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

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

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.

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

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

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.

Computational evaluation of the constraint loss on the fracture toughness of reactor pressure vessel steels

International Nuclear Information System (INIS)

Serrano Garcia, M.

2007-01-01

The Master Curve approach is included on the ASME Code through some Code Cases to assess the reactor pressure vessel integrity. However, the margin definition to be added is not defined as is the margin to be added when the Master Curve reference temperature T 0 is obtained by testing pre-cracked Charpy specimens. The reason is that the T 0 value obtained with this specimen geometry is less conservative than the value obtained by testing compact tension specimens possible due to a loss of constraint. The two parameter fracture mechanics, considered as an extension of the classical fracture mechanics, coupled to a micromechanical fracture models is a valuable tool to assess the effect of constraint loss on fracture toughness. The definition of a parameter able to connect the fracture toughens value to the constraint level on the crack tip will allow to quantify margin to be added to the T 0 value when this value is obtained testing the pre-cracked Charpy specimens included in the surveillance capsule of the reactor pressure vessel. The Nuclear Regulatory Commission (NRC) define on the To value obtained by testing compact tension specimens and ben specimens (as pre-cracked Charpy are) bias. the NRC do not approved any of the direct applications of the Master Curve the reactor pressure vessel integrity assessment until this bias will be quantified in a reliable way. the inclusion of the bias on the integrity assessment is done through a margin to be added. In this thesis the bias is demonstrated an quantified empirical and numerically and a generic value is suggested for reactor pressure vessel materials, so that it can be used as a margin to be added to the T 0 value obtained by testing the Charpy specimens included in the surveillance capsules. (Author) 111 ref

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.

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

Metallurgical failure investigation of a pipe connector fracture of an expansion vessel

International Nuclear Information System (INIS)

Neidel, Andreas

2016-01-01

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.

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

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

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)

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.

Mild traumatic brain injury diagnosis frequently remains unrecorded in subjects with craniofacial fractures.

Science.gov (United States)

Puljula, Jussi; Cygnel, Hanna; Mäkinen, Elina; Tuomivaara, Veli; Karttunen, Vesa; Karttunen, Ari; Hillbom, Matti

2012-12-01

Traumatic brain injuries (TBI) in subjects with craniofacial fractures are usually diagnosed by emergency room physicians. We investigated how often TBI remains unrecorded in these subjects, and whether diagnostic accuracy has improved after the implementation of new TBI guidelines. All subjects with craniofacial fractures admitted to Oulu University Hospital in 1999 and in 2007 were retrospectively identified. New guidelines for improving the diagnostic accuracy of TBI were implemented between 2000 and 2006. Clinical symptoms of TBI were gathered from notes on hospital charts and compared to the recorded diagnoses at discharge. Logistic regression was used to identify independent predictors for TBI to remain unrecorded. Of 194 subjects with craniofacial fracture, 111(57%) had TBI, 40 in 1999 and 71 in 2007. Fifty-one TBIs (46%) remained unrecorded at discharge, 48 being mild and 3 moderate-to-severe. Subjects with unrecorded TBI were significantly less frequently referred to follow-up visits. Failures to record the TBI diagnosis were less frequent (29/71, 41%) in 2007 than in 1999 (22/40, 55%), but the difference was not statistically significant. The most significant independent predictor for this failure was the clinical specialty (other than neurology/neurosurgery) of the examining physician (palcohol intoxication did not hamper the diagnosis of TBI. TBIs remain frequently unrecorded in subjects with craniofacial fractures. Recording of mild TBI slightly but insignificantly improved after the implementation of new guidelines. Copyright © 2012 Elsevier Ltd. All rights reserved.

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

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)

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

International Nuclear Information System (INIS)

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

1998-01-01

Analyses were performed to determine the impact of weld residual stresses in a reactor pressure vessel (RPV) on (1) the generation of pressure temperature (P-T) curves required for maintaining specified fracture prevention margins during nuclear plant startup and shutdown, and (2) the conditional probability of vessel failure due to pressurized thermal shock (PTS) loading. The through wall residual stress distribution in an axially oriented weld was derived using measurements taken from a shell segment of a canceled RPV and finite element thermal stress analyses. The P-T curve derived from the best estimate load analysis and a t / 8 deep flaw, based on K Ic , was less limiting than the one derived from the current methodology prescribed in the ASME Boiler and Pressure Vessel Code. The inclusion of the weld residual stresses increased the conditional probability of cleavage fracture due to PTS loading by a factor ranging from 2 to 4

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.

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)

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)

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

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.

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

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

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)

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 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 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)

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.

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)

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

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

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

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-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

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.

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)

A fracture mechanics method of evaluating structural integrity of a reactor vessel due to thermal shock effects following LOCA condition

International Nuclear Information System (INIS)

Ramani, D.T.

1977-01-01

The importance of knowledge of structural integrity of a reactor vessel due to thermal shock effects, is related to safety and operational requirements in assessing the adequacy and flawless functioing of the nuclear power systems. Followig a loss-of-coolant accident (LOCA) condition the integrity of the reactor vessel due to a sudden thermal shock induced by actuation of emergency core cooling system (ECCS), must be maintained to ensure safe and orderly shutdown of the reactor and its components. The paper encompasses criteria underlaying a fracture mechanics method of analysis to evaluate structural integrity of a typical 950 MWe PWR vessel as a result of very drastic changes in thermal and mechanical stress levels in the reactor vessel wall. The main object of this investigation therefore consists in assessing the capability of a PWR vessel to withstand the most critical thermal shock without inpairing its ability to conserve vital coolant owing to probable crack propagation. (Auth.)

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)

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.

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.).

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

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

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

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.

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

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

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

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)

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

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

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.)

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

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

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

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

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)

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.

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

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.

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

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.)

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.

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

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

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

[Fracture of implant abutment screws and removal of a remaining screw piece

NARCIS (Netherlands)

Broeke, S.M. van den; Baat, C. de

2008-01-01

Fracture of the implant abutment screws is a complication which can render an implant useless. The prevalence of abutment screw fracture does not exceed 2.5% after 10 years. Causes are loosening of implant abutment screw, too few, too short or too narrow implants, implants not inserted perpendicular

Factors in the fail safe approach to pressure vessel assessment

International Nuclear Information System (INIS)

Connors, D.C.; Darlaston, B.J.L.; Hellen, R.A.J.

1978-01-01

The 'leak before break' concept is described in the context of pressure vessel assessment. The factors which determine whether a pipe containing an axial flaw will leak or break under pressure loading are discussed using a post-yield fracture mechanics method. A model is used in which it is assumed that initially the ligament beneath the flaw fails to form a full thickness defect in the pipe. The stability of the full thickness defect at the pressure causing ligament failure is then examined to ascertain whether it would remain at the snap through length or would propagate by fast fracture, to form a leak or a break. The method is used to analyse the results of a series of pipe rupture tests, and it is found that a distinction between leaks and breaks is achieved. (author)

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)

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

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 Shoulder Fractures Email to a friend * required fields ...

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.)

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

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

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)

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

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

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.)

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.

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

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

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.

A computational algorithm addressing how vessel length might depend on vessel diameter

Science.gov (United States)

Jing Cai; Shuoxin Zhang; Melvin T. Tyree

2010-01-01

The objective of this method paper was to examine a computational algorithm that may reveal how vessel length might depend on vessel diameter within any given stem or species. The computational method requires the assumption that vessels remain approximately constant in diameter over their entire length. When this method is applied to three species or hybrids in the...

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

The break remains – elderly people's experiences of a hip fracture 1 year after discharge.

Science.gov (United States)

Zidén, Lena; Scherman, Marianne Hansson; Wenestam, Claes-Göran

2010-01-01

To explore experienced long-term consequences of a hip fracture and conceptions of what influences hip fracture recovery among community-living elderly people 1 year after discharge. Fifteen subjects (13 females and 2 males), aged 66-94, were interviewed. The phenomenographic method was used for analysis. Experiences of insecurity and restricted life dominated the interviews. The descriptive categories within experienced consequences of a hip fracture were: (1) isolated life with more restricted activity and fewer social contacts, with the two sub-categories (a) more insecure and afraid and (b) more limited ability to move, (2) disappointed and sad that identity and life have changed and (3) satisfied with the situation or feeling even better than before the fracture. The categories within conceptions of what influences hip fracture recovery were: (4) own mind and actions influence recovery, (5) treatment and actions from others influence recovery and (6) you cannot influence recovery. The findings accentuate that the negative consequences of a hip fracture are substantial and long-lasting. As it strikes mostly elderly people, who may have experienced earlier losses and growing disabilities, a hip fracture could add to the risk of losing important life values. Furthermore, the findings indicate that all health care professionals who meet the patients need to consider the patients' own experiences and possible fear and not merely focus on the physical injury and disabilities.

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

Remaining life assessment of a high pressure turbine rotor

International Nuclear Information System (INIS)

Nguyen, Ninh; Little, Alfie

2012-01-01

This paper describes finite element and fracture mechanics based modelling work that provides a useful tool for evaluation of the remaining life of a high pressure (HP) steam turbine rotor that had experienced thermal fatigue cracking. An axis-symmetrical model of a HP rotor was constructed. Steam temperature, pressure and rotor speed data from start ups and shut downs were used for the thermal and stress analysis. Operating history and inspection records were used to benchmark the damage experienced by the rotor. Fracture mechanics crack growth analysis was carried out to evaluate the remaining life of the rotor under themal cyclic loading conditions. The work confirmed that the fracture mechanics approach in conjunction with finite element modelling provides a useful tool for assessing the remaining life of high temperature components in power plants.

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

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

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

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

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

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

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

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)

Variable flaw shape analysis for a reactor vessel under pressurized thermal shock loading

International Nuclear Information System (INIS)

Yang, C.Y.; Bamford, W.H.

1984-01-01

A study has been conducted to characterize the response of semi-elliptic surface flaws to thermal shock conditions which can result from safety injection actuation in nuclear reactor vessels. A methodology was developed to predict the behavior of a flaw during sample pressurized thermal shock events. The effects of a number of key variables on the flaw propagation were studied, including fracture toughness of the material and its gradient through the thickness, irradiation effects, effects of warm prestressing, and effects of the stainless steel cladding. The results of these studies show that under thermal shock loading conditions the flaw always tends to elongate along the vessel inside surface from the initial aspect ratio. However, the flaw shape always remains finite rather than becoming continuously long, as has often been assumed in earlier analyses. The final shape and size of the flaws were found to be rather strongly dependent on the effects of warm prestressing and the distribution of neutron flux. The improved methodology results in a more accurate and more realistic treatment of flaw shape changes during thermal shock events and provides the potential for quantifying additional margins for reactor vessel integrity analyses

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.

Mandible Fractures.

Science.gov (United States)

Pickrell, Brent B; Serebrakian, Arman T; Maricevich, Renata S

2017-05-01

Mandible fractures account for a significant portion of maxillofacial injuries and the evaluation, diagnosis, and management of these fractures remain challenging despite improved imaging technology and fixation techniques. Understanding appropriate surgical management can prevent complications such as malocclusion, pain, and revision procedures. Depending on the type and location of the fractures, various open and closed surgical reduction techniques can be utilized. In this article, the authors review the diagnostic evaluation, treatment options, and common complications of mandible fractures. Special considerations are described for pediatric and atrophic mandibles.

In-vessel tritium

International Nuclear Information System (INIS)

Ueda, Yoshio; Ohya, Kaoru; Ashikawa, Naoko; Ito, Atsushi M.; Kato, Daiji; Kawamura, Gakushi; Takayama, Arimichi; Tomita, Yukihiro; Nakamura, Hiroaki; Ono, Tadayoshi; Kawashima, Hisato; Shimizu, Katsuhiro; Takizuka, Tomonori; Nakano, Tomohide; Nakamura, Makoto; Hoshino, Kazuo; Kenmotsu, Takahiro; Wada, Motoi; Saito, Seiki; Takagi, Ikuji; Tanaka, Yasunori; Tanabe, Tetsuo; Yoshida, Masafumi; Toma, Mitsunori; Hatayama, Akiyoshi; Homma, Yuki; Tolstikhina, Inga Yu.

2012-01-01

The in-vessel tritium research is closely related to the plasma-materials interaction. It deals with the edge-plasma-wall interaction, the wall erosion, transport and re-deposition of neutral particles and the effect of neutral particles on the fuel recycling. Since the in-vessel tritium shows a complex nonlinear behavior, there remain many unsolved problems. So far, behaviors of in-vessel tritium have been investigated by two groups A01 and A02. The A01 group performed experiments on accumulation and recovery of tritium in thermonuclear fusion reactors and the A02 group studied theory and simulation on the in-vessel tritium behavior. In the present article, outcomes of the research are reviewed. (author)

The Impact of Type 2 Diabetes on Bone Fracture Healing

Directory of Open Access Journals (Sweden)

Carlos Marin

2018-01-01

Full Text Available Type 2 diabetes mellitus (T2DM is a chronic metabolic disease known by the presence of elevated blood glucose levels. Nowadays, it is perceived as a worldwide epidemic, with a very high socioeconomic impact on public health. Many are the complications caused by this chronic disorder, including a negative impact on the cardiovascular system, kidneys, eyes, muscle, blood vessels, and nervous system. Recently, there has been increasing evidence suggesting that T2DM also adversely affects the skeletal system, causing detrimental bone effects such as bone quality deterioration, loss of bone strength, increased fracture risk, and impaired bone healing. Nevertheless, the precise mechanisms by which T2DM causes detrimental effects on bone tissue are still elusive and remain poorly studied. The aim of this review was to synthesize current knowledge on the different factors influencing the impairment of bone fracture healing under T2DM conditions. Here, we discuss new approaches used in recent studies to unveil the mechanisms and fill the existing gaps in the scientific understanding of the relationship between T2DM, bone tissue, and bone fracture healing.

