture tests (CT test, deep notch test, wide plate tensile test and ESSO test) in order to examine whether welds not subjected to post weld heat treatment are safe against embrittlement fracture under the operating conditions of the vessels. The results proved that the welds of SGV49 steel plates are safe enough under the operating conditions. (author)

Shallow crack effect on brittle fracture of RPV during pressurised thermal shock

International Nuclear Information System (INIS)

Ikonen, K.

1995-12-01

This report describes the study on behaviour of postulated shallow surface cracks in embrittled reactor pressure vessel subjected to pressurised thermal shock loading in an emergency core cooling. The study is related to the pressure vessel of a VVER-440 type reactor. Instead of a conventional fracture parameter like stress intensity factor or J integral the maximum principal stress distribution on a crack tip area is used as a fracture criteria. The postulated cracks locate circumferentially at the inner surface of the reactor pressure wall and they penetrate the cladding layer and open to the inner surface. Axisymmetric and semielliptical crack shapes were studied. Load is formed of an internal pressure acting also on crack faces and of a thermal gradient in the pressure vessel wall. Physical properties of material and loading data correspond real conditions in VVER-440 RPV. The study was carried out by making lot of 2D- and 3D- finite element calculations. Analysing principles and computer programs are explained. Except of studying the shallow crack effect, one objective of the study has also been to develop further expertise and the in-house developed computing system to make effectively elastic-plastic fracture mechanical analyses for real structures under complicated loads. Though the study concerns VVER-440 RPV, the results are of more general interest especially related to thermal loads. (orig.) (11 refs.)

An holistic approach to the problem of reactor ageing. [Pressure vessel embrittlement

Energy Technology Data Exchange (ETDEWEB)

Phythian, W.; McElroy, R.; Druce, S.; Kovan, D. (AEA Reactor Services, Harwell (United Kingdom))

1992-12-01

Understanding the process of ageing in reactors is essential to extending their lives beyond original design. To present a sound case -particularly regarding the level of embrittlement in reactor vessels due to radiation damage - an integrated approach using advanced assessment tools is needed. The techniques developed for the purpose involve, on the microscopic level, advanced neutron dosimetry and high resolution measurement techniques (eg advanced electron beam techniques and small angle neutron scattering) with which an analysis can be done of the radiation damage and the microstructural state of the steel test procedures (tensile, fracture toughness and Charpy impact) on standard and sub-sized specimens, the extent of radiation degradation can be characterised. finally, it is possible to predict how the degradation will evolve using physically-based models of embrittlement. (Author).

Pressure Vessel Steel Research: Belgian Activities

International Nuclear Information System (INIS)

Van Walle, E.; Fabry, A.; Ait Abderrahim, H.; Chaouadi, R.; D'hondt, P.; Puzzolante, J.L.; Van de Velde, J.; Van Ransbeeck, T.; Gerard, R.

1994-03-01

A review of the Belgian research activities on Nuclear Reactor Pressure Vessel Steels (RPVS) and on related Neutron Dosimetry Aspects is presented. Born out of the surveillance programmes of the Belgian nuclear power plants, this research has lead to the development of material saving techniques, like reconstitution and miniaturization, and to improved neutron dosimetry techniques. A physically- justified RPVS fracture toughness indexation methodology, supported by micro-mechanistic modelling, is based on the elaborate use of the instrumented Charpy impact signal. Computational tools for neutron dosimetry allow to reduce the uncertainties on surveillance capsule fluences significantly

Pressure Vessel Steel Research: Belgian Activities

Energy Technology Data Exchange (ETDEWEB)

Van Walle, E; Fabry, A; Ait Abderrahim, H; Chaouadi, R; D` hondt, P; Puzzolante, J L; Van de Velde, J; Van Ransbeeck, T [Centre d` Etude de l` Energie Nucleaire, Mol (Belgium); Gerard, R [TRACTEBEL, Brussels (Belgium)

1994-03-01

A review of the Belgian research activities on Nuclear Reactor Pressure Vessel Steels (RPVS) and on related Neutron Dosimetry Aspects is presented. Born out of the surveillance programmes of the Belgian nuclear power plants, this research has lead to the development of material saving techniques, like reconstitution and miniaturization, and to improved neutron dosimetry techniques. A physically- justified RPVS fracture toughness indexation methodology, supported by micro-mechanistic modelling, is based on the elaborate use of the instrumented Charpy impact signal. Computational tools for neutron dosimetry allow to reduce the uncertainties on surveillance capsule fluences significantly.

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Integral experiments on in-vessel coolability and vessel creep: results and analysis of the FOREVER-C1 test

Energy Technology Data Exchange (ETDEWEB)

Sehgal, B.R.; Nourgaliev, R.R.; Dinh, T.N.; Karbojian, A. [Division of Nuclear Power Safety, Royal Institute of Technology, Drottning Kristinas Vaeg., Stockholm (Sweden)

1999-07-01

This paper describes the FOREVER (Failure Of REactor VEssel Retention) experimental program, which is currently underway at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS). The objectives of the FOREVER experiments are to obtain data and develop validated models (i) on the melt coolability process inside the vessel, in the presence of water (in particular, on the efficacy of the postulated gap cooling to preclude vessel failure); and (ii) on the lower head failure due to the creep process in the absence of water inside and/or outside the lower head. The paper presents the experimental results and analysis of the first FOREVER-C1 test. During this experiment, the 1/10th scale pressure vessel, heated to about 900degC and pressurized to 26 bars, was subjected to creep deformation in a non-stop 24-hours test. The vessel wall displacement data clearly shows different stages of the vessel deformation due to thermal expansion, elastic, plastic and creep processes. The maximum displacement was observed at the lowermost region of the vessel lower plenum. Information on the FOREVER-C1 measured thermal characteristics and analysis of the observed thermal and structural behavior is presented. The coupled nature of thermal and mechanical processes, as well as the effect of other system conditions (such as depressurization) on the melt pool and vessel temperature responses are analyzed. (author)

Influence of structures on fracture and fracture toughness of cemented tungsten carbides

International Nuclear Information System (INIS)

Zhao, W.; Zhang, X.

1987-01-01

A study was made of the influence of structures on fracture and fracture toughness of cemented tungsten carbides with different compositions and grain sizes. The measurement of the fracture toughness of cemented tungsten carbide was carried out using single edge notched beam. The microstructural parameters and the proportion for each fracture mode on the fracture surface were obtained. The brittle fracture of the alloy is mainly due to the interfacial decohesion fracture following the interface of the carbide crystals. It has been observed that there are localized fractures region ahead of the crack tip. The morphology of the crack propagation path as well as the slip structure in the cobalt phase of the deformed region have been investigated. In addition, a study of the correlation between the plane strain fracture toughness and microstructural parameters, such as mean free path of the cobalt phase, tungsten carbide grain size and the contiguity of tungsten carbide crystals was also made

Shallow-crack toughness results for reactor pressure vessel steel

International Nuclear Information System (INIS)

Theiss, T.J.; Shum, D.K.M.; Rolfe, S.T.

1992-01-01

The Heavy Section Steel Technology Program (HSST) is investigating the influence of flaw depth on the fracture toughness of reactor pressure vessel (RPV) steel. To complete this investigation, techniques were developed to determine the fracture toughness from shallow-crack specimens. A total of 38 deep and shallow-crack tests have been performed on beam specimens about 100 mm deep loaded in 3-point bending. Two crack depths (a ∼ 50 and 9 mm) and three beam thicknesses (B ∼ 50, 100, and 150 mm) have been tested. Techniques were developed to estimate the toughness in terms of both the J-integral and crack-tip opening displacement (CTOD). Analytical J-integral results were consistent with experimental J-integral results, confirming the validity of the J-estimation schemes used and the effect of flaw depth on fracture toughness. Test results indicate a significant increase in the fracture toughness associated with the shallow flaw specimens in the lower transition region compared to the deep-crack fracture toughness. There is, however, little or no difference in toughness on the lower shelf where linear-elastic conditions exist for specimens with either deep or shallow flaws. The increase in shallow-flaw toughness compared with deep-flaw results appears to be well characterized by a temperature shift of 35 degree C

Application of high strength steel to nuclear reactor containment vessel

International Nuclear Information System (INIS)

Susukida, H.; Sato, M.; Takano, G.; Uebayashi, T.; Yoshida, K.

1976-01-01

Nuclear reactor containment vessels are becoming larger in size with the increase in the power generating capacity of nuclear power plants. For example, a containment vessel for a PWR power plant with an output of 1,000 MWe becomes an extremely large one if it is made of the conventional JIS SGV 49 (ASTM A 516 Gr. 70) steel plates less than 38 mm in thickness. In order to design the steel containment vessel within the conventional dimensional range, therefore, it is necessary to use a high strength steel having a higher tensile strength than SGV 49 steel, good weldability and a higher fracture toughness and moreover, possessing satisfactory properties without undergoing post-weld heat treatment. The authors conducted a series of verification tests on high strength steel developed by modifying the ASTM A 543 Grade B Class 1 steel with a view to adopting it as a material for the nuclear reactor containment vessels. As the result of evaluation of the test results from various angles, we confirmed that the high strength steel is quite suitable for the manufacture of nuclear reactor containment vessels. (auth.)

[Osteoporosis fracture in a male patient secondary to hypogonadism due to androgen deprivation treatment for prostate cancer].

Science.gov (United States)

Verdú Solans, J; Roig Grau, I; Almirall Banqué, C

2014-01-01

A 84 year-old patient, in therapy with androgen deprivation during the last 5 years due a prostate cancer, is presented with a osteoporotic fracture of the first lumbar vertebra. The pivotal role of the primary care physician, in the prevention of the osteoporosis secondary to the hypogonadism in these patients, is highlighted. Copyright © 2012 Sociedad Española de Médicos de Atención Primaria (SEMERGEN). Publicado por Elsevier España. All rights reserved.

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

Energy Technology Data Exchange (ETDEWEB)

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

1978-06-01

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

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

International Nuclear Information System (INIS)

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

1978-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Falsire: CSNI project for fracture analyses of large-scale international reference experiments (Phase 1). Comparison report

International Nuclear Information System (INIS)

1994-01-01

A summary of the recently completed Phase I of the Project for Fracture Analysis of Large-Scale International Reference Experiments (Project FALSIRE) is presented. Project FALSIRE was created by the Fracture Assessment Group (FAG) of Principal Working Group No. 3 (PWG/3) of the OECD/NEA Committee on the Safety of Nuclear Installations (CSNI), formed to evaluate fracture prediction capabilities currently used in safety assessments of nuclear vessel components. The aim of the Project FALSIRE was to assess various fracture methodologies through interpretive analyses of selected large-scale fracture experiments. The six experiments used in Project FALSIRE (performed in the Federal Republic of Germany, Japan, the United Kingdom, and the U.S.A.) were designed to examine various aspects of crack growth in reactor pressure vessel (RPV) steels under pressurized-thermal-shock (PTS) loading conditions. The analysis techniques employed by the participants included engineering and finite-element methods, which were combined with Jr fracture methodology and the French local approach. For each experiment, analysis results provided estimates of variables such as crack growth, crack-mouth-opening displacement, temperature, stress, strain, and applied J and K values. A comparative assessment and discussion of the analysis results are presented; also, the current status of the entire results data base is summarized. Some conclusions concerning predictive capabilities of selected ductile fracture methodologies, as applied to RPVs subjected to PTS loading, are given, and recommendations for future development of fracture methodologies are made

Survey on application of probabilistic fracture mechanics approach to nuclear piping

International Nuclear Information System (INIS)

Kashima, Koichi

1987-01-01

Probabilistic fracture mechanics (PFM) approach is newly developed as one of the tools to evaluate the structural integrity of nuclear components. This report describes the current status of PFM studies for pressure vessel and piping system in light water reactors and focuses on the investigations of the piping failure probability which have been undertaken by USNRC. USNRC reevaluates the double-ended guillotine break (DEGB) of rector coolant piping as a design basis event for nuclear power plant by using the PFM approach. For PWR piping systems designed by Westinghouse, two causes of pipe break are considered: pipe failure due to the crack growth and pipe failure indirectly caused by failure of component supports due to an earthquake. PFM approach shows that the probability of DEGB from either cause is very low and that the effect of earthquake on pipe failure can be neglected. (author)

Material Fracture Characterization and Toughness Improving Technology Developments

International Nuclear Information System (INIS)

Lee, Bong Sang; Kim, M. C.; Lee, H. J. and others

2005-04-01

Reactor pressure boundary components including pressure vessel and piping are facing a severe aging condition that can degrade the physical-mechanical properties under neutron irradiation, high temperature, high pressure, and corrosive environments. In order to increase the safety of nuclear power plants, it is inevitable to improve the credibility and capability of evaluation technology based on the quantitative fracture mechanics for aging assessment of reactor components. Irradiation embrittlement is the primary aging mechanism of reactor pressure vessel and various techniques have been developed to predict the aging characteristics by using only small volume of irradiated materials. Material database of the domestic structural steels for KSNP's under reactor environments must be very important to play a role in developing an advanced material, in improving the safety of nuclear components, and also in expanding the nuclear industry abroad. This research project has been focused on developing an advanced technology of testing and analysis in the fracture mechanical point of view as well as acquiring test data and improving the performance of nuclear structural steels

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

International Nuclear Information System (INIS)

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

2017-01-01

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

Effects of fracture distribution and length scale on the equivalent continuum elastic compliance of fractured rock masses

Directory of Open Access Journals (Sweden)

Marte Gutierrez

2015-12-01

Full Text Available Fracture systems have strong influence on the overall mechanical behavior of fractured rock masses due to their relatively lower stiffness and shear strength than those of the rock matrix. Understanding the effects of fracture geometrical distribution, such as length, spacing, persistence and orientation, is important for quantifying the mechanical behavior of fractured rock masses. The relation between fracture geometry and the mechanical characteristics of the fractured rock mass is complicated due to the fact that the fracture geometry and mechanical behaviors of fractured rock mass are strongly dependent on the length scale. In this paper, a comprehensive study was conducted to determine the effects of fracture distribution on the equivalent continuum elastic compliance of fractured rock masses over a wide range of fracture lengths. To account for the stochastic nature of fracture distributions, three different simulation techniques involving Oda's elastic compliance tensor, Monte Carlo simulation (MCS, and suitable probability density functions (PDFs were employed to represent the elastic compliance of fractured rock masses. To yield geologically realistic results, parameters for defining fracture distributions were obtained from different geological fields. The influence of the key fracture parameters and their relations to the overall elastic behavior of the fractured rock mass were studied and discussed. A detailed study was also carried out to investigate the validity of the use of a representative element volume (REV in the equivalent continuum representation of fractured rock masses. A criterion was also proposed to determine the appropriate REV given the fracture distribution of the rock mass.

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

International Nuclear Information System (INIS)

Kitagawa, Hideo; Hisada, Toshiaki

1979-01-01

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

Evaluation and prediction of neutron embrittlement in reactor pressure vessel materials. Final report

International Nuclear Information System (INIS)

Hawthorne, J.R.; Menke, B.H.; Loss, F.J.; Watson, H.E.; Hiser, A.L.; Gray, R.A.

1982-12-01

This study evaluates the effects of fast neutron irradiation on the mechanical properties of eight nuclear reactor vessel materials. The materials include submerged arc weldments, three plates, and one forging. The materials are in the unirradiated and irradiated conditions with regard to tensile, Charpy impact, and static and dynamic fracture toughness properties. Correlations between impact and fracture toughness parameters are developed from the experimental results. The observed shifts in transition temperature and the drop in upper-shelf energy are compared with predictions developed from the Regulatory Guide 1.99.1 trend curves

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

Energy Technology Data Exchange (ETDEWEB)

Chowdhury, Tamshuk, E-mail: tamshuk@gmail.com [Deep Sea Technologies, National Institute of Ocean Technology, Chennai, 600100 (India); Sivaprasad, S.; Bar, H.N.; Tarafder, S. [Fatigue & Fracture Group, Materials Science and Technology Division, CSIR-National Metallurgical Laboratory, Jamshedpur, 831007 (India); Bandyopadhyay, N.R. [School of Materials Science and Engineering, Indian Institute of Engineering, Science and Technology, Shibpur, Howrah, 711103 (India)

2016-04-15

Cyclic J-R behaviour of a reactor pressure vessel steel using different methods available in literature has been examined to identify the best suitable method for cyclic fracture problems. Crack opening point was determined by moving average method. The η factor was experimentally determined for cyclic loading conditions and found to be similar to that of ASTM value. Analyses showed that adopting a procedure analogous to the ASTM standard for monotonic fracture is reasonable for cyclic fracture problems, and makes the comparison to monotonic fracture results straightforward. - Highlights: • Different methods of cyclic J-R evaluation compared. • A moving average method for closure point proposed. • η factor for cyclic J experimentally validated. • Method 1 is easier, provides a lower bound and direct comparison to monotonic fracture.

Technology development on analysis program for measuring fracture toughness of irradiated specimens

International Nuclear Information System (INIS)

Shibata, Akira; Takada, Fumiki

2007-03-01

The fracture toughness which represents resistance for brittle or ductile fracture is one of the most important material property concerning linear and non-linear fracture mechanics analyses. In order to respond to needs of collecting data relating to fracture toughness of pressure vessel and austenitic stainless steels, fracture toughness test for irradiated materials has been performed in JMTR hot laboratory. On the other hand, there has been no computer program for analysis of fracture toughness using the test data obtained from the test apparatus installed in the hot cell. Therefore, only load-displacement data have been provided to users to calculate fracture toughness of irradiated materials. Recently, request of analysis of fracture toughness have been increased. Thus a computer program, which calculates the amount of the crack extension, the compliance and the fracture toughness from the data acquired from the test apparatus installed in the hot cell, has been developed. In the program unloading elastic compliance method is applied based on ASTM E1820-01. Through the above development, the request for the fracture toughness analysis can be satisfied and the fracture toughness of irradiated test specimens can be provided to users. (author)

Models for embrittlement recovery due to annealing of reactor pressure vessel steels

International Nuclear Information System (INIS)

Eason, E.D.; Wright, J.E.; Nelson, E.E.; Odette, G.R.; Mader, E.V.

1995-05-01

The reactor pressure vessel (RPV) surrounding the core of a commercial nuclear power plant is subject to embrittlement due to exposure to high energy neutrons. The effects of irradiation embrittlement can be reduced by thermal annealing at temperatures higher than the normal operating conditions. However, a means of quantitatively assessing the effectiveness of annealing for embrittlement recovery is needed. The objective of this work was to analyze the pertinent data on this issue and develop quantitative models for estimating the recovery in 30 ft-lb (41 J) Charpy transition temperature and Charpy upper shelf energy due to annealing. Data were gathered from the Test Reactor Embrittlement Data Base and from various annealing reports. An analysis data base was developed, reviewed for completeness and accuracy, and documented as part of this work. Independent variables considered in the analysis included material chemistries, annealing time and temperature, irradiation time and temperature, fluence, and flux. To identify important variables and functional forms for predicting embrittlement recovery, advanced statistical techniques, including pattern recognition and transformation analysis, were applied together with current understanding of the mechanisms governing embrittlement and recovery. Models were calibrated using multivariable surface-fitting techniques. Several iterations of model calibration, evaluation with respect to mechanistic and statistical considerations, and comparison with the trends in hardness data produced correlation models for estimating Charpy upper shelf energy and transition temperature after irradiation and annealing. This work provides a clear demonstration that (1) microhardness recovery is generally a very good surrogate for shift recovery, and (2) there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes

Radiation embrittlement in pressure vessels of power reactors

International Nuclear Information System (INIS)

Kempf, Rodolfo; Fortis, Ana M.