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)

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.

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.

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

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

Subcritical crack growth in the ligament between the instrumentation rods of the BBR pressure vessel bottom

International Nuclear Information System (INIS)

Marci, G.; Bazant, E.; Kautz, H.R.

1978-01-01

A fracture mechanics fatigue analysis is made for an assumed crack emanating from the bore of an instrumentation rod. This assumed crack has partially penetrated the Inconel buttering of the 22 Ni Mo Cr 37 on which the structural Inconel welds are laid. Our analysis shows that the assumed crack could only penetrate 26% of the remaining ligament of the Inconel structural weld as a result of the fatigue crack growth during the entire operating life of the pressure vessel. Therefore a leak caused by a flaw missed during pre-service and in-service non-destructive testing can be excluded. (author)

Model of T-Type Fracture in Coal Fracturing and Analysis of Influence Factors of Fracture Morphology

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Yuwei Li

2018-05-01

Full Text Available Special T-type fractures can be formed when coal is hydraulically fractured and there is currently no relevant theoretical model to calculate and describe them. This paper first establishes the height calculation model of vertical fractures in multi-layered formations and deduces the stress intensity factor (SIF at the upper and lower sides of the fracture in the process of vertical fracture extension. Combined with the fracture tip stress analysis method of fracture mechanics theory, the horizontal bedding is taken into account for tensile and shear failure, and the critical mechanical conditions for the formation of horizontal fracture in coal are obtained. Finally, the model of T-type fracture in coal fracturing is established, and it is verified by fracturing simulation experiments. The model calculation result shows that the increase of vertical fracture height facilitates the increase of horizontal fracture length. The fracture toughness of coal has a significant influence on the length of horizontal fracture and there is a threshold. When the fracture toughness is less than the threshold, the length of horizontal fracture remains unchanged, otherwise, the length of horizontal fracture increases rapidly with the increase of fracture toughness. When the shear strength of the interface between the coalbed and the interlayer increases, the length of the horizontal fracture of the T-type fracture rapidly decreases.

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)

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.

Analisis Remaining Life dan Penjadwalan Program Inspeksi pada Pressure Vessel dengan Menggunakan Metode Risk Based Inspection (RBI

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Dyah Arina Wahyu Lillah

2017-01-01

Full Text Available Seiring perkembangan eksplorasi minyak dan gas bumi di dunia, perusahaan minyak dan gas di Indonesia juga turut berlomba-lomba untuk mendapatkan ladang minyak dan gas bumi sebanyak-banyaknya. Perkembangan ini turut dipengaruhi oleh aturan-aturan pemerintah mengenai keselamatan dan pencegahan bahaya baik pada unit yang dikelola maupun tenaga kerja pengelola. Untuk itu semua perlatan-peralatan (unit kerja harus dijamin kehandalaannya agar tidak menimbulkan bahaya baik bagi pekerja maupun lingkungan. Subjek penelitian dalam tugas akhir ini ialah pada pressure vessel yang dimiliki oleh Terminal LPG Semarang. Kemungkinan bahaya yang dapat menyebabkan kerusakan pada pressure vessel perlu dianalisis agar dapat meminimalkan resiko yang akan terjadi. Metode Risk Based Inspection (RBI diharapkan dapat meminimalkan resiko yang ada pada pressure vessel. Penilaian resiko dalam tugas akhir ini mengacu pada standar API RP 581. Untuk mengetahui besarnya resiko yang ada pada plant, maka terlebih dahulu harus dihitung besarnya probabilitas kegagalan dan konsekuensi apabila terjadi kegagalan. Langkah selanjutnya ialah membandingkan besarnya resiko yang didapat dengan target resiko yang dimiliki oleh perusahaan. Dari hasil perbandingan ini dapat diketahui tingkat resiko pressure vessel, sehingga dapat ditentukan jadwal inspeksi dan metode inspeksi yang tepat.

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

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

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

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

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

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

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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.

Effect of length and diameter of fiber reinforced composite post (FRC on fracture resistance of remaining tooth structure

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Mahdiyeh seifi

2013-03-01

Full Text Available Introduction: Post and core has been considered for endodontically treated tooth, especially in cases with severe damage crowns. Recently fiber reinforced composite posts (FRC post have been used in the treatment of endodontically treated teeth. Because the length and diameter of posts are effective in stress distribution, the purpose of this study is to evaluate the effect of length and diameter of FRC post on fracture resistance. Methods: In this experimental study, 36 glass fiber posts with combination of 7mm, 9mm, and 12mm length and 1.1mm, 1.3mm and 1.5mm diameter were divided into 9 groups of 4. These posts were cemented in root canals by Panavia. Samples were tested with 45° compressive forces for the evaluation of fracture resistance. Datas were analyzed using SPSS soft ware and One- way and Two-way ANOVA analyses. Results: Fracture resistance did not increase significantly with the effect of length and diameter simultaneously (P=0.85. Samples with 12mm length and 1.5mm diameter had the greatest fracture resistance (1023/33N±239/22. The minimum fracture resistance had occurred in post with 7mm length and 1.5mm diameter (503/13N ±69/18. Fracture resistance increased significantly by increasing the length and the same diameter. Conclusion: It can be concluded that fracture resistance is affected by the length and not the diameter of FRC post.

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

Life extension of the BR2 aluminium vessel

International Nuclear Information System (INIS)

Koonen, E.; Fabry, A.; Chaouadi, R.; Verwerft, M.; Raedt, C. de; Winckel, S. van; Wacquier, W.; Dadoumont, J.; Verwimp, A.

2000-01-01

The BR2 reactor has recently undergone a major refurbishment comprising the replacement of all vessel internals. The vessel itself however was not replaced. An important requalification programme has been executed to prove that the vessel would remain fit during the contemplated life extension period of BR2. Representative material samples could be obtained from the shroud surrounding the vessel. A comprehensive in-service inspection was carried out and a vessel surveillance programme has been established. (author)

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.

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)

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

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

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)

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

[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.

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

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.)

Structural materials for ITER in-vessel component design

Energy Technology Data Exchange (ETDEWEB)

Kalinin, G. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Gauster, W. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Matera, R. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Tavassoli, A.-A.F. [CEA Centre d`Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France); Rowcliffe, A. [Oak Ridge National Lab., TN (United States); Fabritsiev, S. [Research Inst. of Electrophysical Apparatus, St. Petersburg (Russian Federation); Kawamura, H. [JAERI, IMTR Project, Ibaraki (Japan). Blanket Irradiation Lab.

1996-10-01

The materials proposed for ITER in-vessel components have to exhibit adequate performance for the operating lifetime of the reactor or for specified replacement intervals. Estimates show that maximum irradiation dose to be up to 5-7 dpa (for 1 MWa/m{sup 2} in the basic performance phase (BPP)) within a temperature range from 20 to 300 C. Austenitic SS 316LN-ITER Grade was defined as a reference option for the vacuum vessel, blanket, primary wall, pipe lines and divertor body. Conventional technologies and mill products are proposed for blanket, back plate and manifold manufacturing. HIPing is proposed as a reference manufacturing method for the primary wall and blanket and as an option for the divertor body. The existing data show that mechanical properties of HIPed SS are no worse than those of forged 316LN SS. Irradiation will result in property changes. Minimum ductility has been observed after irradiation in an approximate temperature range between 250 and 350 C, for doses of 5-10 dpa. In spite of radiation-induced changes in tensile deformation behavior, the fracture remains ductile. Irradiation assisted corrosion cracking is a concern for high doses of irradiation and at high temperatures. Re-welding is one of the critical issues because of the need to replace failed components. It is also being considered for the replacement of shielding blanket modules by breeding modules after the BPP. (orig.).

Multiple Sclerosis Increases Fracture Risk: A Meta-Analysis

Directory of Open Access Journals (Sweden)

Guixian Dong

2015-01-01

Full Text Available Purpose. The association between multiple sclerosis (MS and fracture risk has been reported, but results of previous studies remain controversial and ambiguous. To assess the association between MS and fracture risk, a meta-analysis was performed. Method. Based on comprehensive searches of the PubMed, Embase, and Web of Science, we identified outcome data from all articles estimating the association between MS and fracture risk. The pooled risk ratios (RRs with 95% confidence intervals (CIs were calculated. Results. A significant association between MS and fracture risk was found. This result remained statistically significant when the adjusted RRs were combined. Subgroup analysis stratified by the site of fracture suggested significant associations between MS and tibia fracture risk, femur fracture risk, hip fracture risk, pelvis fracture risk, vertebrae fracture risk, and humerus fracture risk. In the subgroup analysis by gender, female MS patients had increased fracture risk. When stratified by history of drug use, use of antidepressants, hypnotics/anxiolytics, anticonvulsants, and glucocorticoids increased the risk of fracture risk in MS patients. Conclusions. This meta-analysis demonstrated that MS was significantly associated with fracture risk.

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)

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

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

Reactor vessel decommissioning project. Final report

International Nuclear Information System (INIS)

Schoonen, D.H.

1984-09-01

This report describes a reactor vessel decommissioning project; it documents and explains the project objectives, scope, performance results, and sodium removal process. The project was successfully completed in FY-1983, within budget and without significant problems or adverse impact on the environment. Waste generated by the operation included the reactor vessel, drained sodium, and liquid, solid, and gaseous wastes which were significantly less than project estimates. Personnel radiation exposures were minimized, such that the project total was one-half the predicted exposure level. Except for the sodium removed, the material remaining in the reactor vessel is essentially the same as when the vessel arrived for processing

Fracturing formations in wells

Energy Technology Data Exchange (ETDEWEB)

Daroza, R A

1964-05-15

This well stimulation method comprises introducing through the well bore a low-penetrating, dilatant fluid, and subjecting the fluid to sufficient pressure to produce fractures in the formation. The fluid is permitted to remain in contact with the formation so as to become diluted by the formation fluids, and thereby lose its properties of dilatancy. Also, a penetrating fluid, containing a propping agent suspended therein, in introduced into contact with the fractures at a pressure substantially reduced with respect to that pressure which would have been required, prior to the fracturing operation performed using the low-penetrating dilatant fluid. The propping agent is deposited within the fractures, and thereafter, fluid production is resumed from the fractured formation. (2 claims)

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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)

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)

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

Fracture mechanics analysis and evaluation for the RPV of the Chinese Qinshan 300 MW NPP and PTS

International Nuclear Information System (INIS)

He Yinbiao; Isozaki, Toshikuni

2000-03-01

One of the most severe accident conditions of a reactor pressure vessel (RPV) in service is the loss of coolant accident (LOCA). Cold safety injection water is pumped into the downcomer of the RPV through inlet nozzles, while the internal pressure may remain at high level. Such an accident is called pressurized thermal shock (PTS) transient according to 10 CFR 50.61 definition. This paper illustrates the fracture mechanics analysis for the existing RPV of the Chinese Qinshan 300 MW nuclear power plant (NPP) under the postulated PTS transients that include SB-LOCA, LB-LOCA of Qinshan NPP and Rancho Seco transients. 3-D models with the flaw depth range a/w=0.05∼0.9 (a: flaw depth; w: wall thickness) were used to probe what kind of flaw and what kind of transient are most dangerous for the RPV in the end of life (EOF). Both the elastic and elastic-plastic material models were used in the stress analysis and fracture mechanics analysis. The different types of flaw and the influence of the stainless steel cladding on the fracture analysis were investigated for different PTS transients. comparing with the material initiation crack toughness K IC , the fracture evaluation for the RPV in question under PTS transients are performed in this paper. (author)

Benign compression fractures of the spine: signal patterns

International Nuclear Information System (INIS)

Ryu, Kyung Nam; Choi, Woo Suk; Lee, Sun Wha; Lim, Jae Hoon

1992-01-01

Fifteen patients with 38 compression fractures of the spine underwent magnetic resonance(MR) imaging. We retrospectively evaluated MR images in those benign compression fractures. MR images showed four patterns in T1-weighted images. MR imaging patterns were normal signal(21), band like low signal(8), low signal with preservation of peripheral portion of the body(8), and diffuse low signal through the vertebral body(1). The low signal portions were changed to high signal intensities in T2-weighted images. In 7 of 15 patients (11 compression fractures), there was a history of trauma, and the remaining 8 patients (27 compression fractures) had no history of trauma. Benign compression fractures of trauma, remained 8 patients (27 compression fractures) were non-traumatic. Benign compression fractures of the spine reveal variable signal intensities in MR imagings. These patterns of benign compression fractures may be useful in interpretation of MR imagings of the spine

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

Identification of a progenitor cell population destined to form fracture fibrocartilage callus in Dickkopf-related protein 3-green fluorescent protein reporter mice.