2007-01-01

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

Pisiform fractures

International Nuclear Information System (INIS)

Fleege, M.A.; Jebson, P.J.; Renfrew, D.L.; El-Khoury, G.Y.; Steyers, C.M. Jr.

1991-01-01

Fractures of the pisiform are often missed due to improper radiographic evaluation and a tendency to focus on other, more obvious injuries. Delayed diagnosis may result in disabling sequelae. A high index of clinical suspicion and appropriate radiographic examination will establish the correct diagnosis. Ten patients with pisiform fracture are presented. The anatomy, mechanism of injury, clinical presentation, radiographic features, and evaluation of this injury are discussed. (orig.)

Design of the ITER vacuum vessel

International Nuclear Information System (INIS)

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

1995-01-01

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

Fracture Toughness Evaluation of Kori-1 RPV Beltline Weld for a Long-Term Operation

International Nuclear Information System (INIS)

Lee, Bong-Sang; Kim, Min-Chul; Ahn, Sang-Bok; Kim, Byung-Chul; Hong, Jun-Hwa

2007-01-01

Irradiation embrittlement of RPV (reactor pressure vessel) material is the most important aging issue for a long-term operation of nuclear power plants. KORI unit 1, which is the first PWR in Korea, is approaching its initial licensing life of 30 years. In order to operate the reactor for another 10 years and more, it should be demonstrated that the irradiation embrittlement of the reactor will be adequately managed by ensuring that the fracture toughness properties have a certain level of the safety margin. The current regulation requires Charpy V-notch impact data through conventional surveillance tests. It is based on the assumption that Charpy impact test results are well correlated with the fracture toughness properties of many engineering steels. However, Charpy V-notch impact data may not be adequate to estimate the fracture toughness of certain materials, such as Linde 80 welds. During the last decade, a tremendous number of fracture toughness data on many RPV steels have been produced in accordance with the new standard test method, the so-called master curve method. ASTM E1921 represents a revolutionary advance in characterizing fracture toughness of RPV steels, since it permits establishing the ductile to brittle transition portion of the fracture toughness curve with direct measurements on a relatively small number of relatively small specimens, such as pre-cracked Charpy specimens. Actual fracture toughness data from many different RPV steels revealed that the Charpy test estimations are generally conservative with the exception of a few cases. Recent regulation codes in USA permit the master curve fracture toughness methodology in evaluating an irradiation embrittlement of commercial nuclear reactor vessels

Coupled hydrological-mechanical effects due to excavation of underground openings in unsaturated fractured rocks

International Nuclear Information System (INIS)

Montazer, P.

1985-01-01

One of the effects of excavating an underground opening in fractured rocks is a modification of the state of the stress in the rock mass in the vicinity of the opening. This effect causes changes in the geometry of the cross sections of the fracture planes, which in turn results in modification of the hydrologic properties of the fractures of the rock mass. The significance of the orientation of the fractures and their stiffness on the extent of the modification of the hydrologic properties as a result of excavation of underground openings is demonstrated. A conceptual model is presented to illustrate the complexity of the coupled hydrological-mechanical phenomena in the unsaturated zone. This conceptual model is used to develop an investigative program to assess the extent of the effect at a proposed repository site for storing high-level nuclear wastes

Midland reactor pressure vessel flaw distribution

International Nuclear Information System (INIS)

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

1993-12-01

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

Earthquake-proof supporting structure in reactor vessel

International Nuclear Information System (INIS)

Sakurai, Akio; Sekine, Katsuhisa; Madokoro, Manabu; Katoono, Shin-ichi; Konno, Mutsuo; Suzuki, Takuro.

1990-01-01

Conventional earthquake-proof structure comprises a vessel vibration stopper integrated to a reactor vessel, powder for restricting the horizontal displacements, a safety vessel surrounds the outer periphery of the reactor vessel and a safety vessel vibration stopper integrated therewith, which are fixed to buildings. However, there was a problem that a great amount of stresses are generated in the base of the reactor vessel vibration stopper due to reaction of the powders which restrict thermal expansion. In order to remarkably reduce the reaction of the powers, powders are charged into a spaces formed between each of the reactor vessel vibration stopper, the safety vessel vibration stopper and the flexible member disposed between them. According to this constitution, the reactor vessel vibration stopper does not undergo a great reaction of the powers upon thermal expansion of the reactor vessel to moderate the generated stresses, maintain the strength and provide earthquake-proof supporting function. (N.H.)

Confinement Vessel Assay System: Calibration and Certification Report

Energy Technology Data Exchange (ETDEWEB)

Frame, Katherine C. [Los Alamos National Laboratory; Bourne, Mark M. [Los Alamos National Laboratory; Crooks, William J. [Los Alamos National Laboratory; Evans, Louise [Los Alamos National Laboratory; Gomez, Cipriano [Retired CMR-OPS: OPERATIONS; Mayo, Douglas R. [Los Alamos National Laboratory; Miko, David K. [Los Alamos National Laboratory; Salazar, William R. [Los Alamos National Laboratory; Stange, Sy [Los Alamos National Laboratory; Vigil, Georgiana M. [Los Alamos National Laboratory

2012-07-17

Los Alamos National Laboratory has a number of spherical confinement vessels (CVs) remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1 to 2 inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the vessels. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of SNM in CVs before and after cleanout. Prior to cleanout, the system will be used to perform a verification measurement of each vessel. After cleanout, the system will be used to perform safeguards-quality assays of {le} 100-g {sup 239}Pu equivalent in a vessel for safeguards termination. The system was calibrated in three different mass regions (low, medium, and high) to cover the entire plutonium mass range that will be assayed. The low mass calibration and medium mass calibration were verified for material positioned in the center of an empty vessel. The systematic uncertainty due to position bias was estimated using an MCNPX model to simulate the response of the system to material localized at various points along the inner surface of the vessel. The background component due to cosmic ray spallation was determined by performing measurements of an empty vessel and comparing to measurements in the same location with no vessel present. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements of CVs before and after cleanout.

Confinement Vessel Assay System: Calibration and Certification Report

International Nuclear Information System (INIS)

Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Gomez, Cipriano; Mayo, Douglas R.; Miko, David K.; Salazar, William R.; Stange, Sy; Vigil, Georgiana M.

2012-01-01

Los Alamos National Laboratory has a number of spherical confinement vessels (CVs) remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1 to 2 inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the vessels. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of SNM in CVs before and after cleanout. Prior to cleanout, the system will be used to perform a verification measurement of each vessel. After cleanout, the system will be used to perform safeguards-quality assays of (le) 100-g 239 Pu equivalent in a vessel for safeguards termination. The system was calibrated in three different mass regions (low, medium, and high) to cover the entire plutonium mass range that will be assayed. The low mass calibration and medium mass calibration were verified for material positioned in the center of an empty vessel. The systematic uncertainty due to position bias was estimated using an MCNPX model to simulate the response of the system to material localized at various points along the inner surface of the vessel. The background component due to cosmic ray spallation was determined by performing measurements of an empty vessel and comparing to measurements in the same location with no vessel present. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements of CVs before and after cleanout.

Fracture of the acetabulum with femoral artery injury presenting late: A case report

Directory of Open Access Journals (Sweden)

Sivaprasad Kalyanasundaram

2016-02-01

Full Text Available This study reports a rare case of both column acetabulum fracture with femoral artery injury that presented late and was managed with arterial reconstruction and fracture fixation.A thirty-one year old man sustained both column acetabular fracture on the left in a motor vehicle accident. On admission there was no obvious neuro-vascular deficit. During surgery for the fracture after 7 days of the injury the femoral artery was found to be severely crushed with no blood flow. The anterior column of the acetabulum was stabilised followed by resection and reconstruction of the femoral artery. The post-operative period was uneventful and he was discharged normally. At 6 months from injury the fractures had united well with excellent limb circulation and good lower limb function.Femoral artery injury with acetabular fracture is rare and late presentations are unreported hitherto. The results of fracture stabilisation and vessel reconstruction seem to be excellent. Literature of similar injuries is reviewed. Keywords: Acetabular fractures, Both column fractures, Anterior column fractures, Vascular injury, Femoral artery injury

Fracture toughness evaluation of a low upper-shelf weld metal from the Midland Reactor using the master curve

International Nuclear Information System (INIS)

McCabe, D.E.; Sokolov, M.A.; Nanstad, R.K.

1997-01-01

The primary objective of the Heavy-Section Steel Irradiation (HSSI) Program Tenth Irradiation Series was to develop a fracture mechanics evaluation of weld metal WF-70, which was taken from the beltline and nozzle course girth weld joints of the Midland Reactor vessel. This material became available when Consumers Power Company of Midland, Michigan, decided to abort plans to operate their nuclear power plant. WF-70 is classified as a low upper-shelf steel primarily due to the Linde 80 flux that was used in the submerged-arc welding process. The master curve concept is introduced to model the transition range fracture toughness when the toughness is quantified in terms of K Jc values. K Jc is an elastic-plastic stress intensity factor calculated by conversion from J c ; i.e., J-integral at onset of cleavage instability

Avulsion fractures of the scapula

International Nuclear Information System (INIS)

Heyse-Moore, G.H.; Stoker, D.J.