Science.gov (United States)

Mori, Yu; Adams, Douglas; Hagiwara, Yusuke; Yoshida, Ryu; Kamimura, Masayuki; Itoi, Eiji; Rowe, David W

2016-11-01

Fracture healing is a complex biological process involving the proliferation of mesenchymal progenitor cells, and chondrogenic, osteogenic, and angiogenic differentiation. The mechanisms underlying the proliferation and differentiation of mesenchymal progenitor cells remain unclear. Here, we demonstrate Dickkopf-related protein 3 (Dkk3) expression in periosteal cells using Dkk3-green fluorescent protein reporter mice. We found that proliferation of mesenchymal progenitor cells began in the periosteum, involving Dkk3-positive cell proliferation near the fracture site. In addition, Dkk3 was expressed in fibrocartilage cells together with smooth muscle α-actin and Col3.6 in the early phase of fracture healing as a cell marker of fibrocartilage cells. Dkk3 was not expressed in mature chondrogenic cells or osteogenic cells. Transient expression of Dkk3 disappeared in the late phase of fracture healing, except in the superficial periosteal area of fracture callus. The Dkk3 expression pattern differed in newly formed type IV collagen positive blood vessels and the related avascular tissue. This is the first report that shows Dkk3 expression in the periosteum at a resting state and in fibrocartilage cells during the fracture healing process, which was associated with smooth muscle α-actin and Col3.6 expression in mesenchymal progenitor cells. These fluorescent mesenchymal lineage cells may be useful for future studies to better understand fracture healing.

Clinical Utility of Chest Computed Tomography in Patients with Rib Fractures CT Chest and Rib Fractures

OpenAIRE

Chapman, Brandon C.; Overbey, Douglas M.; Tesfalidet, Feven; Schramm, Kristofer; Stovall, Robert T.; French, Andrew; Johnson, Jeffrey L.; Burlew, Clay C.; Barnett, Carlton; Moore, Ernest E.; Pieracci, Fredric M.

2016-01-01

Background Chest CT is more sensitive than a chest X-ray (CXR) in diagnosing rib fractures; however, the clinical significance of these fractures remains unclear. Objectives The purpose of this study was to determine the added diagnostic use of chest CT performed after CXR in patients with either known or suspected rib fractures secondary to blunt trauma. Methods Retrospective coho...

Neutron Assay System for Confinement Vessel Disposition

International Nuclear Information System (INIS)

Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Mayo, Douglas R.; Miko, David K.; Salazar, William R.; Stange, Sy; Valdez, Jose I.; 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-inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the CVs. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of special nuclear material (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 CVAS has been tested and calibrated in preparation for verification and safeguards measurements.

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.)

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)

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.

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.)

Method for temporary shielding of reactor vessel internals

International Nuclear Information System (INIS)

Grimm, N.P.; Sejvar, J.

1991-01-01

This patent describes a method for shielding stored internals for reactor vessel annealing. It comprises removing nuclear fuel from the reactor vessel containment building; removing and storing upper and lower core internals under water in a refueling canal storage area; assembling a support structure in the refueling canal between the reactor vessel and the stored internals; introducing vertical shielding tanks individually through a hatch in the containment building and positioning each into the support structure; introducing horizontal shielding tanks individually through a hatch in the containment building and positioning each above the stored internals and vertical tanks; draining water from the refueling canal to the level of a flange of the reactor vessel; placing an annealing apparatus in the reactor vessel; pumping the remaining water from the reactor vessel; and annealing the reactor vessel

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

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.

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)

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

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

Why ductile fracture mechanics

International Nuclear Information System (INIS)

Ritchie, R.O.

1983-01-01

Until recently, the engineering application of fracture mechanics has been specific to a description of macroscopic fracture behavior in components and structural parts which remain nominally elastic under loading. While this approach, termed linear elastic fracture mechanics, has been found to be invaluable for the continuum analysis of crack growth in brittle and high strength materials, it is clearly inappropriate for characterizing failure in lower strength ductile alloys where extensive inelastic deformation precedes and accompanies crack initiation and subsequent propagation. Accordingly, much effort has been devoted in recent years toward the development of nonlinear or ductile fracture mechanics methodology to characterize fracture behavior under elastic/plastic conditions; an effort which has been principally motivated by problems in nuclear industry. In this paper, the concepts of ductile (elastic/plastic) fracture mechanics are introduced and applied to the problem of both stationary and nonstationary cracks. Specifically, the limitations inherent in this approach are defined, together with a description of the microstructural considerations and applications relevant to the failure of ductile materials by fracture, fatigue, and creep

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

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

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

Lung vessel segmentation in CT images using graph-cuts

Science.gov (United States)

Zhai, Zhiwei; Staring, Marius; Stoel, Berend C.

2016-03-01

Accurate lung vessel segmentation is an important operation for lung CT analysis. Filters that are based on analyzing the eigenvalues of the Hessian matrix are popular for pulmonary vessel enhancement. However, due to their low response at vessel bifurcations and vessel boundaries, extracting lung vessels by thresholding the vesselness is not sufficiently accurate. Some methods turn to graph-cuts for more accurate segmentation, as it incorporates neighbourhood information. In this work, we propose a new graph-cuts cost function combining appearance and shape, where CT intensity represents appearance and vesselness from a Hessian-based filter represents shape. Due to the amount of voxels in high resolution CT scans, the memory requirement and time consumption for building a graph structure is very high. In order to make the graph representation computationally tractable, those voxels that are considered clearly background are removed from the graph nodes, using a threshold on the vesselness map. The graph structure is then established based on the remaining voxel nodes, source/sink nodes and the neighbourhood relationship of the remaining voxels. Vessels are segmented by minimizing the energy cost function with the graph-cuts optimization framework. We optimized the parameters used in the graph-cuts cost function and evaluated the proposed method with two manually labeled sub-volumes. For independent evaluation, we used 20 CT scans of the VESSEL12 challenge. The evaluation results of the sub-volume data show that the proposed method produced a more accurate vessel segmentation compared to the previous methods, with F1 score 0.76 and 0.69. In the VESSEL12 data-set, our method obtained a competitive performance with an area under the ROC curve of 0.975, especially among the binary submissions.

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)

TMI-2 Vessel Investigation Project integration report

Energy Technology Data Exchange (ETDEWEB)

Wolf, J. R.; Rempe, J. L.; Stickler, L. A.; Korth, G. E.; Diercks, D. R.; Neimark, L. A.; Akers, D W; Schuetz, B. K.; Shearer, T L; Chavez, S. A.; Thinnes, G. L.; Witt, R. J.; Corradini, M L; Kos, J. A. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

1994-03-01

The Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) was an international effort that was sponsored by the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. The primary objectives of the VIP were to extract and examine samples from the lower head and to evaluate the potential modes of failure and the margin of structural integrity that remained in the TMI-2 reactor vessel during the accident. This report presents a summary of the major findings and conclusions that were developed from research during the VIP. Results from the various elements of the project are integrated to form a cohesive understanding of the vessel`s condition after the accident.

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

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

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

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

46 CFR 2.10-101 - Annual vessel inspection fee.

Science.gov (United States)

2010-10-01

... inspections, damage surveys, repair and modification inspections, change in vessel service inspections, permit... of international certificates. (d) Entitlement to inspection services for the current year remains... MODUs 4,695 Semi-submersible MODUs 8,050 Nautical School Vessels: Length not greater than 100 feet 835...

Evaluation of the conditions imposed by the fracture surface geometry on water seepage through fractured porous media

International Nuclear Information System (INIS)

Fuentes, Nestor O.; Faybishenko, B.

2003-01-01

In order to determine the geometric patterns of the fracture surfaces that imposes conditions on the fluid flow through fractured porous media, a series a fracture models have been analyzed using the RIMAPS technique and the variogram method. Results confirm that the main paths followed by the fluid channels are determined by the surface topography and remain constant during water seepage evolution. Characteristics scale lengths of both situations: fracture surface and the flow of water, are also found. There exists a relationship between the scale lengths corresponding to each situation. (author)

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

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.)

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

Management of civilian ballistic fractures.

Science.gov (United States)

Seng, V S; Masquelet, A C

2013-12-01

The management of ballistic fractures, which are open fractures, has often been studied in wartime and has benefited from the principles of military surgery with debridement and lavage, and the use of external fixation for bone stabilization. In civilian practice, bone stabilization of these fractures is different and is not performed by external fixation. Fifteen civilian ballistic fractures, Gustilo II or IIIa, two associated with nerve damage and none with vascular damage, were reviewed. After debridement and lavage, ten internal fixations and five conservative treatments were used. No superficial or deep surgical site infection was noted. Fourteen of the 15 fractures (93%) healed without reoperation. Eleven of the 15 patients (73%) regained normal function. Ballistic fractures have a bad reputation due to their many complications, including infections. In civilian practice, the use of internal fixation is not responsible for excessive morbidity, provided debridement and lavage are performed. Civilian ballistic fractures, when they are caused by low-velocity firearms, differ from military ballistic fractures. Although the principle of surgical debridement and lavage remains the same, bone stabilization is different and is similar to conventional open fractures. Copyright © 2013 Elsevier Masson SAS. All rights reserved.

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.)

Fractures of the proximal humerus

DEFF Research Database (Denmark)

Brorson, Stig

2013-01-01

Fractures of the proximal humerus have been diagnosed and managed since the earliest known surgical texts. For more than four millennia the preferred treatment was forceful traction, closed reduction, and immobilization with linen soaked in combinations of oil, honey, alum, wine, or cerate......, classification of proximal humeral fractures remains a challenge for the conduct, reporting, and interpretation of clinical trials. The evidence for the benefits of surgery in complex fractures of the proximal humerus is weak. In three systematic reviews I studied the outcome after locking plate osteosynthesis...

Scaphoid fractures in children

Directory of Open Access Journals (Sweden)

Gajdobranski Đorđe

2014-01-01

Full Text Available Introduction. Scaphoid fractures are rare in childhood. Diagnosis is very difficult to establish because carpal bones are not fully ossified. In suspected cases comparative or delayed radiography is used, as well as computerized tomography, magnetic resonance imaging, ultrasound and bone scintigraphy. Majority of scaphoid fractures are treated conservatively with good results. In case of delayed fracture healing various types of treatment are available. Objective. To determine the mechanism of injury, clinical healing process, types and outcome of treatment of scaphoid fractures in children. Methods. We retrospectively analyzed patients with traumatic closed fracture of the scaphoid bone over a ten-year period (2002-2011. The outcome of the treatment of “acute” scaphoid fracture was evaluated using the Mayo Wrist Score. Results. There were in total 34 patients, of mean age 13.8 years, with traumatic closed fracture of the scaphoid bone, whose bone growth was not finished yet. Most common injury mechanism was fall on outstretched arm - 76% of patients. During the examined period 31 children with “acute” fracture underwent conservative treatment, with average immobilization period of 51 days. Six patients were lost to follow-up. In the remaining 25 patients, after completed rehabilitation, functional results determined by the Mayo Wrist Score were excellent. Conclusion. Conservative therapy of “acute” scaphoid fractures is an acceptable treatment option for pediatric patients with excellent functional results.

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.)

Change in fracture risk and fracture pattern after bariatric surgery: nested case-control study.

Science.gov (United States)

Rousseau, Catherine; Jean, Sonia; Gamache, Philippe; Lebel, Stéfane; Mac-Way, Fabrice; Biertho, Laurent; Michou, Laëtitia; Gagnon, Claudia

2016-07-27

surgery were more likely to have fractures than were obese or non-obese controls, and this risk remained higher after surgery. Fracture risk was site specific, changing from a pattern associated with obesity to a pattern typical of osteoporosis after surgery. Only biliopancreatic diversion was clearly associated with fracture risk; however, results for Roux-en-Y gastric bypass and sleeve gastrectomy remain inconclusive. Fracture risk assessment and management should be part of bariatric care. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://group.bmj.com/group/rights-licensing/permissions.

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)

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

Integrating Multiple Autonomous Underwater Vessels, Surface Vessels and Aircraft into Oceanographic Research Vessel Operations

Science.gov (United States)

McGillivary, P. A.; Borges de Sousa, J.; Martins, R.; Rajan, K.

2012-12-01

Autonomous platforms are increasingly used as components of Integrated Ocean Observing Systems and oceanographic research cruises. Systems deployed can include gliders or propeller-driven autonomous underwater vessels (AUVs), autonomous surface vessels (ASVs), and unmanned aircraft systems (UAS). Prior field campaigns have demonstrated successful communication, sensor data fusion and visualization for studies using gliders and AUVs. However, additional requirements exist for incorporating ASVs and UASs into ship operations. For these systems to be optimally integrated into research vessel data management and operational planning systems involves addressing three key issues: real-time field data availability, platform coordination, and data archiving for later analysis. A fleet of AUVs, ASVs and UAS deployed from a research vessel is best operated as a system integrated with the ship, provided communications among them can be sustained. For this purpose, Disruptive Tolerant Networking (DTN) software protocols for operation in communication-challenged environments help ensure reliable high-bandwidth communications. Additionally, system components need to have considerable onboard autonomy, namely adaptive sampling capabilities using their own onboard sensor data stream analysis. We discuss Oceanographic Decision Support System (ODSS) software currently used for situational awareness and planning onshore, and in the near future event detection and response will be coordinated among multiple vehicles. Results from recent field studies from oceanographic research vessels using AUVs, ASVs and UAS, including the Rapid Environmental Picture (REP-12) cruise, are presented describing methods and results for use of multi-vehicle communication and deliberative control networks, adaptive sampling with single and multiple platforms, issues relating to data management and archiving, and finally challenges that remain in addressing these technological issues. Significantly, the

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

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

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

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)

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.

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)

Assessment of reactor vessel integrity (ARVI)

Energy Technology Data Exchange (ETDEWEB)

Sehgal, B.R. E-mail: sehgal@ne.kth.se; Theerthan, A.; Giri, A.; Karbojian, A.; Willschuetz, H.G.; Kymaelaeinen, O.; Vandroux, S.; Bonnet, J.M.; Seiler, J.M.; Ikkonen, K.; Sairanen, R.; Bhandari, S.; Buerger, M.; Buck, M.; Widmann, W.; Dienstbier, J.; Techy, Z.; Kostka, P.; Taubner, R.; Theofanous, T.; Dinh, T.N

2003-04-01

The cost-shared project ARVI (assessment of reactor vessel integrity) involves a total of nine organisations from Europe and USA. The objective of the ARVI Project is to resolve the safety issues that remain unresolved for the melt vessel interaction phase of the in-vessel progression of a severe accident. The work consists of experiments and analysis development. Four tests were performed in the EC-FOREVER Programme, in which failure was achieved in-vessels employing the French pressure vessel steel. The tests were analysed with the commercial code ANSYS-Multiphysics, and the codes SYSTUS+ and PASULA, and quite good agreement was achieved for the failure location. Natural convection experiments in stratified pools have been performed in the SIMECO and the COPO facilities, which showed that much greater heat is transferred downwards for immiscible layers or before layers mix. A model for gap cooling and a set of simplified models for the system codes have been developed. MVITA code calculations have been performed for the Czech and Hungarian VVERs, towards evaluation of the in-vessel melt retention accident management scheme. Tests have been performed at the ULPU facility with organised flow for vessel external cooling. Considerable enhancement of the critical heat flux (CHF) was obtained. The ARVI Project has reached the halfway stage. This paper presents the results obtained thus far from the project.

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

Neutron Assay System for Con?nement Vessel Disposition

International Nuclear Information System (INIS)

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

2012-01-01

Waste will be removed from confinement vessels remaining from 1970s-era experiments. Los Alamos has 9+ spherical confinement vessels remaining from experiments. Each vessel contains ∼ 500 lbs of radioactive debris such as actinide metals and oxides, metals, powdered silica, graphite, and wires and hardware. In order to dispose of the vessels, debris and contamination must be removed. Neutron assay system was designed to assay vessels before and after cleanout. System requirements are: (1) Modular and moveable; (2) Capable of detecting ∼100g 239 Pu equivalent in a 2-inch thick steel sphere with 6 foot diameter; and (3) Capable of safeguards-quality assays. Initial design parameters arethe use of 4-atm 3 He tubes with length of 6 feet, and 3 He tubes embedded in polyethelene for moderation. This paper describes the calibration of the Confinement Vessel Assay System (CVAS) and quantification of its uncertainties. Assay uncertainty depends on five factors: (1) Statistical uncertainty in the assay measurement; (2) Statistical uncertainty in the background measurement; (3) Statistical uncertainty in the isotopics determination - This should be much smaller than the other uncertainties; (4) Systematic uncertainty due to position bias; and (5) Systematic uncertainty due to fluctuations in cosmic ray spallation. This one can be virtually eliminated by performing the background measurement with an empty vessel - but that may not be possible. We used modeling and experiments to quantify the systematic uncertainties. The calibration assumes a uniform distribution of material, but reality will be different. MCNPX modeling was used to quantify the positional bias. The model was benchmarked to build confidence in its results. Material at top of vessel is 44% greater than amount assayed, according to singles. Material near 19-tube detector is 38% less than amount assayed, according to singles. Cosmic ray spallation contributes significantly to the background. Comparing rates

Reactor pressure vessels safety and reliability - certainty and uncertainty

International Nuclear Information System (INIS)

O'Neil, R.

1977-01-01

In the paper, it is suggested that the hazard to the population which would result from vessel failure rate of the order of 10 -6 to 10 -7 per vessel year could be acceptable to society on the basis of other natural and man-made risks. The paper considers the problems of demonstrating safety by calculation based on fracture mechanics, and indicates some of the uncertainties, and inconsistencies in the theory, particularly the effect of cracks in locally degraded volumes of material. The phenomenon of crack arrest is considered, and attention is drawn to the uncertainties as indicated at least by some tests. There is need for speedy resolution of this problem. The uncertainties in material properties, heat treatment and residual stresses are considered, and a proposed upper limit for residual defects ('original sin') is proposed. (orig.) [de

Draft fracture mechanics code case for American Society of Mechanical Engineers NUPACK rules

International Nuclear Information System (INIS)

McConnell, P.; Sorenson, K.; Nickell, R.; Saegusa, T.

2004-01-01

The containment boundaries of most spent-fuel casks certified for use in the United States by the Nuclear Regulatory Commission are constructed with stainless steel, a material that is ductile in an engineering sense at all temperatures and for which, therefore, fracture mechanics principles are not relevant for the containment application. Ferritic materials may fail in a nonductile manner at sufficiently low temperatures, so fracture mechanics principles may be applied to preclude nonductile fracture. Because of the need to transport and store spent nuclear fuel safely in all types of climatic conditions, these vessels have regulatory lowest service temperatures that range down to -40 C (-40 F) for transport application. Such low service temperatures represent a severe challenge in terms of fracture toughness to many ferritic materials. Linear-elastic and elastic-plastic fracture mechanics principles provide a methodology for evaluating ferritic materials under such conditions

Materials surveillance program for C-E NSSS reactor vessels

International Nuclear Information System (INIS)

Koziol, J.J.

1977-01-01

Irradiation surveillance programs for light water NSSS reactor vessels provide the means by which the utility can assess the extent of neutron-induced changes in the reactor vessel materials. These programs are conducted to verify, by direct measurement, the conservatism in the predicted radiation-induced changes and hence the operational parameters (i.e., heat-up, cooldown, and pressurization rates). In addition, such programs provide assurance that the scheduled adjustments in the operational parameters are made with ample margin for safe operation of the plant. During the past 3 years, several documents have been promulgated establishing the criteria for determining both the initial properties of the reactor vessel materials as well as measurement of changes in these initial properties as a result of irradiation. These documents, ASTM E-185-73, ''Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels,'' and Appendix H to 10 CFR 50, ''Reactor Vessel Material Surveillance Program Requirements,'' are complementary to each other. They are the result of a change in the basic philosophy regarding the design and analysis of reactor vessels. In effect, the empirical ''transition temperature approach,'' which was used for design, was replaced by the ''analytical fracture mechanics approach.'' The implementation of this technique was described in Welding Research Council Bulletin 1975 and Appendix G to ASME Code Section III. Further definition of requirements appears in Appendix G to 10 CFR 50 published in July 1973. It is the intent of this paper to describe (1) a typical materials surveillance program for the reactor vessel of a Combustion Engineering NSSS, and (2) how the results of such programs, as well as experimental programs provide feed-back for improvement of materials to enhance their radiation resistance and thereby further improve the safety and reliability of future plants. (author)

Distribution of the In-Vessel Diagnostics in ITER Tokamak

Energy Technology Data Exchange (ETDEWEB)

González, Jorge, E-mail: Jorge.Gonzalez@iter.org [Rüecker Lypsa, Carretera del Prat, 65, Cornellá de Llobregat (Spain); Clough, Matthew; Martin, Alex; Woods, Nick; Suarez, Alejandro [ITER Organization, Route de Vinon sur Verdon-CS 90 046 13067 Saint Paul Lez Durance (France); Martinez, Gonzalo [Technical University Of Catalonia (UPC), Barcelona-Tech, Barcelona (Spain); Stefan, Gicquel; Yunxing, Ma [ITER Organization, Route de Vinon sur Verdon-CS 90 046 13067 Saint Paul Lez Durance (France)

2017-01-15

The ITER In-Vessel Diagnostics have been distributed around the In-Vessel shell to understand burning plasma physics and assist in machine operation. Each diagnostics component has its own requirements, constraints, and even exclusion among them for the highly complex In-Vessel environment. The size of the plasma, the requirement to be able to align the blanket system to the magnetic centre of the machine, the cooling requirements of the blanket system and the size of the pressure vessel itself all add to the difficulties of integrating these systems into the remaining space available. The available space for the cables inside the special trays (in-Vessel looms) is another constraint to allocate In-Vessel electrical sensors. Besides this, there are issues with the Assembly sequences and surface & volumetric neutron heating considerations that have imposed several additional restrictions.

Scaphoid Fracture

Directory of Open Access Journals (Sweden)

Esther Kim, BS

2018-04-01

-evaluated in two weeks with repeat plain films.4 Further evaluation with MRI or bone scan should be considered if repeat plain films are negative and clinical suspicion remains high.3 In conclusion, scaphoid fractures can be difficult to identify; therefore, a high level of suspicion and low threshold for further imaging or intervention is warranted. Topics: Scaphoid, fracture, orthopedics, FOOSH injury

Niobium Application, Metallurgy and Global Trends in Pressure Vessel Steels

Science.gov (United States)

Jansto, Steven G.

Niobium-containing high strength steel materials have been developed for a variety of pressure vessel applications. Through the application of these Nb-bearing steels in demanding applications, the designer and end user experience improved toughness at low temperature, excellent fatigue resistance and fracture toughness and excellent weldability. These enhancements provide structural engineers the opportunity to further improve the pressure vessel design and performance. The Nb-microalloy alloy designs also result in reduced operational production cost at the steel operation, thereby embracing the value-added attribute Nb provides to both the producer and the end user throughout the supply chain. For example, through the adoption of these Nb-containing structural materials, several design-manufacturing companies are considering improved designs which offer improved manufacturability, lower overall cost and better life cycle performance.

The metrological problems of irradiation embrittlement of reactor pressure vessel steel

International Nuclear Information System (INIS)

Vodenicharov, S.; Kamenova, Ts.

1993-01-01

Neutron irradiation of reactor pressure vessel steels increases the T k -values of transition temperature from ductile to brittle fracture. This effect is very important in emergency situations, when the water cooling injection in the reactor results in high thermal gradients. In such cases there is a risk from the appearance of a brittle fracture with catastrophic crack propagation speed at relatively low stresses. That is why the T k -value determination is very important for the safe operation of the reactor systems. Some advanced experimental methods for T k -testing and control have been discussed in the present article and the standards of different countries have been compared. The methods applying subsize specimens and welding-restored specimens have been reviewed. (author)

NDE and fracture mechanics evaluation of bottom-head weld indications in a BWR reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Brickstad, B [Swedish Plant Inspectorate, Stockholm (Sweden)

1988-12-31

This document deals with the Non Destructive Examination (NDE) and the fracture mechanics evaluation of bottom head welds in a BWR. The NDE equipment is presented, together with the geometry of evaluated flaw regions. After the fracture mechanics evaluation, it appeared that the plant results fulfilled the usual conditions, and the plant was allowed to operate one more year. (TEC).

The prediction of failure of welded steel plates for pressure vessels

International Nuclear Information System (INIS)

Gulvin, T.F.; Terry, P.; Webster, S.E.

1980-01-01

The avoidance of brittle fracture in pressure vessels and other welded steel fabrications is a matter of considerable importance. The application of fracture mechanics to this problem has led to the evolution of a philosophy which, among other things, permits estimation of tolerable crack sizes so that practical working limits can be set on inspection procedures and on material and welded joint toughness levels. The use of small-scale fracture mechanics tests, particularly crack opening displacement (COD) tests, to make the necessary material and/or joint property assessments is now commonplace. A number of possible analytical techniques have been considered. Initially, the COD data have been used with the Burdekin-Dawes design curve approach taking into account all the possible stress conditions and, it can be demonstrated, that consistently safe predictions of allowable flaw sizes can be made. Also considered, is the fracture analysis diagram approach of Harrison, Loosemore, and Milne in which a combination of fracture toughness data and mechanical property data is used to assess the probability of failure. Similarly the approach defined by Irvine and Quirk which makes use of mechanical property data without resorting to fracture mechanics and finally, a simple approach using uniaxial tensile property data only has also been examined. Comparisons have been made between these four approaches using, initially, the body of data referring to fracture tests on a single material with various defect sizes and aspect ratios, etc. and latterly, to the specific cases in the body of data on high strength steels. The results are discussed. (author)

Behavior of underclad cracks in reactor pressure vessels - evaluation of mechanical analyses with tests on cladded mock-ups

International Nuclear Information System (INIS)

Moinereau, D.; Rousselier, G.; Bethmont, M.