1982-01-01

Fractures of the scapula due to direct violence are relatively common. Wilber and Evans [18] reported 40 scapular fractures and reviewed the literature. All those injured has received direct trauma to the shoulder and they were able to divide their cases into two groups, based on anatomical location and functional results. Scapular fractures due to avulsion of the muscular attachments are uncommon and, as reports of these injuries in the literature are usually confined to single cases, no classification has been established which takes account of the anatomical sites at which these fractures occur and the mechanism of injury involved. In this paper the more common sites of avulsion injury of the scapula are described and illustrated by case reports. In several of these the skeletal injury resulted from muscle contraction against a resisted force on the upper limb during the course of an accident. This mechanism has been implicated in fractures of the coracoid and acromion, but is shown in this paper to contribute to other avulsion fractures. (orig.)

Guidelines for Application of the Master Curve Approach to Reactor Pressure Vessel Integrity in Nuclear Power Plants

International Nuclear Information System (INIS)

Lyssakov, V.N.; Kang, K.S.

2005-01-01

These guidelines have been developed under an International Atomic Energy Agency (IAEA) Co-ordinated Research Project (CRP) titled ''Surveillance Programme Results Application to Reactor Pressure Vessel Integrity Assessment.'' The IAEA has sponsored a series of five CRPs that have led to a focus on measuring the best irradiation fracture parameters using relatively small test specimens for assuring structural integrity of reactor pressure vessel (RPV) materials in Nuclear Power Plants (NPPs)

Probabilistic atlas based labeling of the cerebral vessel tree

Science.gov (United States)

Van de Giessen, Martijn; Janssen, Jasper P.; Brouwer, Patrick A.; Reiber, Johan H. C.; Lelieveldt, Boudewijn P. F.; Dijkstra, Jouke

2015-03-01

Preoperative imaging of the cerebral vessel tree is essential for planning therapy on intracranial stenoses and aneurysms. Usually, a magnetic resonance angiography (MRA) or computed tomography angiography (CTA) is acquired from which the cerebral vessel tree is segmented. Accurate analysis is helped by the labeling of the cerebral vessels, but labeling is non-trivial due to anatomical topological variability and missing branches due to acquisition issues. In recent literature, labeling the cerebral vasculature around the Circle of Willis has mainly been approached as a graph-based problem. The most successful method, however, requires the definition of all possible permutations of missing vessels, which limits application to subsets of the tree and ignores spatial information about the vessel locations. This research aims to perform labeling using probabilistic atlases that model spatial vessel and label likelihoods. A cerebral vessel tree is aligned to a probabilistic atlas and subsequently each vessel is labeled by computing the maximum label likelihood per segment from label-specific atlases. The proposed method was validated on 25 segmented cerebral vessel trees. Labeling accuracies were close to 100% for large vessels, but dropped to 50-60% for small vessels that were only present in less than 50% of the set. With this work we showed that using solely spatial information of the vessel labels, vessel segments from stable vessels (>50% presence) were reliably classified. This spatial information will form the basis for a future labeling strategy with a very loose topological model.

Employing Eigenvalue Ratios to Generate Prior Fracture-like Features for Stochastic Hydrogeophysical Characterization of a Fractured Aquifer System

Science.gov (United States)

Brewster, J.; Oware, E. K.

2017-12-01

Groundwater hosted in fractured rocks constitutes almost 65% of the principal aquifers in the US. The exploitation and contaminant management of fractured aquifers require fracture flow and transport modeling, which in turn requires a detailed understanding of the structure of the aquifer. The widely used equivalent porous medium approach to modeling fractured aquifer systems is inadequate to accurately predict fracture transport processes due to the averaging of the sharp lithological contrast between the matrix and the fractures. The potential of geophysical imaging (GI) to estimate spatially continuous subsurface profiles in a minimally invasive fashion is well proven. Conventional deterministic GI strategies, however, produce geologically unrealistic, smoothed-out results due to commonly enforced smoothing constraints. Stochastic GI of fractured aquifers is becoming increasing appealing due to its ability to recover realistic fracture features while providing multiple likely realizations that enable uncertainty assessment. Generating prior spatial features consistent with the expected target structures is crucial in stochastic imaging. We propose to utilize eigenvalue ratios to resolve the elongated fracture features expected in a fractured aquifer system. Eigenvalues capture the major and minor directions of variability in a region, which can be employed to evaluate shape descriptors, such as eccentricity (elongation) and orientation of features in the region. Eccentricity ranges from zero to one, representing a circularly sharped to a line feature, respectively. Here, we apply eigenvalue ratios to define a joint objective parameter consisting of eccentricity (shape) and direction terms to guide the generation of prior fracture-like features in some predefined principal directions for stochastic GI. Preliminary unconditional, synthetic experiments reveal the potential of the algorithm to simulate prior fracture-like features. We illustrate the strategy with a

Study on closed pressure vessel test. Effect of heat rate, sample weight and vessel size on pressure rise due to thermal decomposition; Mippeigata atsuryoku yoki shiken ni kansuru kenkyu. Atsuryoku hassei kyodo ni oyobosu kanetsusokudo, shiryoryo oyobi youki saizu no eikyo

Energy Technology Data Exchange (ETDEWEB)

Aoki, Kenji.; Akutsu, Yoshiaki.; Arai, Mitsuru.; Tamura, Masamitsu. [The University of Tokyo, Tokyo (Japan). School of Engineering

1999-02-28

We have attempted to devise a new closed pressure vessel test apparatus in order to evaluate the violence of thermal decomposition of self-reactive materials and have examined some influencing factors, such as heat rate, sample weight, filling factor (sample weight/vessel size) and vessel size on Pmax (maximum pressure rise) and dP/dt (rate of pressure rise) due to their thermal decomposition. As a result, the following decreasing orders of Pmax and dP/dt were shown. Pmax: ADCA>BPZ>AIBN>TCP dP/dt: AIBN>BPZ>ADCA>TCP Moreover, Pmax was not almost influenced by heat rate, while dP/dt increased with an increase in heat rate in the case of BPZ. Pmax and dP/dt increased with an increase in sample weight and the degree of increase depended on the kinds of materials. In addition, it was shown that Pmax and dP/dt increased with an increase in vessel size at a constant filling factor. (author)

Development of a plastic fracture methodology for nuclear systems

International Nuclear Information System (INIS)

Marston, T.U.; Jones, R.L.; Kanninen, M.F.; Mowbray, D.F.

1981-01-01

This paper describes research conducted to develop a fundamental basis for flaw tolerance assessment procedures suitable for components exhibiting ductile behavior. The research was composed of an integrated combination of stable crack growth experiments and elastic-plastic analyses. A number of candidate fracture criteria were assembled and investigated to determine the proper basis for plastic fracture mechanics assessments. The results demonstrate that many different fracture criteria can be used as the basis of a resistance curve approach to predicting stable crack growth and fracture instability. While all have some disadvantages, none is completely unacceptable. On balance, the best criteria were found to be the J-integral for initiation and limited amounts of stable crack growth and the local crack-tip opening angle for extended amounts of stable growth. A combination of the two, which may preserve the advantages of each while reducing their disadvantages, also was suggested by these results. The influence of biaxial and mixed flat/shear fracture behavior was investigated and found to not alter the basic results. Further work in the development of simplified ductile fracture analyses for routine engineering assessments of nuclear pressure vessels and piping evolving from this research is also described

Resolution of the Task A-11 reactor-vessel materials-toughness safety issue. Appendices C-K

International Nuclear Information System (INIS)

1982-10-01

The central problem in the unresolved safety issue A-11, Reactor Vessel Materials Toughness, was to provide guidance in performing analyses required by 10 CFR Part 50, Appendix G, Section V.C. for reactor pressure vessels (RPVs) which fail to meet the toughness requirement during service life as a result of neutron radiation embrittlement. Although the methods of linear-elastic fracture mechanics (LEFM) were adequate for low-temperature RPV problems, they were inapplicable under operating conditions because vessel steels, even those which exhibit less than 50 ft-lb of C/sub v/ energy, were relatively tough at temperatures where the impact energy reached its upper shelf values. A technical team of recognized experts was organized to assist the NRC staff in addressing the problem. Using the foundation of the tearing modulus concept, which had been developed under earlier NRC sponsorship, relationships were obtained which provided approximate solutions to the problem of RPV fracture with assumed beltline region flaws. The first paper of this report is a summary of the problem, the solutions, and the results of verification analyses. The details are provided in a series of appendices in Volumes I and II

GUNSHOT FRACTURES OF TIBIA AND FEMUR - EXCELLENT ...

African Journals Online (AJOL)

2011-10-10

Oct 10, 2011 ... fractures due to gunshot injury grafted with reamed bone marrow and immobilised with Surgical ... open fractures, which pose a challenging problem .... Table 2. Gustillo-Anderson Classification of fractures and infection.

Chest and spine radiography abnormality in blunt chest trauma correlated with major vessel injury in an unselected patient population

International Nuclear Information System (INIS)

Fernandez, G.; Kadir, S.; Encarnacion, C.