1993-01-01

Innocuity of underclad flaws in the reactor pressure vessels must be demonstrated in the French safety analyses, particularly in the case of a severe transient at the end of the pressure vessel lifetime, because of the radiation embrittlement of the vessel material. Safety analyses are usually performed with elastic and elasto-plastic analyses taking into account the effect of the stainless steel cladding. EDF has started a program including experiments on large size cladded specimens and their interpretations. The purpose of this program is to evaluate the different methods of fracture analysis used in safety studies. Several specimens made of ferritic steel A508 C1 3 with stainless steel cladding, containing small artificial defects, are loaded in four-point bending. Experiments are performed at very low temperature to simulate radiation embrittlement and to obtain crack instability by cleavage fracture. Three tests have been performed on mock-ups containing a small underclad crack (with depth about 5 mn) and a fourth test has been performed on one mock-up with a larger crack (depth about 13 mn). In each case, crack instability occurred by cleavage fracture in the base metal, without crack arrest, at a temperature of about - 170 deg C. Each test is interpreted using linear elastic analysis and elastic-plastic analysis by two-dimensional finite element computations. The fracture are conservatively predicted: the stress intensity factors deduced from the computations (K cp or K j ) are always greater than the base metal toughness. The comparison between the elastic analyses (including two plasticity corrections) and the elastic-plastic analyses shows that the elastic analyses are often conservative. The beneficial effect of the cladding in the analyses is also shown : the analyses are too conservative if the cladding effects is not taken into account. (authors). 9 figs., 6 tabs., 10 refs

Neutron irradiation embrittlement of reactor pressure vessel steel 20 MnMoNi55 weld

International Nuclear Information System (INIS)

Ghoneim, M.M.

1987-05-01

The effect of neutron irradiation on the mechanical and fracture properties of an 'improved' 20 MnMoNi 55 Pressure Vessel Steel (PVS) weld was investigated. In addition to very low residual element content, especially Cu (0.035 wt.%), and relatively higher Ni content (0.9 wt.%), this steel has higher strength (30% more) than the steels used currently in nuclear reactor pressure vessels. The material was irradiated to 3.5x10 19 and 7x10 19 n/cm 2 (E > 1 Mev) at 290 0 C and 2.5x10 19 n/cm 2 (E > 1 MeV) at 160 0 C in FRJ-1 and FRJ-2 research reactors at KFA, Juelich, F.R.G. Test methods used in the evaluation included instrumented impact testing of standard and precracked Charpy specimens, tensile, and fracture toughness testing. Instrumented impact testing provided load and energy vs. time (deflection) data in addition to energy absorption data. The results indicated that the investigated high strength improved steel is more resistant to irradiation induced embrittlement than conventional PVSs. (orig./IHOE)

Use of fracture mechanics in engineering problems

Energy Technology Data Exchange (ETDEWEB)

Carter, C S

1965-02-26

If an engineering material containing a crack is subjected to a slowly increasing load, applied so that the crack tends to open, a small zone of plastic yielding develops at the crack tip. This zone increases in size with increasing load, and has the effect of resisting the tendency of the crack to extend. The basic concepts of fracture mechanics are outlined and the significance of crack toughness as measured by KDcU and KD1cU which relate the applied stress and crack size for unstable fracture prior to general yielding is discussed. The methods available for crack-toughness evaluation are indicated, and the mathematical expressions describing KDcU and KD1cU for a variety of geometrical situations are quoted. This approach to the design of fracture- resistant structures has been used in a number of fields in the U.S. and could be of value to the British steam turbine, aerospace, and pressure-vessel industries for design, inspection, and material selection. (64 refs.)

Proximal Tibial Epiphysis Fracture in a 13-Year-Old Male Athlete

Directory of Open Access Journals (Sweden)

Ioannis M. Stavrakakis

2017-01-01

Full Text Available Fractures of the proximal epiphysis of the tibia are rare, representing 0.5 to 3.0% of all epiphyseal injuries. These injuries can damage the popliteal vessels and their bifurcation, affecting the blood supply of the lower limb, as well as the nerves below the knee. Epiphyseal growth arrest is also a potential complication, leading to various angular deformities. We present a case of a 13-year-old male athlete with a posteriorly displaced Salter-Harris type II fracture of the proximal epiphysis of the left tibia who was treated conservatively with closed reduction and cast immobilization.

Effect of Stress State on Fracture Features

Science.gov (United States)

Das, Arpan

2018-02-01

Present article comprehensively explores the influence of specimen thickness on the quantitative estimates of different ductile fractographic features in two dimensions, correlating tensile properties of a reactor pressure vessel steel tested under ambient temperature where the initial crystallographic texture, inclusion content, and their distribution are kept unaltered. It has been investigated that the changes in tensile fracture morphology of these steels are directly attributable to the resulting stress-state history under tension for given specimen dimensions.

TMI-2 Vessel Investigation Project integration report

International Nuclear Information System (INIS)

Wolf, J.R.; Rempe, J.L.; Stickler, L.A.; Korth, G.E.; Diercks, D.R.; Neimark, L.A.; Akers, D.W.; Schuetz, B.K.; Shearer, T.L.; Chavez, S.A.; Thinnes, G.L.; Witt, R.J.; Corradini, M.L.; Kos, J.A.

1994-03-01

The Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) was an international effort that was sponsored by the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. The primary objectives of the VIP were to extract and examine samples from the lower head and to evaluate the potential modes of failure and the margin of structural integrity that remained in the TMI-2 reactor vessel during the accident. This report presents a summary of the major findings and conclusions that were developed from research during the VIP. Results from the various elements of the project are integrated to form a cohesive understanding of the vessel's condition after the accident

Blood Vessels in Allotransplantation.

Science.gov (United States)

Abrahimi, P; Liu, R; Pober, J S

2015-07-01

Human vascularized allografts are perfused through blood vessels composed of cells (endothelium, pericytes, and smooth muscle cells) that remain largely of graft origin and are thus subject to host alloimmune responses. Graft vessels must be healthy to maintain homeostatic functions including control of perfusion, maintenance of permselectivity, prevention of thrombosis, and participation in immune surveillance. Vascular cell injury can cause dysfunction that interferes with these processes. Graft vascular cells can be activated by mediators of innate and adaptive immunity to participate in graft inflammation contributing to both ischemia/reperfusion injury and allograft rejection. Different forms of rejection may affect graft vessels in different ways, ranging from thrombosis and neutrophilic inflammation in hyperacute rejection, to endothelialitis/intimal arteritis and fibrinoid necrosis in acute cell-mediated or antibody-mediated rejection, respectively, and to diffuse luminal stenosis in chronic rejection. While some current therapies targeting the host immune system do affect graft vascular cells, direct targeting of the graft vasculature may create new opportunities for preventing allograft injury and loss. © Copyright 2015 The American Society of Transplantation and the American Society of Transplant Surgeons.

Femoral head vitality after intracapsular hip fracture

International Nuclear Information System (INIS)

Stroemqvist, B.

1983-01-01

Femoral head vitality before, during and at various intervals from the operation was determined by tetracycline labeling and/or 99 sp (m)Tc-MDP scintimetry. In a three-year follow-up, healing prognosis could be determined by scintimetry 3 weeks from operation; deficient femoral head vitality predicting healing complications and retained vitality predicting uncomplicated healing. A comparison between pre- and postoperative scintimetry indicated that further impairment of the femoral head vitality could be caused by the operative procedure, and as tetracycline labeling prior to and after fracture reduction in 370 fractures proved equivalent, it was concluded that the procedure of osteosynthesis probably was responsible for capsular vessel injury, using a four-flanged nail. The four-flanged nail was compared with a low-traumatic method of osteosynthesis, two hook-pins, in a prospective randomized 14 month study, and the postoperative femoral head vitality was significantly better in the hook-pin group. This was also clearly demonstrated in a one-year follow-up for the fractures included in the study. Parallel to these investigations, the reliability of the methods of vitality determination was found satisfactory in methodologic studies. For clinical purpose, primary atraumatic osteosynthesis, postoperative prognostic scintimetry and early secondary arthroplasty when indicated, was concluded to be the appropriate approach to femoral neck fracture treatment. (Author)

Evidence-based medicine: Mandible fractures.

Science.gov (United States)

Morrow, Brad T; Samson, Thomas D; Schubert, Warren; Mackay, Donald R

2014-12-01

After studying this article, the participant should be able to: 1. Describe the anatomy and subunits of the mandible. 2. Review the cause and epidemiology of mandible fractures. 3. Discuss the preoperative evaluation and diagnostic imaging. 4. Understand the principles and techniques of mandible fracture reduction and fixation. The management of mandibular fractures has undergone significant improvement because of advancements in plating technology, imaging, and instrumentation. As the techniques in management continue to evolve, it is imperative for the practicing physician to remain up-to-date with the growing body of scientific literature. The objective of this Maintenance of Certification article is to present a review of the literature so that the physician may make treatment recommendation based on the best evidence available. Pediatric fractures have been excluded from this article.

The statistical background to proposed ASME/MPC fracture toughness reference curves

International Nuclear Information System (INIS)

Oldfield, W.

1981-01-01

The ASME Pressure Vessel Codes define, in Sec. 11, lower bound fracture toughness curves. These curves are used to predict the lower bound fracture toughness on the basis of the RT test procedure. This test is used to remove heat to heat differences, by permitting the lower bound (reference) curve to be moved along the temperature scale according to the measured RT. Numerous objections have been raised to the procedure, and a Subcommittee (the ASME/MPC Working Group on Reference Toughness) is currently revising the codified procedures for fracture toughness prediction. The task has required a substantial amount of statistical work, since the new procedure are to have a statistical basis. Using initiation fracture toughness (J-Integral R curve procedures in the ductile domain) it was shown that when CVN energy data is properly transformed it is highly correlated with valid fracture toughness measurements. A single functional relationship can be used to predict the mean fracture toughness for a sample of steel from a set of CVN energy measurements, and the coefficients of the function tabulated. More importantly, the approximate lower statistical bounds to the initiation fracture toughness behaviour can be similarly predicted, and coefficients for selected bounds have also been tabulated. (orig.)

A Comprehensive Fracture Prevention Strategy in Older Adults

DEFF Research Database (Denmark)

Blain, H; Masud, T; Dargent-Molina, P

2016-01-01

Prevention of fragility fractures in older people has become a public health priority, although the most appropriate and cost-effective strategy remains unclear. In the present statement, the Interest Group on Falls and Fracture Prevention of the European Union Geriatric Medicine Society (EUGMS...

Fracture Characterization in Reactive Fluid-Fractured Rock Systems Using Tracer Transport Data

Science.gov (United States)

Mukhopadhyay, S.

2014-12-01

Fractures, whether natural or engineered, exert significant controls over resource exploitation from contemporary energy sources including enhanced geothermal systems and unconventional oil and gas reserves. Consequently, fracture characterization, i.e., estimating the permeability, connectivity, and spacing of the fractures is of critical importance for determining the viability of any energy recovery program. While some progress has recently been made towards estimating these critical fracture parameters, significant uncertainties still remain. A review of tracer technology, which has a long history in fracture characterization, reveals that uncertainties exist in the estimated parameters not only because of paucity of scale-specific data but also because of knowledge gaps in the interpretation methods, particularly in interpretation of tracer data in reactive fluid-rock systems. We have recently demonstrated that the transient tracer evolution signatures in reactive fluid-rock systems are significantly different from those in non-reactive systems (Mukhopadhyay et al., 2013, 2014). For example, the tracer breakthrough curves in reactive fluid-fractured rock systems are expected to exhibit a long pseudo-state condition, during which tracer concentration does not change by any appreciable amount with passage of time. Such a pseudo-steady state condition is not observed in a non-reactive system. In this paper, we show that the presence of this pseudo-steady state condition in tracer breakthrough patterns in reactive fluid-rock systems can have important connotations for fracture characterization. We show that the time of onset of the pseudo-steady state condition and the value of tracer concentration in the pseudo-state condition can be used to reliably estimate fracture spacing and fracture-matrix interface areas.

Stress Fractures of the Foot.

Science.gov (United States)

Hossain, Munier; Clutton, Juliet; Ridgewell, Mark; Lyons, Kathleen; Perera, Anthony

2015-10-01

Stress fractures of the foot and ankle may be more common among athletes than previously reported. A low threshold for investigation is warranted and further imaging may be appropriate if initial radiographs remain inconclusive. Most of these fractures can be treated conservatively with a period of non-weight-bearing mobilization followed by gradual return to activity. Early surgery augmented by bone graft may allow athletes to return to sports earlier. Risk of delayed union, nonunion, and recurrent fracture is high. Many of the patients may also have risk factors for injury that should be modified for a successful outcome. Copyright © 2015 Elsevier Inc. All rights reserved.

Contemporary management of subtrochanteric fractures.

Science.gov (United States)

Joglekar, Siddharth B; Lindvall, Eric M; Martirosian, Armen

2015-01-01

Cephalomedullary interlocking nails that allow for trochanteric entry and minimally invasive fixation have revolutionized the contemporary management of subtrochanteric fractures with improved union rates and decreased incidence of fixation failure. The most successful alternative to intramedullary fixation remains the angled blade plate. Despite biomechanical superiority of contemporary intramedullary implants to previous intramedullary devices, the importance of achieving and maintaining satisfactory fracture reduction prior to and during hardware insertion cannot be overemphasized. In comminuted and more challenging fractures, additional techniques, such as limited open reduction with clamps and/or cables, can allow for canal restoration and more anatomic reductions prior to and/or during nail insertion. Published by Elsevier Inc.