1989-01-01

To assess the true incidence of major vessel injury, the authors retrospectively reviewed all arch aortograms obtained for blunt chest trauma (BCT) during a 24-month period beginning December 1986. Aortograms were correlated with preangiographic chest radiographic and operative findings. The goals of this review were to examine the usefulness of commonly employed screening criteria for aortography and determine whether thoracic spine fractures imply a decreased likelihood of aortic injury. One hundred twenty aortograms were obtained during this period. The incidence of aortic laceration was 6.7%, and 7.5% had brachiocerebral vascular injury. Only 51% of chest radiographs were suggestive of vascular injury. Two patients with subtle radiographic findings had aortic laceration. One patient with a burst fracture of T-4 had aortic laceration. The results of this review indicate the incidence of great vessel injury is as high as that of injury to the aorta itself and that the presence of spine fractures does not exclude vascular injury

Hydraulic fracture propagation modeling and data-based fracture identification

Science.gov (United States)

Zhou, Jing

Successful shale gas and tight oil production is enabled by the engineering innovation of horizontal drilling and hydraulic fracturing. Hydraulically induced fractures will most likely deviate from the bi-wing planar pattern and generate complex fracture networks due to mechanical interactions and reservoir heterogeneity, both of which render the conventional fracture simulators insufficient to characterize the fractured reservoir. Moreover, in reservoirs with ultra-low permeability, the natural fractures are widely distributed, which will result in hydraulic fractures branching and merging at the interface and consequently lead to the creation of more complex fracture networks. Thus, developing a reliable hydraulic fracturing simulator, including both mechanical interaction and fluid flow, is critical in maximizing hydrocarbon recovery and optimizing fracture/well design and completion strategy in multistage horizontal wells. A novel fully coupled reservoir flow and geomechanics model based on the dual-lattice system is developed to simulate multiple nonplanar fractures' propagation in both homogeneous and heterogeneous reservoirs with or without pre-existing natural fractures. Initiation, growth, and coalescence of the microcracks will lead to the generation of macroscopic fractures, which is explicitly mimicked by failure and removal of bonds between particles from the discrete element network. This physics-based modeling approach leads to realistic fracture patterns without using the empirical rock failure and fracture propagation criteria required in conventional continuum methods. Based on this model, a sensitivity study is performed to investigate the effects of perforation spacing, in-situ stress anisotropy, rock properties (Young's modulus, Poisson's ratio, and compressive strength), fluid properties, and natural fracture properties on hydraulic fracture propagation. In addition, since reservoirs are buried thousands of feet below the surface, the

Primer: Fracture mechanics in the nuclear power industry

International Nuclear Information System (INIS)

Wessel, E.T.; Server, W.L.; Kennedy, E.L.

1990-01-01

This Primer is intended to familiarize utility engineers with the fracture mechanics technology and to provide the basis for a working knowledge of the subject. It is directed towards all the engineering disciplines that are involved either directly or indirectly with the structural reliability of electrical power generation equipment and systems. These engineering disciplines include such areas as: design and stress analysis, metallurgy and materials, nondestructive inspection and quality control, structural analysis and reliability engineering, chemical engineering and water chemistry control, and architectural engineering. This Primer does not provide a comprehensive, in-depth treatment of all the detailed aspects involved in fracture mechanics. It does, however, provide sufficient information and a common vocabulary that should enable engineers to: read and converse intelligently about the subject, understand and utilize ASME Codes and Regulatory Guides involving fracture mechanics, absorb technical information presented and discussed at various technical meetings, and begin to apply this technology towards actual engineering problems encountered in the course of their work. Example problems are provided to further enhance an understanding of fracture mechanics. Also, Appendix A describes fracture mechanics computer codes available through EPRI to analyze rotors, reactor pressure vessels and piping

Progress in elastic-plastic fracture mechanics and its applications

International Nuclear Information System (INIS)

Paris, P.C.; Zahalak, G.I.

1980-01-01

This paper surveys recent developments in the application of J-Integral methods to problems of elastic-plastic fracture. The analytical and experimental development of the J-Integral concept over the last ten years is reviewed briefly. Tearing instability theory is presented in general terms, and specific applications of the theory are discussed. Principles of fracture-proof design are shown to follow naturally from the tearing instability theory. These principles are illustrated first for simple structures, and then generalized to more complex configurations and loading conditions. Examples include multiple member tension structures, beams, frames, nuclear reactor pressure vessel nozzles and piping, and beams on elastic foundations. It is concluded that J-integral based methods offer the best immediate opportunity for the development of sound analytical techniques for treating important practical problems of elastic-plastic fracture

The prospect of modern thermomechanics in structural integrity calculations of large-scale pressure vessels

Science.gov (United States)

Fekete, Tamás

2018-05-01

Structural integrity calculations play a crucial role in designing large-scale pressure vessels. Used in the electric power generation industry, these kinds of vessels undergo extensive safety analyses and certification procedures before deemed feasible for future long-term operation. The calculations are nowadays directed and supported by international standards and guides based on state-of-the-art results of applied research and technical development. However, their ability to predict a vessel's behavior under accidental circumstances after long-term operation is largely limited by the strong dependence of the analysis methodology on empirical models that are correlated to the behavior of structural materials and their changes during material aging. Recently a new scientific engineering paradigm, structural integrity has been developing that is essentially a synergistic collaboration between a number of scientific and engineering disciplines, modeling, experiments and numerics. Although the application of the structural integrity paradigm highly contributed to improving the accuracy of safety evaluations of large-scale pressure vessels, the predictive power of the analysis methodology has not yet improved significantly. This is due to the fact that already existing structural integrity calculation methodologies are based on the widespread and commonly accepted 'traditional' engineering thermal stress approach, which is essentially based on the weakly coupled model of thermomechanics and fracture mechanics. Recently, a research has been initiated in MTA EK with the aim to review and evaluate current methodologies and models applied in structural integrity calculations, including their scope of validity. The research intends to come to a better understanding of the physical problems that are inherently present in the pool of structural integrity problems of reactor pressure vessels, and to ultimately find a theoretical framework that could serve as a well

Behaviour of the nozzle corner region during the first phase of the fatigue test on scaled models of pressure vessels (JRC Vessel A)

International Nuclear Information System (INIS)

Jovanovic, A.; Lucia, A.C.; Brunnhuber, R.; Elbaz, J.M.; Schwarz, U.

1987-01-01

The work presented here deals mainly with the stress and fracture mechanics aspects of the first phase of the structural reliability experimental program based on the scaled experimental vessel at Ispra. The overall research and experiments make also part of the structural reliability assessment extrapolation pattern for the full-scale structures. The work presented here deals with the problem of the nozzle corner cracks induced by the fatigue under the pressure cycling

An experimental study on feasibility of ex-vessel cooling through the external guide vessel

International Nuclear Information System (INIS)

Kang, Kyoung-Ho; Kim, Jong-Hwan; Park, Rae-Jun; Kim, Sang-Baik

2000-01-01

This paper presents the results of a series of experiments for assessing the efficacy of ex-vessel cooling through the external guide vessel during a severe accident. Four tests were performed in the LAVA test facility at KAERI, varying the boundary conditions at the outer surface of the vessel. The first test was a dry condition test conducted without cooling the outside of the vessel. On the other hand, in the second test, the cooling of the vessel surface was produced by gravity-driven forced injection of water along the annular gap of 25 mm between the vessel and the external guide vessel. Water flow rate was about 0.85 kg/s and total mass of available water was 300 kg. For the evaluation of the water flow rate effect, the third test was performed with a pool type cooling in the annulus without any circulation of water. These two external cooling tests were performed under elevated pressure of about 1.6 MPa. Finally, the fourth test was conducted under atmospheric pressure to evaluate the effect of system pressure on boiling heat transfer characteristics. In the dry test and the pool type ex-vessel cooling test performed under atmospheric pressure, the vessel was failed by a melt penetration at about 40 degree upper position from the vessel bottom, which is coincident with the boundary of the Al 2 O 3 /Fe melt separated layers. On the other hand, in both of the ex-vessel cooling tests conducted under elevated pressure of about 1.6 MPa, the vessel didn't fail. Compared with the pool boiling test, the vessel experienced effective cooling due to the inlet flow in the forced flow test. Synthesized the results of the tests, it was shown that the heat removal with ex-vessel cooling through the guide vessel is feasible, but the additional evaluations should be performed to guarantee enough thermal margin. (author)

Fatigue and insufficiency fractures

International Nuclear Information System (INIS)

Lodwick, G.S.; Rosenthal, D.I.; Kattapuram, S.V.; Hudson, T.M.

1987-01-01

The incidence of stress fracture is increasing. In our younger society this is due largely to a preocupation with physical conditioning, but in our elderly population it is due to improved recognition and better methods of detection and diagnosis. Stress fracture of the elderly is an insufficiency fracture which occurs in the spine, the pelvis, the sacrum and other bones afflicted with disorders which cause osteopenia. Stress fracture is frequently misdiagnosed as a malignant lesion of bone resulting in biopsy. Scintiscanning provides the greatest frequency of detection, while computed tomography often provides the definitive diagnosis. With increased interest and experience a better insight into the disease has been achieved, and what was once thought of as a simple manifestation of mechanical stress is now known to be an orderly, complex pattern of physiological changes in bone which conform to a model by Frost. The diffuse nature of these changes can be recognized by scintigraphy, radiography and magnetic resonance imaging. 27 refs.; 8 figs

Hydraulic Fracturing and Microseismicity: Global Perspective in Oil Exploration

Directory of Open Access Journals (Sweden)

J.R. Kayal

2017-09-01

Full Text Available Induced microseismicity is a common phenomenon in oil and gas reservoirs due to changes in internal stress accompanied by hydraulic fracturing and oil-gas extraction. These microseismicity can be monitored to understand the direction and type of hydraulic fracturing and pre-existing faults by precise hypocenter location and focal mechanism studies. Normal as well as strike-slip faulting earthquakes occur due to opening up of new cracks/fractures, and thrust/reverse faulting earthquakes due to compaction or closing of existing fractures. Further, frequency-magnitude relation (b-value and fractal dimension (D-value of the spatial and temporal clusterization of induced microseismicity may be much useful to characterize the fractures / existing faults and the stress regimes. Seismic tomography, on the other hand, can image the heterogeneous velocity structures / perturbations in the reservoir due to fractures and oil-gas-water contents. A few global case studies are illustrated to understand these processes and to draw attention towards importance of these studies in oil industries.

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

Prestressed concrete reactor vessels: review of design and failure criteria

International Nuclear Information System (INIS)

Endebrock, E.G.

1975-03-01

The design and failure criteria of prestressed concrete reactor vessels (PCRVs) are reviewed along with the analysis methods. The mechanical properties of concrete under multiaxial stresses are not adequately quantified or described to permit an accurate analysis of a PCRV. Structural analysis of PCRVs almost universally utilizes a finite element which encounters difficulties in numerical solution of the governing equations and in treatment of fractured elements. (U.S.)

Fractures in elderly due to falls registered in a traumatology tertiary reference hospital in the year 2004 - doi:10.5020/18061230.2007.p226

Directory of Open Access Journals (Sweden)

Marcelo de Carvalho Filgueiras

2012-01-01

Full Text Available The falls followed by fractures in elderly patients consist in an important cause of morbidity and mortality, being a public health problem, due to the high incidence of cases, their seriousness and socioeconomic cost. They may range from a small lesion to the loss of mobility and even death. This study aimed at analyzing the topographical distribution, from the anatomical aspect, of post-fall fractures in elders registered at the Dr. José Frota Institute, in Fortaleza – Ceará, relating it to the gender, the age group, internment period and possible complications. This was an epidemiological, descriptive and retrospective study based on data collected from documented evidence. The study population was composed of patients over 60, victims of post-fall fractures, attended at the institution, in the year 2004. As results, it was verified that 50.25% of the fractures occurred in the neck of the femur, followed by femoral shaft fractures (15.57%, cranial fractures (12.57% and radium (9.58%. Approximately 36.5% stayed interned from 1 to 10 days, the majority of the cases being female (55.69%, with predominance of the age group above 81 years old. In males, the most affected age group was that of 76 to 80 years old, comprising 28.38% of the men. From this research, it was possible to delineate a profile of the risk factors of fractures in elderly, allowing studies directed to susceptive groups.

Reactor pressure vessel thermal annealing

International Nuclear Information System (INIS)

Lee, A.D.

1997-01-01

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

Application of the RKR model for evaluating the fracture toughness of pressure vessel steel in the transition temperature region

International Nuclear Information System (INIS)

Yang, Won Jon; Huh, Moo Young; Lee, Bong Sang; Hong, Jun Hwa

2002-01-01

Fracture toughness of a SA 533 B-1 steel was characterized in ductile-brittle transition temperature region by means of a RKR-type model. The original RKR model has been used to predict the plane strain fracture toughness (K IC ) behaviors in lower shelf region by assuming two material parameters, ie, the critical fracture stress and the characteristic distance. In this study, the fracture surface of every specimen was thoroughly investigated using scanning electron microscope to locate the actual cleavage initiation and to measure the cleavage initiation distance (CID) from the initial crack. The local fracture stress (σ f * ) of material was determined from the elastic-plastic stress field at the measured cleavage initiation location in the notched and precracked specimen. The local fracture stress of the precracked specimens was much higher than that of the notched specimen. The measured CIDs were strongly dependent on the test temperature and also on the fracture toughness. Based on the observations, it is found that, in the RKR-type cleavage fracture models, the characteristic distance should not be treated as a constant material parameter in the ductile-brittle transition region where the cleavage initiation controls the overall fracture process

Pressurized Thermal Shock Analysis for OPR1000 Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Bhowmik, P. K.; Shamim, J. A.; Gairola, A.; Suh, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of)

2014-10-15

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

Studies of Transport Properties of Fractures: Final Report

Energy Technology Data Exchange (ETDEWEB)

Stephen R. Brown

2006-06-30

We proposed to study several key factors controlling the character and evolution of fracture system permeability and transport processes. We suggest that due to surface roughness and the consequent channeling in single fractures and in fracture intersections, the tendency of a fracture system to plug up, remain permeable, or for permeability to increase due to chemical dissolution/precipitation conditions will depend strongly on the instantaneous flow channel geometry. This geometry will change as chemical interaction occurs, thus changing the permeability through time. To test this hypothesis and advance further understanding toward a predictive capability, we endeavored to physically model and analyze several configurations of flow and transport of inert and chemically active fluids through channels in single fractures and through fracture intersections. This was an integrated program utilizing quantitative observations of fractures and veins in drill core, quantitative and visual observations of flow and chemical dissolution and precipitation within replicas of real rough-walled fractures and fracture intersections, and numerical modeling via lattice Boltzmann methods.

The role of ductile ligaments and warm prestress on the re-initiation of fracture from a crack arrested during thermal shock

International Nuclear Information System (INIS)

Smith, E.

1982-01-01

The protection offered by warm prestress can be important for preserving a nuclear pressure vessel's integrity during a postulated emergency condition involving a loss of coolant, when the emergency core cooling water subjects the pressure vessel to a thermal shock. There are two aspects to the problem: (a) the initial extension of a defect into the vessel wall, and (b) the subsequent re-initiation of fracture at an arrested crack tip. This note considers the effect of warm prestress on the re-initiation of fracture from an arrested crack, and emphasizes the role of ductile ligaments. It is argued that the warm prestress concept is applicable, thus complementing the limited experimental results provided by the HSST Thermal Shock experimental programme. (orig.)

Study on the application of 50 mm thick welded joints without PWHT for containment vessels

Energy Technology Data Exchange (ETDEWEB)

Watanabe, Nozomu; Sakai, Yoshiyuki; Hayashi, Kazutoshi; Higashikubo, Tomohiro (Mitsubishi Heavy Industries. Ltd., Kobe Shipyard and Machinery Works (Japan)); Iida, Kunihiro (Shibaura Inst. of Tech., Dept. of Mechanical Engineering, Tokyo (Japan)); Satou, Masanobu (Mitsubishi Heavy Industries. Ltd., Tkasago Research and Development Center (Japan))

1992-01-01

In order to investigate the propriety of the use of 50 mm thick SGV480 carbon steel which is equivalent to ASTM A516 Gr. 70 without post weld heat treatment for containment vessels, the authors have certified the basic properties of base metal and welded joints of 50 mm thick SGV480 steel plates. The results showed that fracture thoughness of welded joints is high without PWHT and the steel is safe enough without PWHT against embrittlement fracture under the operating conditions. (orig.).

Study on the application of 50 mm thick welded joints without PWHT for containment vessels

International Nuclear Information System (INIS)

Watanabe, Nozomu; Sakai, Yoshiyuki; Hayashi, Kazutoshi; Higashikubo, Tomohiro; Iida, Kunihiro; Satou, Masanobu

1992-01-01

In order to investigate the propriety of the use of 50 mm thick SGV480 carbon steel which is equivalent to ASTM A516 Gr. 70 without post weld heat treatment for containment vessels, the authors have certified the basic properties of base metal and welded joints of 50 mm thick SGV480 steel plates. The results showed that fracture thoughness of welded joints is high without PWHT and the steel is safe enough without PWHT against embrittlement fracture under the operating conditions. (orig.)

Integrated conjugate heat transfer analysis method for in-vessel retention with external reactor vessel cooling - 15477

International Nuclear Information System (INIS)

Park, J.W.; Bae, J.H.; Seol, W.C.

2015-01-01

An integrated conjugate heat transfer analysis method for the thermal integrity of a reactor vessel under external reactor vessel cooling conditions is developed to resolve light metal layer focusing effect issue. The method calculates steady-state 3-dimensional temperature distribution of a reactor vessel using coupled conjugate heat transfer between in-vessel 3-layered stratified corium (metallic pool, oxide pool and heavy metal) and polar-angle dependent boiling heat transfer at the outer surface of a reactor vessel. The 3-layer corium heat transfer model is utilizing lumped-parameter thermal-resistance circuit method and ex-vessel boiling regimes are parametrically considered. The thermal integrity of a reactor vessel is addressed in terms of un-molten thickness profile. The vessel 3-dimensional heat conduction is validated against a commercial code. It is found that even though the internal heat flux from the metal layer goes far beyond critical heat flux (CHF) the heat flux from the outermost nodes of the vessel may be maintained below CHF due to massive vessel heat diffusion. The heat diffusion throughout the vessel is more pronounced for relatively low heat generation rate in an oxide pool. Parametric calculations are performed considering thermal conditions such as peak heat flux from a light metal layer, heat generation in an oxide pool and external boiling conditions. The major finding is that the most crucial factor for success of in-vessel retention is not the mass of the molten light metal above the oxide pool but the heat generation rate inside the oxide pool and the 3-dimensional vessel heat transfer provides a much larger minimum vessel wall thickness. (authors)

Method of burying vessel containing radioactive waste

International Nuclear Information System (INIS)

Koga, Yoshihito.

1989-01-01

A float having an inert gas sealed therein is attached to a tightly closed vessel containing radioactive wastes. The vessel is inserted and kept in a small hole for burying the tightly closed vessel in an excavated shaft in rocks such as of granite or rock salts, while filling bentonite as shielding material therearound. In this case, the float is so adjusted that the apparent specific gravity is made equal or nearer between the tightly closed vessel and the bentonite, so that the rightly closed vessel does not sink and cause direct contact with the rocks even if bentonite flows due to earthquakes, etc. This can prevent radioactivity contamination through water in the rocks. (S.K.)

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)

Thermo-mechanical behaviour of FBTR reactor vessel due to natural convection in cover gas space

International Nuclear Information System (INIS)

Srinivasan, G.; Varadarajan, S.; Kapoor, R.P.