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.

Experience with the WWER-440 MW reactor pressure vessel in-service inspections and evaluation of their results

International Nuclear Information System (INIS)

Brumovsky, M.; Kralovec, J.; Prepechal, J.; Sulc, J.

1989-01-01

The Power Machinery Plant of Skoda Works in Plzen carries out in-service inspections of WWER-440 MW reactor pressure vessels by means of remote controlled inspection equipment - the TRC reactor test system, and some other inspections devices. The results of the in-service inspections were evaluated by methods based on the fracture mechanics approach, the knowledge of stress and strain distribution, and the operating history of the pressure vessels. Examples of types of defects found and their analysis are shown. (author). 1 tab

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

Preliminary assessment of the effects of biaxial loading on reactor pressure vessel structural-integrity-assessment technology

International Nuclear Information System (INIS)

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

1996-01-01

Effects of biaxial loading on shallow-flaw fracture toughness were studied to determine potential impact on structural integrity assessment of a reactor pressure vessel (RPV) under pressurized thermal shock (PTS) transient loading and pressure-temperature (PT) loading produced by reactor heatup and cooldown transients. Biaxial shallow-flaw fracture-toughness tests results were also used to determine the parameter controlling fracture in the transition temperature range, and to develop a related dual-parameter fracture-toughness correlation. Shallow-flaw and biaxial loading effects were found to reduce the conditional probability of crack initiation by a factor of nine when the shallow-flaw fracture-toughness K Jc data set, with biaxial-loading effects adjustments, was substituted in place of ASME Code K Ic data set in PTS analyses. Biaxial loading was found to reduce the shallow-flaw fracture toughness of RPV steel such that the lower-bound curve was located between ASME K Ic and K IR curves. This is relevant to future development of P-T curve analysis procedures. Fracture in shallow-flaw biaxial samples tested 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

Inservice inspection of Halden BWR pressure vessel

International Nuclear Information System (INIS)

Foerli, O.; Hernes, T.

1978-01-01

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

Fracture size and transmissivity correlations: Implications for transport simulations in sparse three-dimensional discrete fracture networks following a truncated power law distribution of fracture size

Science.gov (United States)

Hyman, J.; Aldrich, G. A.; Viswanathan, H. S.; Makedonska, N.; Karra, S.

2016-12-01

We characterize how different fracture size-transmissivity relationships influence flow and transport simulations through sparse three-dimensional discrete fracture networks. Although it is generally accepted that there is a positive correlation between a fracture's size and its transmissivity/aperture, the functional form of that relationship remains a matter of debate. Relationships that assume perfect correlation, semi-correlation, and non-correlation between the two have been proposed. To study the impact that adopting one of these relationships has on transport properties, we generate multiple sparse fracture networks composed of circular fractures whose radii follow a truncated power law distribution. The distribution of transmissivities are selected so that the mean transmissivity of the fracture networks are the same and the distributions of aperture and transmissivity in models that include a stochastic term are also the same.We observe that adopting a correlation between a fracture size and its transmissivity leads to earlier breakthrough times and higher effective permeability when compared to networks where no correlation is used. While fracture network geometry plays the principal role in determining where transport occurs within the network, the relationship between size and transmissivity controls the flow speed. These observations indicate DFN modelers should be aware that breakthrough times and effective permeabilities can be strongly influenced by such a relationship in addition to fracture and network statistics.

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

International Nuclear Information System (INIS)

Server, W.L.

1984-01-01

Radiation embrittlement of ferritic pressure vessel steels increases the ductile-brittle transition temperature and decreases the upper-shelf level of toughness as measured by Charpy impact tests. A thermal anneal cycle well above the normal operating temperature of the vessel can restore most of the original Charpy V-notch energy properties. A test reactor pressure vessel has been wet annealed at less than 343 0 C (650 0 F), and annealing of the Belgian BR-3 reactor vessel has recently taken place. An industry survey indicates that dry annealing a reactor vessel in-place is feasible, but solvable engineering problems do exist. The materials with highest radiation sensitivity in the older reactor vessels are submerged-arc weld metals with high copper and nickel concentrations. The limited Charpy V-notch and fracture toughness data available for five such welds were reviewed. The review suggested that significant recovery results from annealing at 454 0 C (850 0 F) for one week. Two of the main concerns with a localized heat treatment at 454 0 C (850 0 F) are the degree of distortion that may occur after the annealing cycle and the extent of residual stresses. A thermal and structural analysis of a reactor vessel for distortions and residual stresses found no problems with the reactor vessel itself but did indicate a rotation at the nozzle region of the vessel that would plastically deform the attached primary piping. Further analytical studies are needed. An American Society for Testing and Materials (ASTM) task group is upgrading and revising the ASTM Recommended Guide for In-Service Annealing of WaterCooled Nuclear Reactor Vessels (E 509-74) with emphasis on the materials and surveillance aspects of annealing rather than system engineering problems. System safety issues are the province of organizations other than ASTM (for example, the American Society of Mechanical Engineers Boiler and Pressure Vessel Code body)

Study of brittle fracture in thick walled structures using small and large specimens

International Nuclear Information System (INIS)

Norrthon, J.-O.; Carlsson, J.

1979-01-01

A Russian pressure vessel steel has been investigated for fracture toughness data, Ksub(Ic) and Jsub(Ic). Four large specimens have been tested and the results correlated to data from several smaller specimens. Onset of crack growth has been detected by a high frequency electric method. (author)

A comprehensive fracture prevention strategy in older adults

DEFF Research Database (Denmark)

Blain, H.; Masud, T.; Dargent-Molina, P.

2016-01-01

Prevention of fragility fractures in older people has become a public health priority, although the most appropriate and cost-effective strategy remains unclear. In the present statement, the Interest group on falls and fracture prevention of the European union geriatric medicine society (EUGMS...... of osteoporosis and osteoarthritis (ECCEO), outlines its views on the main points in the current debate in relation to the primary and secondary prevention of falls, the diagnosis and treatment of bone fragility, and the place of combined falls and fracture liaison services for fracture prevention in older people....

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.)

Numerical modeling of ductile tearing effects on cleavage fracture toughness

International Nuclear Information System (INIS)

Dodds, R.H. Jr.; Tang, M.; Anderson, T.L.

1994-05-01

Experimental studies demonstrate a significant effect of specimen size, a/W ratio and prior ductile tearing on cleavage fracture toughness values (J c ) measured in the ductile-to-brittle transition region of ferritic materials. In the lower-transition region, cleavage fracture often occurs under conditions of large-scale yielding but without prior ductile crack extension. The increased toughness develops when plastic zones formed at the crack tip interact with nearby specimen surfaces which relaxes crack-tip constraint (stress triaxiality). In the mid-to-upper transition region, small amounts of ductile crack extension (often c -values. Previous work by the authors described a micromechanics fracture model to correct measured J c -values for the mechanistic effects of large-scale yielding. This new work extends the model to also include the influence of ductile crack extension prior to cleavage. The paper explores development of the new model, provides necessary graphs and procedures for its application and demonstrates the effects of the model on fracture data sets for two pressure vessel steels (A533B and A515)

Validation of favor code linear elastic fracture solutions for finite-length flaw geometries

International Nuclear Information System (INIS)

Dickson, T.L.; Keeney, J.A.; Bryson, J.W.

1995-01-01

One of the current tasks within the US Nuclear Regulatory Commission (NRC)-funded Heavy Section Steel Technology Program (HSST) at Oak Ridge National Laboratory (ORNL) is the continuing development of the FAVOR (Fracture, analysis of Vessels: Oak Ridge) computer code. FAVOR performs structural integrity analyses of embrittled nuclear reactor pressure vessels (RPVs) with stainless steel cladding, to evaluate compliance with the applicable regulatory criteria. Since the initial release of FAVOR, the HSST program has continued to enhance the capabilities of the FAVOR code. ABAQUS, a nuclear quality assurance certified (NQA-1) general multidimensional finite element code with fracture mechanics capabilities, was used to generate a database of stress-intensity-factor influence coefficients (SIFICs) for a range of axially and circumferentially oriented semielliptical inner-surface flaw geometries applicable to RPVs with an internal radius (Ri) to wall thickness (w) ratio of 10. This database of SIRCs has been incorporated into a development version of FAVOR, providing it with the capability to perform deterministic and probabilistic fracture analyses of RPVs subjected to transients, such as pressurized thermal shock (PTS), for various flaw geometries. This paper discusses the SIFIC database, comparisons with other investigators, and some of the benchmark verification problem specifications and solutions

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.

Circumferential wires as a supplement to intramedullary nailing in unstable trochanteric hip fractures

DEFF Research Database (Denmark)

Ban, Ilija; Birkelund, Lasse; Palm, Henrik

2012-01-01

Fixation of unstable trochanteric fractures is challenging. Application of a circumferential wire may facilitate bone contact and avoid postoperative fracture displacement. However, the use of circumferential wires remains controversial due to possible disturbance of the blood supply to the under......Fixation of unstable trochanteric fractures is challenging. Application of a circumferential wire may facilitate bone contact and avoid postoperative fracture displacement. However, the use of circumferential wires remains controversial due to possible disturbance of the blood supply...... to the underlying bone. We evaluated the results of applied circumferential wires, concentrating mainly on complications and reoperations....

Review of the TMI-2 accident evaluation and vessel investigation projects

Energy Technology Data Exchange (ETDEWEB)

Ladekarl Thomsen, Knud

1998-03-01

The results of the TMI-2 Accident Evaluation Programme and the Vessel Investigation Project have been reviewed as part of a literature study on core meltdown and in-vessel coolability. The emphasis is placed on the late phase melt progression, which is of special relevance to the NKS-sponsored RAK-2.1 project on Severe Accident Phenomenology. The body of the report comprises three main sections, The TMI-2 Accident Scenario, Core Region and Relocation Path Investigations, and Lower Head Investigations. In the final discussion, the lower head gap formation mechanism is explained in terms of thermal contraction and fracturing of the debris crust. This model seems more plausible than the MAAP model based on creep expansion of the lower head. (au) 1 tab., 33 ills., 31 refs.

[Operative treatment of sacroiliac joint fracture and dislocation in Tile C pelvic fracture with Colorado 2 system].

Science.gov (United States)

Liu, Shuping; Zhou, Qing; Liu, Yuehong; Chen, Xi; Zhou, Yu; Zhang, Desheng; Fang, Zhi; Xu, Wei

2011-12-01

To explore the effectiveness of Colorado 2 system in the stability reconstruction of sacroiliac joint fracture and dislocation in Tile C pelvic fracture. Between February 2009 and January 2011, 8 cases of Tile C pelvic fracture were treated with Colorado 2 system. There were 3 males and 5 females with an average age of 34.4 years (range, 22-52 years). Fractures were caused by traffic accident in 3 cases, by falling from height in 3 cases, and by crash of heavy object in 2 cases. According to Tile classification, 5 cases were classified as C1-2, 2 cases as C1-3, and 1 case as C2. The time between injury and operation was 5-10 days (mean, 7 days). After skeletal traction reduction, Colorado 2 system was used to fix sacroiliac joint, and reconstruction plate or external fixation was selectively adopted. The postoperative X-ray films showed that the reduction of vertical and rotatory dislocation was satisfactory, posterior pelvic ring achieved effective stability. All the incisions healed by first intention, and no blood vessel or nerve injury occurred. Eight patients were followed up 6-24 months (mean, 12 months). No loosening or breakage of internal fixation was observed and no re-dislocation of sacroiliac joint occurred. The bone healing time was 6-12 months (mean, 9 months). According to Majeed's functional criterion, the results were excellent in 5 cases, good in 2 cases, and fair in 1 case at last follow-up. Colorado 2 system could provide immediate stability of pelvic posterior ring and good maintenance of reduction effect, which is an effective method in the therapy of sacroiliac joint fracture and dislocation in Tile C pelvic fracture.

J estimation scheme for cracks near the cladding of a reactor pressure vessel

International Nuclear Information System (INIS)

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

1992-01-01

The evaluation of flaws near the cladding is an important issue in term of risk of fast fracture of main vessel. This study analyses different K estimation schemes. These different K values are compared with respect to the toughness of the material K IC for different crack situations; the results confirm the validity of the proposal in the French RCC M Code for the plastic zone correction

An Improved Rate-Transient Analysis Model of Multi-Fractured Horizontal Wells with Non-Uniform Hydraulic Fracture Properties

Directory of Open Access Journals (Sweden)

Youwei He

2018-02-01

Full Text Available Although technical advances in hydraulically fracturing and drilling enable commercial production from tight reservoirs, oil/gas recovery remains at a low level. Due to the technical and economic limitations of well-testing operations in tight reservoirs, rate-transient analysis (RTA has become a more attractive option. However, current RTA models hardly consider the effect of the non-uniform production on rate decline behaviors. In fact, PLT results demonstrate that production profile is non-uniform. To fill this gap, this paper presents an improved RTA model of multi-fractured horizontal wells (MFHWs to investigate the effects of non-uniform properties of hydraulic fractures (production of fractures, fracture half-length, number of fractures, fracture conductivity, and vertical permeability on rate transient behaviors through the diagnostic type curves. Results indicate obvious differences on the rate decline curves among the type curves of uniform properties of fractures (UPF and non-uniform properties of fractures (NPF. The use of dimensionless production integral derivative curve magnifies the differences so that we can diagnose the phenomenon of non-uniform production. Therefore, it’s significant to incorporate the effects of NPF into the RDA models of MFHWs, and the model proposed in this paper enables us to better evaluate well performance based on long-term production data.