1988-01-01

Fast Breeder Test Reactor is a 40 MW(t), loop type sodium cooled reactor, similar in design to Rapsodie. The Reactor Assembly, which is the heart of FBTR, comprises the Reactor Vessel (RV) housed in a safety vessel within a concrete cell (A1 Cell). The RV which supports the core is shielded at the top by two rotatable plugs which are stacked with layers of borated graphite and steel. The smaller plug (SRP), is mounted excentric to the larger one (LRP). A nominal annular gap of 16 mm is provided between RV and LRP and between LRP and SRP to enable free rotation of the plugs. Stainless Steel insulation is fixed inside the steel vessel, to avoid overheating of the A1 Cell concrete. The core is supported by the Grid Plate (GP), bolted to the RV. During preheating, sodium charging and isothermal runs upto 350 0 C, temperature asymmetries were noticed in the reactor vessel wall in the cover gas space. This was attributable to convection currents in the annulus between RV and LRP. The asymmetries also resulted in a lateral shift of the grid plate. This paper discusses our experience in suppressing these convection currents, and minimising the grid plate shift

Reactor pressure vessel behaviour with a small crack in the cladding

International Nuclear Information System (INIS)

Fayolle, P.; Churier-Bossennec, H.; Faidy, C.

1990-01-01

This paper reports on fracture mechanic analysis of a PWR reactor pressure vessel with a 3.5 mm embedded circumferential crack in the cladding under a small lost of cooling accident transient. Different RTNDT level and effect of irradiation on material properties are considered. The study compares simplified one-dimensional and two-dimensional elastic approach and complete elastoplastic approach using J-parameter. The results show: good correlation between the different elastic approaches, important conservatism of the elastic approach compared to elastoplastic approach, no influence of irradiated material properties. The behavior of a vessel with this type of crack is acceptable for RTNDT less than 135 deg and safety injection temperature of 60 deg

Detection of travel time delay caused by dilation of an artificial fracture due to pressurization; Jinko chika kiretsu kaatsu ni tomonau toka danseiha denpa jikan henka no kenshutsu

Energy Technology Data Exchange (ETDEWEB)

Tanaka, K; Moriya, H; Asanuma, H; Niitsuma, H [Tohoku University, Sendai (Japan). Faculty of Engineering

1996-10-01

By revealing the relation between dilation of a subsurface fracture due to pressurization and travel time delay, it may be possible to measure the information as to the subsurface fracture system as a geothermal reservoir. In this study, field experiment was conducted to clarify the relation between the travel time delay of elastic waves and the dilation of fracture, pressure, and incident angles. The travel time delay of P-wave and S-wave tended to increase with the pressurization. When incident angle was about 90{degree} against the fracture, the increase was ranging between 0 and 0.2 ms. The magnitude of this delay could not be explained only by the opening of main fracture. It was considered that there were micro-crack zones around the main fracture. The difference of P-S delay depended on the pressurization and change of the pressure. The delay depended on the incident angle against the fracture. The delay of S-wave showed the polarized wave direction dependency. However, the obtained results might greatly depend on the analytical method and parameters. 4 refs., 10 figs.

The incidence of pelvic fractures with traumatic lower limb amputation in modern warfare due to improvised explosive devices.

Science.gov (United States)

Cross, A M; Davis, C; Penn-Barwell, J; Taylor, D M; De Mello, W F; Matthews, J J

2014-01-01

A frequently-seen injury pattern in current military experience is traumatic lower limb amputation as a result of improvised explosive devices (IEDs). This injury can coexist with fractures involving the pelvic ring. This study aims to assess the frequency of concomitant pelvic fracture in IED-related lower limb amputation. A retrospective analysis of the trauma charts, medical notes, and digital imaging was undertaken for all patients arriving at the Emergency Department at the UK military field hospital in Camp Bastion, Afghanistan, with a traumatic lower limb amputation in the six months between September 2009 and April 2010, in order to determine the incidence of associated pelvic ring fractures. Of 77 consecutive patients with traumatic lower limb amputations, 17 (22%) had an associated pelvic fracture (eleven with displaced pelvic ring fractures, five undisplaced fractures and one acetabular fracture). Unilateral amputees (n = 31) had a 10% incidence of associated pelvic fracture, whilst 30 % of bilateral amputees (n = 46) had a concurrent pelvic fracture. However, in bilateral, trans-femoral amputations (n = 28) the incidence of pelvic fracture was 39%. The study demonstrates a high incidence of pelvic fractures in patients with traumatic lower limb amputations, supporting the routine pre-hospital application of pelvic binders in this patient group.

Facial ulcerations due to Acinetobacter baumannii: Vessel thrombosis with bacterial mycelia

Directory of Open Access Journals (Sweden)

Dong Ming Li

2014-01-01

Full Text Available A 14-year-old girl presented with a 2-week history of progressive facial ulcerations that did not respond to cephalexin and topical dexamethasone. Biopsy on the ulcer showed rod-shaped bacteria and actinomycetes-like mycelia in the vessel walls and within thrombi. Tissue culture yielded Acinetobacter baumannii, which was resistant to cephalexin. A favourite outcome was achieved with minocycline treatment. This is the first case report of A. baumannii-related vasculitis.

Prestressed reactor vessel for nuclear power plants

International Nuclear Information System (INIS)

Schoening, J.; Schwiers, H.G.

1982-01-01

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

Preliminary assessment of the fracture behavior of weld material in full-thickness clad beams

International Nuclear Information System (INIS)

Keeney, J.A.; Bass, B.R.; McAfee, W.J.; Iskander, S.K.

1994-10-01

This report describes a testing program that utilizes full-thickness clad beam specimens to quantify fracture toughness for shallow cracks in material for which metallurgical conditions are prototypic of those found in reactor pressure vessels (RPVs). The beam specimens are fabricated from a section of an RPV wall (removed from a canceled nuclear plant) that includes weld, plate, and clad material. Metallurgical factors potentially influencing fracture toughness for shallow cracks in the beam specimens include material gradients due to welding and cladding applications, as well as material inhomogeneities in welded regions due to reheating in multiple weld passes. A summary of the testing program includes a description of the specimen geometry, material properties, the testing procedure, and the experimental results form three specimens. The yield strength of the weld material was determined to be 36% higher than the yield strength of the base material. An irradiation-induced increase in yield strength of the weld material could result in a yield stress that exceeds the upper limit where code curves are valid. The high yield strength for prototypic weld material may have implications for RPV structural integrity assessments. Analyses of the test data are discussed, including comparisons of measured displacements with finite-element analysis results, applications of toughness estimation techniques, and interpretations of constraint conditions implied by stress-based constraint methodologies. Metallurgical conditions in the region of the cladding heat-affected zone are proposed as a possible explanation for the lower-bound fracture toughness measured with one of the shallow-crack clad beam specimens. Fracture toughness data from the three clad beam specimens are compared with other shallow- and deep-crack uniaxial beam and cruciform data generated previously from A 533 Grade B plate material

Progress in generating fracture data base as a function of loading rate and temperature using small-scale tests

International Nuclear Information System (INIS)

Couque, H.; Hudak, S.J. Jr.

1993-01-01

Structural integrity assessment of nuclear pressure vessels requires small specimen fracture testing to generate data over a wide range of material loading, and temperature conditions. Small scale testing is employed since extensive testing is required including small radiation embrittled samples from nuclear surveillance capsules. However, current small scale technology does not provide the needed dynamic fracture toughness relevant to the crack arrest/reinitiation events that may occur during pressurized thermal shock transients following emergency shutdown. This paper addresses the generation of this much needed dynamic toughness data using a novel experimental-computational approach involving a coupled pressure bars (CPB) technique and a viscoplastic dynamic fracture code. CPB data have been generated to testing temperatures never before reached: 37 to 100 degrees C -- 60 to 123 degrees C above the nil ductility transition temperature. Fracture behavior of pressure vessel steel from lower shelf to upper shelf temperatures and previous toughness estimates for the 10 6 MPa√m s -1 loading rate regime are assessed in light of the new CPB data. 26 refs., 14 figs., 3 tabs

early operative management of pilon fractures using the ...

African Journals Online (AJOL)

Background: Pilon fractures is a management challenge due to complexity of fracture pattern and ... excellent exposure for accurate fracture reduction and no wound complications in our case ... dehiscence, superficial infection, osteomyelitis,.

A powerful methodology for reactor vessel pressurized thermal shock analysis

International Nuclear Information System (INIS)

Boucau, J.; Mager, T.

1994-01-01

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

Offshore wind transport and installation vessel

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

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

Concepts and possibilities of fracture mechanics for fracture safety assessment

International Nuclear Information System (INIS)

Blauel, J.

1980-01-01

In very tough materials for pressure vessels and pipelines of nuclear plants, cracking begins in a stable manner and only after macroscopic plastic deformations and crack blunting. It is possible to describe this elasto-plastic fracture behaviour and to quantify the safety margin compared to the assessment criteria based on linear elastic stressing and initiation by the concept of the J integral, the crack peak width and the crack resistance Jsub(R) curve. The numerous problems of details still open and the partly very limited validity range should not prevent the further investigation into the great possibilities of this concept and making greater use of the interpretation of large scale tests. (orig./RW) [de

Fracture toughness of A533B Part III - variability of A533B fracture toughness as determined from Charpy data

International Nuclear Information System (INIS)

Druce, S.G.; Eyre, B.L.