Fracture in Duchenne Muscular Dystrophy: Natural History and Vitamin D Deficiency.

Science.gov (United States)

Perera, Nadia; Sampaio, Hugo; Woodhead, Helen; Farrar, Michelle

2016-08-01

The present study examined the natural history of fracture and vitamin D levels in Duchenne muscular dystrophy patients, who are vulnerable to osteoporosis and fractures. Retrospective analysis of a cohort of 48 Duchenne muscular dystrophy patients revealed that 43% of patients experienced ≥1 fracture. Fracture probabilities at ages 6, 9, 12, and 15 years were 4%, 9%, 31%, and 60% respectively, accelerating around the time of ambulation loss (mean age 11.8 ± 2.7 years). Chronic corticosteroid therapy was utilized in 69% of patients and was associated with all vertebral fractures. A history of vitamin D deficiency occurred in 84%, and 35% were currently deficient. Despite chronic vitamin D supplementation, 38% remained deficient. These results demonstrate that osteoporosis and fracture remain major concerns in Duchenne muscular dystrophy. Bone health should be optimized well before loss of ambulation, however current levels of vitamin D supplementation may be inadequate given high levels of deficiency. © The Author(s) 2016.

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

Calculation of neutron fluence in the region of the pressure vessel for the history of different reactors by using the Monte-Carlo-method

International Nuclear Information System (INIS)

Barz, H.U.; Bertram, W.

1992-01-01

Embrittlement of pressure vessel material caused by neutron irradiation is a very important problem for VVER-440 reactors. For the estimation of the fracture risk highly reliable neutron fluence values are necessary. For this reason a special theoretical determination of space dependent neutron fluences has been performed mainly on the basis of Monte-Carlo calculations. The described method allows the accurate calculation of neutron fluences near the pressure vessel in the height of the core region for all reactor histories and loading cycles in an efficient manner. To illustrate the accuracy of the suggested method a comparison with experimental results was done. The calculated neutron fluence values can be used for planning the loading schemes of each reactor according to the safety requirements against brittle fracture. (orig.)

Irradiation effects in low-alloy reactor pressure vessel steels (Heavy-Section Steel Technology program series 4 and 5)

International Nuclear Information System (INIS)

McGowan, J.J.; Nanstad, R.K.; Thoms, K.R.; Menke, B.H.

1985-01-01

This report presents studies on the irradiation effects in low-alloy reactor pressure vessel steels. The Fourth Heavy-Section Steel Technology (HSST) Irradiation Series, almost completed, was aimed at elastic-plastic and fully plastic fracture toughness of low-copper weldments (''current practice welds''). A typical nuclear pressure vessel plate steel was included for statistical purposes. The Fifth HSST Irradiation Series, now in progress, is aimed at determining the shape of the K/sub IR/ curve after significant radiation-induced shift of the transition temperatures. This series includes irradiated test specimens of thicknesses up to 100 mm and weldment compositions typical of early nuclear power reactor pressure vessel welds. 27 refs., 22 figs

Analysis of PITFL injuries in rotationally unstable ankle fractures.

Science.gov (United States)

Warner, Stephen J; Garner, Matthew R; Schottel, Patrick C; Hinds, Richard M; Loftus, Michael L; Lorich, Dean G

2015-04-01

Reduction and stabilization of the syndesmosis in unstable ankle fractures is important for ankle mortise congruity and restoration of normal tibiotalar contact forces. Of the syndesmotic ligaments, the posterior inferior tibiofibular ligament (PITFL) provides the most strength for maintaining syndesmotic stability, and previous work has demonstrated the significance of restoring PITFL function when it remains attached to a posterior malleolus fracture fragment. However, little is known regarding the nature of a PITFL injury in the absence of a posterior malleolus fracture. The goal of this study was to describe the PITFL injury pattern based on magnetic resonance imaging (MRI) and intraoperative observation. A prospective database of all operatively treated ankle fractures by a single surgeon was used to identify all supination-external rotation (SER) types III and IV ankle fracture patients with complete preoperative orthogonal ankle radiographs and MRI. All patients with a posterior malleolus fracture were excluded. Using a combination of preoperative imaging and intraoperative findings, we analyzed the nature of injuries to the PITFL. In total, 185 SER III and IV operatively treated ankle fractures with complete imaging were initially identified. Analysis of the preoperative imaging and operative reports revealed 34% (63/185) had a posterior malleolus fracture and were excluded. From the remaining 122 ankle fractures, the PITFL was delaminated from the posterior malleolus in 97% (119/122) of cases. A smaller proportion (3%; 3/122) had an intrasubstance PITFL rupture. Accurate and stable syndesmotic reduction is a significant component of restoring the ankle mortise after unstable ankle fractures. In our large cohort of rotationally unstable ankle fractures without posterior malleolus fractures, we found that most PITFL injuries occur as a delamination off the posterior malleolus. This predictable PITFL injury pattern may be used to guide new methods for

Comparative Evaluation of Enalapril and Losartan in Pharmacological Correction of Experimental Osteoporosis and Fractures of Its Background

Science.gov (United States)

Rajkumar, D. S. R.; Faitelson, A. V.; Gudyrev, O. S.; Dubrovin, G. M.; Pokrovski, M. V.; Ivanov, A. V.

2013-01-01

In the experiment on the white Wistar female rats (222 animals), the osteoprotective effect of enalapril and losartan was studied on experimental models of osteoporosis and osteoporotic fractures. It was revealed that in rats after ovariectomy, the endothelial dysfunction of microcirculation vessels of osteal tissue develops, resulting in occurrence of osteoporosis and delay of consolidation of experimental fractures. Enalapril and losartan prevented the reduction of microcirculation in bone, which was reflected in slowing the thinning of bone trabeculae and in preventing the occurrence of these microfractures, as well as increasing quality of experimental fractures healing. PMID:23401845

Comparative Evaluation of Enalapril and Losartan in Pharmacological Correction of Experimental Osteoporosis and Fractures of Its Background

Directory of Open Access Journals (Sweden)

D. S. R. Rajkumar

2013-01-01

Full Text Available In the experiment on the white Wistar female rats (222 animals, the osteoprotective effect of enalapril and losartan was studied on experimental models of osteoporosis and osteoporotic fractures. It was revealed that in rats after ovariectomy, the endothelial dysfunction of microcirculation vessels of osteal tissue develops, resulting in occurrence of osteoporosis and delay of consolidation of experimental fractures. Enalapril and losartan prevented the reduction of microcirculation in bone, which was reflected in slowing the thinning of bone trabeculae and in preventing the occurrence of these microfractures, as well as increasing quality of experimental fractures healing.

Crack blunting, cleavage fracture in transition area and stable crack growth - investigated using the nonlinear fracture mechanics method

International Nuclear Information System (INIS)

Heerens, J.

1990-01-01

A procedure is developed which allows to estimate crack tip blunting using the stress-strain curve of the material and the J-integral. The second part deals with cleavage fracture in a quenched and tempered pressure vessel steel. It was found that within the ductile to brittle transition regime the fracture toughness is controlled by cleavage initiated at 'weak spots of the material' and by the normal stresses at the weak spots. In the last part of the paper the influence of specimen size on J-, Jm- and δ 5 -R-curves for side grooved CT-specimens under fully plastic condition is investigated. In order to characterize constraint-effects the necking of the specimens was measured. For specimens having similar constraint the parameters Jm and δ 5 yielded size independent R-curves over substantial larger amounts of crack extension than the J-integral. (orig.) With 114 figs., 10 tabs [de

Potential effect of fracture technology on IPTS [Integrated Pressurized Thermal Shock] analysis (Fracture toughness: Kla and Klc and warm prestressing)

International Nuclear Information System (INIS)

Dickson, T.L.

1990-01-01

A major nuclear plant life extension issue to be confronted in the 1990's is pressure vessel integrity for the pressurized thermal shock (PTS) loading condition. Governing criteria associated with PTS are included in ''The PTS Rule'' (10 CFR 50.61) and Regulatory Guide 1.154: Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors. The results of the Integrated Pressurized Water Reactors. The results of the Integrated Pressurized Thermal Shock (IPTS) Program, along with risk assessments and fracture analyses performed by the NRC and reactor system vendors, contributed to the derivation of the PTS Rule. Over the last several years, the Heavy Section Steel Technology (HSST) Program at the Oak Ridge National Laboratory (ORNL) has performed a series of large-scale fracture-mechanics experiments. The Thermal Shock Experiments (TSE), Pressurized Thermal Shock Experiments (PTSE), and Wide Plate Experiments (WPE) produced K IC and K Ia data that suggest increased mean K IC and K Ia curves relative to the ones used in the IPTS study. Also, the PTSE and WPE have demonstrated that prototypical nuclear reactor pressure vessel steels are capable of arresting a propagating crack at K I values considerably above 220 MPa√m, the implicit limit of the ASME Code and the limit used in the IPTS studies. This document provides a discussion of the results of these experiments

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

Acoustic emission measurements at the pressure vessel ZB2

International Nuclear Information System (INIS)

Tirbonod, B.; Hanacek, L.

1990-01-01

The work presented here is the Swiss contribution to the project 'Zwischenbehaelter 2 (ZB2)' hosted by the 'Bundesministerium fuer Forschung und Technologie' of the Federal Republic of Germany. One of the crack-like defects introduced at the inside surface of the thick-walled pressure vessel ZB2 was locally monitored by acoustic emission. The measurement system was broadband (0.5 - 5 MHz) and allowed a threedimensional location of the source. The vessel was subjected to different tests. Signals were recorded during the second series of hydrotests, fast pressure cycles and fatigue test at 50 C. About 1 signal per hydrotest or cycle was recorded. For the hydrotests the signals were recorded generally at loading in the intermediate range of pressure; the sources were located in the artificial defect. Recurrent and non recurrent signals were recorded during the fatigue test. At loading, signals were captured up to the maximum pressure and for the recurrent signals at well defined pressure ranges. All the sources (except one, located in the base material ahead of the artificial defect) were situated in the artificial defect. The pressure and location depended on the loading phase and on the cycle range. The measurements were discussed by describing the signals by measurement, signal and source parameters. The goal was to identify the source mechanism and to assess the growth of the defect. For the hydrotests the identification of the mechanism at loading remains open. For the fatigue test the source situated in the base material was attributed to a primary mechanism; this source could assess the growth of the defect on the basis of linear elastic fracture mechanics. A secondary mechanism was suggested for recurrent sources active at loading. For all the tests, the sources active at unloading were attributed to a secondary mechanism. (author)

Ex-vessel nuclear fuel transfer system

International Nuclear Information System (INIS)

Wade, E.E.

1978-01-01

A system for transferring fuel assemblies between a fuel transfer area and a fuel storage area while the fuel assemblies remain completely submerged in a continuous body of coolant is described. A fuel transfer area filled with reactor coolant communicating with the reactor vessel below the reactor coolant level provides a transfer area for fuel assemblies in transit to and from the reactor vessel. A positioning mechanism comprising at least one rotatable plug disposed on a fuel transfer tank located outside the reactor vessel cooperates with either the fuel transfer area or the fuel storage area to position a fuel assembly in transit. When in position, a transporting mechanism cooperating with the positioning mechanism lifts or lowers a chosen fuel assembly. The transporting mechanism together with the positioning mechanism are capable of transferring a fuel assembly between the fuel transfer area and the fuel storage area

Prosthetic replacement for proximal humeral fractures.

Science.gov (United States)

Kontakis, George; Tosounidis, Theodoros; Galanakis, Ioannis; Megas, Panagiotis

2008-12-01

The ideal management of complex proximal humeral fractures continues to be debatable. Evolution of proximal humeral fracture management, during the past decade, led to the implementation of many innovations in surgical treatment. Even though the pendulum of treatment seems to swing towards new trends such as locked plating, hemiarthroplasty remains a valid and reliable option that serves the patient's needs well. Hemiarthroplasty is indicated for complex proximal humeral fractures in elderly patients with poor bone stock and when internal fixation is difficult or unreliable. Hemiarthroplasty provides a better result when it is performed early post-injury. Stem height, retroversion and tuberosity positioning are technical aspects of utmost importance. Additionally reverse total shoulder arthroplasty is an alternative new modality that can be used as a primary solution in selected patients with proximal humeral fracture treatment. Failed hemiarthroplasty and fracture sequelae can be successfully managed with reverse total shoulder arthroplasty. Individual decision-making and tailored treatment that takes into consideration the personality of the fracture and the patient's characteristics should be used.

Twenty years of fracture mechanics and flaw evaluation applications in the ASME Nuclear Code

International Nuclear Information System (INIS)

Riccardella, P.C.