1978-08-01

This is the final part of a series of three reports examining the upper shelf fracture toughness of A533B Class 1 pressure vessel steel. Part I (AERE R 8968) critically reviews the current elasto plastic fracture mechanics methodologies employed to characterise toughness following extensive yielding and Part II (AERE R 8969) examines several sources of fracture mechanics data pertinent to A533B Class 1 in the longitudinal (RW) orientation. Part III is a review of the effects of (i) position and orientation within the plate (ii) welding processes and post weld heat treatment and (iii) neutron irradiation as measured by Charpy impact testing. It is concluded that the upper shelf factor energy is dependent on orientation and position and can be reduced by welding, extended post weld heat treatments and neutron irradiation. Neutron irradiation effects are known to be strongly dependent on composition and metallurgical conditions, but an explanation for the variability following extended post weld treatments has yet to be resolved. (author)

Opinion of the IRSN on serviceability of the 900 MWe reactor vessel - Answers to demands of the Nuclear Permanent Department of December 2005 - Mechanical aspect

International Nuclear Information System (INIS)

2010-05-01

As three demands had been made to EDF in December 2005 regarding the serviceability and more particularly the mechanical behaviour of the 900 MWe reactor vessels, this report discusses the evolution brought to models and proposed by EDF to correct the defect plasticity and take residual stresses into account. This discussion notably concerns the defect height and length range, and the admissible residual stress, but also the use of safety coefficients, transient application, fluence and the brittle-ductile transition temperature. This report from the French Nuclear Safety and Radioprotection Institute (IRSN) outlines the failure risks associated to the vessel in some specific nuclear power stations. Recommendations are made regarding the residual stress amplitude, the risk of fracture by cleavage, and actions to correct fracture risk margins on vessels which do not comply with regulatory criteria

THE HIP FRACTURE IS AN INJURY AND A DISEASE AT THESAME TIME

Directory of Open Access Journals (Sweden)

Radko Komadina

2008-12-01

Very useful is Charlson’s comorbidity index with 19 typical geriatric diseases, predictingdeath in hospitalized elderly with fragility fracture (heart, lung, kidneys, vessels, DM,tumor, liver, dementia, coagulopathies. If the patient has not any comorbidity, his oneyear mortality is estimated to be 12 %. With 1–2 comorbidities estimated mortality is 26 %,with 3–4 comorbidities 52 %, with 5 or more comorbidities the mortality is above 85 %. Inhip fracture Charlson index is on average 3.4. All comorbidities benefit from early operation and early mobilization

Coupled Fracture and Flow in Shale in Hydraulic Fracturing

Science.gov (United States)

Carey, J. W.; Mori, H.; Viswanathan, H.

2014-12-01

Production of hydrocarbon from shale requires creation and maintenance of fracture permeability in an otherwise impermeable shale matrix. In this study, we use a combination of triaxial coreflood experiments and x-ray tomography characterization to investigate the fracture-permeability behavior of Utica shale at in situ reservoir conditions (25-50 oC and 35-120 bars). Initially impermeable shale core was placed between flat anvils (compression) or between split anvils (pure shear) and loaded until failure in the triaxial device. Permeability was monitored continuously during this process. Significant deformation (>1%) was required to generate a transmissive fracture system. Permeability generally peaked at the point of a distinct failure event and then dropped by a factor of 2-6 when the system returned to hydrostatic failure. Permeability was very small in compression experiments (fashion as pressure increased. We also observed that permeability decreased with increasing fluid flow rate indicating that flow did not follow Darcy's Law, possibly due to non-laminar flow conditions, and conformed to Forscheimer's law. The coupled deformation and flow behavior of Utica shale, particularly the large deformation required to initiate flow, indicates the probable importance of activation of existing fractures in hydraulic fracturing and that these fractures can have adequate permeability for the production of hydrocarbon.

Rehabilitation after falls and fractures.

Science.gov (United States)

Dionyssiotis, Y; Dontas, I A; Economopoulos, D; Lyritis, G P

2008-01-01

Falls are one of the most common geriatric problems threatening the independence of older persons. Elderly patients tend to fall more often and have a greater tendency to fracture their bones. Fractures occur particularly in osteoporotic people due to increased bone fragility, resulting in considerable reduction of quality of life, morbidity, and mortality. This article provides information for the rehabilitation of osteoporotic fractures pertaining to the rehabilitation of the fractured patient, based on personal experience and literature. It also outlines a suggested effective and efficient clinical strategy approach for preventing falls in individual patients.

Clavicular fractures: Classification, diagnosis, therapy

International Nuclear Information System (INIS)

Schunk, K.; Strunk, H.; Schild, H.; Lohr, S.

1988-01-01

Clavicular fracture is one of the most frequent skeletal lesions. In most cases the median third of the clavicula is affected (this is due to the peculiar biomechanical structure). Accompanying lesions and complications of clavicular fractures are rare. A total of 13 X-ray diagnostic techniques are described of clavicular fractures. X-ray film should, as a matter of principle, always be taken in two planes. Definitely the major part of clavicular fractures are treated conservatively (rucsac dressing), whereas surgery is reserved for few and strictly defined indications. (orig.) [de

Clavicular fractures: Classification, diagnosis, therapy

Energy Technology Data Exchange (ETDEWEB)

Schunk, K.; Strunk, H.; Schild, H.; Lohr, S.

1988-09-01

Clavicular fracture is one of the most frequent skeletal lesions. In most cases the median third of the clavicula is affected (this is due to the peculiar biomechanical structure). Accompanying lesions and complications of clavicular fractures are rare. A total of 13 X-ray diagnostic techniques are described of clavicular fractures. X-ray film should, as a matter of principle, always be taken in two planes. Definitely the major part of clavicular fractures are treated conservatively (rucsac dressing), whereas surgery is reserved for few and strictly defined indications.

Thoracic Outlet Syndrome in a Volleyball Player Due to Nonunion of the First Rib Fracture.

Science.gov (United States)

Puttmann, Kathleen T; Satiani, Bhagwan; Vaccaro, Patrick

2016-11-01

Fracture of the first rib with ensuing callus formation is a rare cause of thoracic outlet syndrome. We report a case of a 17-year-old female volleyball player who presented with months of chronic arm pain. Radiographic imaging demonstrated nonunion fracture of the first rib. Physical therapy had been unsuccessful in relieving the pain, and surgical management was performed with resection of the first rib through a transaxillary approach with complete resolution of symptoms. Inflammation surrounding such fractures may destroy tissue planes, making dissection more technically difficult.

Elastic-plastic fracture mechanics study of thermal shock cracking

International Nuclear Information System (INIS)

Hirano, K.; Kobayashi, H.; Nakazawa, H.

1980-01-01

This paper describes thermal shock experiments conducted on a nuclear pressure vessel steel (A533 Grade B Class 1), an AISI304 steel and a tool steel (JIS SKD62) using both a new thermal shock test facility and method. Analysis of their quasi-static thermal stress intensity factors is performed on the basis of linear-elastic fracture mechanics; and a thermal shock fracture toughness value, Ksub(tsc) is evaluated. Then elastic-plastic fracture toughness tests are carried out in the same high temperature range of the thermal shock experiment, and a relation between the stretched zone width, SZW, formed as a result of the fatigue precrack tip plastic blunting and the J-integral is clarified. An elastic-plastic thermal shock fracture toughness value, Jsub(tsc), is evaluated from a critical value of the stretched zone width, SZWsub(tsc), at the initiation of the thermal shock cracking by using the relation between SZW and J. The Jsub(tsc) value is compared with an elastic-plastic fracture toughness value, Jsub(Ic), and the difference between these Jsub(tsc) and Jsub(Ic) values is discussed on the basis of fractography. (author)

Statistical evaluation of fracture characteristics of RPV steels in the ductile-brittle transition temperature region

International Nuclear Information System (INIS)

Kang, Sung Sik; Chi, Se Hwan; Hong, Jun Hwa

1998-01-01

The statistical analysis method was applied to the evaluation of fracture toughness in the ductile-brittle transition temperature region. Because cleavage fracture in steel is of a statistical nature, fracture toughness data or values show a similar statistical trend. Using the three-parameter Weibull distribution, a fracture toughness vs. temperature curve (K-curve) was directly generated from a set of fracture toughness data at a selected temperature. Charpy V-notch impact energy was also used to obtain the K-curve by a K IC -CVN (Charpy V-notch energy) correlation. Furthermore, this method was applied to evaluate the neutron irradiation embrittlement of reactor pressure vessel(RPV) steel. Most of the fracture toughness data were within the 95 percent confidence limits. The prediction of a transition temperature shift by statistical analysis was compared with that from the experimental data. (author)

Electrical discharge machining for vessel sample removal

International Nuclear Information System (INIS)

Litka, T.J.

1993-01-01

Due to aging-related problems or essential metallurgy information (plant-life extension or decommissioning) of nuclear plants, sample removal from vessels may be required as part of an examination. Vessel or cladding samples with cracks may be removed to determine the cause of cracking. Vessel weld samples may be removed to determine the weld metallurgy. In all cases, an engineering analysis must be done prior to sample removal to determine the vessel's integrity upon sample removal. Electrical discharge machining (EDM) is being used for in-vessel nuclear power plant vessel sampling. Machining operations in reactor coolant system (RCS) components must be accomplished while collecting machining chips that could cause damage if they become part of the flow stream. The debris from EDM is a fine talclike particulate (no chips), which can be collected by flushing and filtration

Microstructural parameters and yielding in a quenched and tempered Cr-Mo-V pressure vessel steel

International Nuclear Information System (INIS)

Toerroenen, Kari

1979-01-01

In this work the plastic deformation behaviour of a Cr-Mo-V pressure vessel steel is studied at ambient and low temperatures. To produce a wide range of microstructures, different austenitizing, quenching and tempering treatments are performed. The microstructures, including grain and dislocation structures as well as carbides, are evaluated. A qualitative model is proposed for the martensitic and bainitic transformations explaining the morphology and crystallography of the transformation products. Based on microstructural observations of undeformed and deformed materials, as well as the tensile test results, the role of various obstacles to dislocation motion in plastic deformation is evaluated. Finally the strength increment, its temperature dependence and the effect due to combinations of various obstacles are analyzed. The results are intended to serve as basis for further fracture behaviour analyses. (author)

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

International Nuclear Information System (INIS)

Cheverton, R.D.

1982-01-01

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