1991-01-01

The paper presents a retrospective on the development and applications of fracture mechanics-based toughness requirements and flaw evaluation methodology in Sections III and XI of the ASME Code. Section III developments range from the rules and requirements for thick section Class 1 pressure vessels to thinner section components in other Classes. Section XI applications include flaw acceptance standards and evaluation methodology for various components ranging from pressure vessels to thins section piping of carbon and austenitic steels. The experience gained in operating plant applications of these rules and procedures are also discussed

Local approach of cleavage fracture applied to a vessel with subclad flaw. A benchmark on computational simulation

International Nuclear Information System (INIS)

Moinereau, D.; Brochard, J.; Guichard, D.; Bhandari, S.; Sherry, A.; France, C.

1996-10-01

A benchmark on the computational simulation of a cladded vessel with a 6.2 mm sub-clad flaw submitted to a thermal transient has been conducted. Two-dimensional elastic and elastic-plastic finite element computations of the vessel have been performed by the different partners with respective finite element codes ASTER (EDF), CASTEM 2000 (CEA), SYSTUS (Framatome) and ABAQUS (AEA Technology). Main results have been compared: temperature field in the vessel, crack opening, opening stress at crack tips, stress intensity factor in cladding and base metal, Weibull stress σ w and probability of failure in base metal, void growth rate R/R 0 in cladding. This comparison shows an excellent agreement on main results, in particular on results obtained with local approach. (K.A.)

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

International Nuclear Information System (INIS)

Gomes, Paulo de Tarso Vida

2005-01-01

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

Reactor pressure vessel steels[1997 Scientific Report of the Belgian Nuclear Research Centre

Energy Technology Data Exchange (ETDEWEB)

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

1998-07-01

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.

Pressure vessel steels: influence of chemical composition on irradiation sensitivity

International Nuclear Information System (INIS)

Ghoniem, M.M.; Hammad, F.H.

1998-01-01

Neutron irradiation of the steels used in the construction of the nuclear reactor pressure vessels can lead to the embrittlement of these materials, increasing the ductile-to-brittle transition temperature and decreasing the fracture energy, which can limit the plant life. The knowledge of irradiation embrittlement and the means for minimizing such degradation is therefore important in the field of assuring the safety of the nuclear power plants. Irradiation embrittlement is quite a complex process. It involves many variables. The most important of these are irradiation temperature, neutron fluence (neutron dose), neutron flux (neutron dose rate), and chemical composition of the irradiated material. This paper is concerned with the effect of chemical composition, the role of residual and alloying elements in the irradiation embrittlement of nuclear reactor pressure vessel steels in light water reactors. It presents a critical review for the published work in this field through the last 25 years

Allowable minimum upper shelf toughness for nuclear reactor pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.

1988-05-01

The paper develops methodology and procedure for determining the allowable minimum upper shelf toughness for continued safe operation of nuclear reactor pressure vessels. Elastic-plastic fracture mechanics analysis method based on the J-integral tearing modulus (J/T) approach is used. Closed from expressions for the applied J and tearing modulus are presented for finite length, part-throughwall axial flaw with aspect ratio of 1/6. Solutions are then presented for Section III, Appendix G flaw. A simple flaw evaluation procedure that can be applied quickly by utility engineers is presented. An attractive feature of the simple procedure is that tearing modulus calculations are not required by the user, and a solution for the slope of the applied J/T line is provided. Results for the allowable minimum upper shelf toughness are presented for a range of reactor pressure vessel thickness and heatup/cooldown rates.

Allowable minimum upper shelf toughness for nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Zahoor, A.

1988-01-01

The paper develops methodology and procedure for determining the allowable minimum upper shelf toughness for continued safe operation of nuclear reactor pressure vessels. Elastic-plastic fracture mechanics analysis method based on the J-integral tearing modulus (J/T) approach is used. Closed from expressions for the applied J and tearing modulus are presented for finite length, part-throughwall axial flaw with aspect ratio of 1/6. Solutions are then presented for Section III, Appendix G flaw. A simple flaw evaluation procedure that can be applied quickly by utility engineers is presented. An attractive feature of the simple procedure is that tearing modulus calculations are not required by the user, and a solution for the slope of the applied J/T line is provided. Results for the allowable minimum upper shelf toughness are presented for a range of reactor pressure vessel thickness and heatup/cooldown rates. (orig.)

Femoral neck stress fractures (fnsf) in military recruits

International Nuclear Information System (INIS)

Majeed, N.U.; Naqvi, A.N.; Majeed, H.

2012-01-01

Objective: To identify patterns of Femoral Neck Stress Fractures (FNSF), its presentation and outcome of its treatment in PMA (Pakistan Military Academy) cadets. These findings would help suggest guidelines for their appropriate management. Study design: Case Series Place and duration of study: CMH Abbottabad and CMH Rawalpindi from May 2005 to January 2008. Materials and Methods: Twenty cases (20 hips in 18 patients) of FNSF were included in the study. Only male cadets from Pakistan Military Academy (PMA) were included. Diagnosis was made clinically and was confirmed by radiographs or bone scan. Incomplete fractures were managed conservatively where as complete fractures were fixed surgically. Results: All compression fractures healed conservatively where as tension fractures needed surgical fixation in all the cases, except one where fracture remained incomplete. Rest of tension fractures converted from incomplete fractures to complete fractures and hence needed surgical stabilization. There was no problem of avascular necrosis of femoral head (AVNFH) in any patient. Conclusion: FNSF are uncommon injuries with potentially serious complications and are difficult to diagnose clinically. When diagnosed early and managed appropriately, they carry good prognosis. (author)

Capacity of Prestressed Concrete Containment Vessels with Prestressing Loss

International Nuclear Information System (INIS)

SMITH, JEFFREY A.

2001-01-01

Reduced prestressing and degradation of prestressing tendons in concrete containment vessels were investigated using finite element analysis of a typical prestressed containment vessel. The containment was analyzed during a loss of coolant accident (LOCA) with varying levels of prestress loss and with reduced tendon area. It was found that when selected hoop prestressing tendons were completely removed (as if broken) or when the area of selected hoop tendons was reduced, there was a significant impact on the ultimate capacity of the containment vessel. However, when selected hoop prestressing tendons remained, but with complete loss of prestressing, the predicted ultimate capacity was not significantly affected for this specific loss of coolant accident. Concrete cracking occurred at much lower levels for all cases. For cases where selected vertical tendons were analyzed with reduced prestressing or degradation of the tendons, there also was not a significant impact on the ultimate load carrying capacity for the specific accident analyzed. For other loading scenarios (such as seismic loading) the loss of hoop prestressing with the tendons remaining could be more significant on the ultimate capacity of the containment vessel than found for the accident analyzed. A combination of loss of prestressing and degradation of the vertical tendons could also be more critical during other loading scenarios

Mandibular ramus fractures: a rarity.

Science.gov (United States)

Kale, Tejraj Pundalik; Kotrashetti, S M; Louis, Archana; Lingaraj, J B; Sarvesh, B U

2013-01-01

To determine the incidence of mandibular ramus fractures in KLE's PK Hospital and to analyze the outcome of open reduction and internal fixation of these fractures. Using a retrospective study design, records of all trauma patients who reported to the Department of Oral and Maxillofacial Surgery, KLE's PK Hospital Belgaum, between the years January 2006 to October 2011 was obtained from the medical records office. The data variables that were analyzed were the name, age, sex, cause of injury, pretreatment occlusion, treatment given, period of MMF and post-treatment occlusion. Total number of mandibular fracture cases was 298. Ramus fractures were 10 in number which accounted for 3.3% of fractures. The age range of these 10 patients was seen to be between 20 to 80 years with the average age being 35.6 years. Of these 10 patients, 9 were male and 1 was female and 7 patients were treated by open reduction and internal fixation and the remaining 3 by closed reduction. The average period of MMF was 3 days for the patients who underwent open reduction and internal fixation. There was improvement in occlusion in all 10 patients post-treatment and there was no complication reported in any of the cases. Ramus fractures accounted for 3.3% of all mandibular fractures. Open reduction and internal fixation of ramus fractures ensures adequate functional and anatomic reduction. This study makes an attempt to throw a light on the increasing incidence of ramus fractures and a successful management of these fractures by open reduction and internal fixation. How to cite this article: Kale TP, Kotrashetti SM, Louis A, Lingaraj JB, Sarvesh BU. Mandibular Ramus Fractures: A Rarity. J Contemp Dent Pract 2013;14(1):39-42. Source of support: Nil Conflict of interest: None declared.

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

Dictionary of pressure vessel and piping technology

International Nuclear Information System (INIS)

Schmitz, H.P.

1987-01-01

This dictionary is the result of many years of evaluation of technical terminology taken from the salient non-German rules, regulations, standards and specifications such as ANSI, API, ASME, ASNT, ASTM, BSI, EJMA, TEMA, and WRC (see bibliography) and of comparing these with the corresponding German rules, regulations, etc., as well as examining relevant technical documentation. This dictionary fills the gap left by existing dictionaries. The following specialized factors are given special attention: pressure vessels, tanks, heat exchangers, piping, valves and fittings, expansion joints, flanges, giving particular consideration to the fields of materials, welding, strength calculation, design and construction, fracture mechanics, destructive and non-destructive testing, as well as heat and mass transfer. (orig./HP) [de

Fracture strengths of chair‑side‑generated veneers cemented with ...

African Journals Online (AJOL)

Introduction: CAD/CAM (computer‑aided design and computer‑aided manufacturing) systems have refreshed the idea of chair‑side production of restorations, but the fracture of ceramic veneers remains a problem. Cementation with glass fibers may improve the fracture strengths and affect the failure modes of ...

A service laboratory's view of the status and direction of reactor vessel surveillance

International Nuclear Information System (INIS)

Norris, E.B.

1981-01-01

Advances in testing techniques and analysis procedures have had a minor impact to date on the conduct of reactor vessel material surveillance programs. However, major thrusts in the near future will be associated with the development of elastic-plastic fracture toughness data on irradiated materials and improvements in analysis techniques for projecting surveillance results to the pressure vessel wall. In this regard, increased emphasis will be placed on the development of R-curves from the results of J-integral tests. Also, efforts will be increased to develop a better understanding of neutron irradiation embrittlement mechanisms, to determine if a time dependency of damage can lead to saturation and to evaluate the significance of small variations in irradiation temperature on the embrittlement response

Ipsilateral humeral neck and shaft fractures

Directory of Open Access Journals (Sweden)

Zhu Bin

2017-01-01

Full Text Available Background/Aim. Fractures of the proximal humerus or shaft are common, however, ipsilateral neck and shaft humerus fracture is a rare phenomenon. This combination injury is challenging for orthopaedic surgeons because of its complex treatment options at present. The purpose of this study was to review a series of ipsilateral humeral neck and shaft fractures to study the fracture pattern, complications and treatment outcomes of each treatment options used. Methods. A total of six patients (four female and two male with the average age of 42.8 years (range: 36–49 years was collected and reviewed retrospectively. Two of them were treated with double plates and four with antegrade intramedullary nail. According to the Neer’s classification, all proximal fractures were two-part surgical neck fractures. All humeral shaft fractures were located at the middle of one third. Five fractures were simple transverse (A3, one fragmented wedge fracture (B3. One patient had associated radial nerve palsy. Results. All surgical neck fractures except one united uneventfully in the average time span of 8.7 weeks. Four humeral shaft fractures healed in near anatomic alignment. The remaining two patients had the nonunion with no radiological signs of fracture healing. The average University of California, Los Angeles End-Results (UCLA score was 23.1. On the contrary, the average American Shoulder and Elbow Surgeon's (ASES score was 73.3. The patients treated with antegrade intramedullary nails presented 70.5 points. The ASES scores were 79 in the double plates group. Conclusions. Ipsilateral humeral shaft and neck fracture is extremely rare. Both antegrade intramedullar nailing and double plates result in healing of fractures. However the risk of complication is lower in the double plating group.

Lower Length Scale Model Development for Embrittlement of Reactor Presure Vessel Steel

Energy Technology Data Exchange (ETDEWEB)

Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, Daniel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Chakraborty, Pritam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, Xianming [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

This report summarizes the lower-length-scale effort during FY 2016 in developing mesoscale capabilities for microstructure evolution, plasticity and fracture in reactor pressure vessel steels. During operation, reactor pressure vessels are subject to hardening and embrittlement caused by irradiation induced defect accumulation and irradiation enhanced solute precipitation. Both defect production and solute precipitation start from the atomic scale, and manifest their eventual effects as degradation in engineering scale properties. To predict the property degradation, multiscale modeling and simulation are needed to deal with the microstructure evolution, and to link the microstructure feature to material properties. In this report, the development of mesoscale capabilities for defect accumulation and solute precipitation are summarized. A crystal plasticity model to capture defect-dislocation interaction and a damage model for cleavage micro-crack propagation is also provided.

Greater trochanteric fracture with occult intertrochanteric extension.

Science.gov (United States)

Reiter, Michael; O'Brien, Seth D; Bui-Mansfield, Liem T; Alderete, Joseph

2013-10-01

Proximal femoral fractures are frequently encountered in the emergency department (ED). Prompt diagnosis is paramount as delay will exacerbate the already poor outcomes associated with these injuries. In cases where radiography is negative but clinical suspicion remains high, magnetic resonance imaging (MRI) is the study of choice as it has the capability to depict fractures which are occult on other imaging modalities. Awareness of a particular subset of proximal femoral fractures, namely greater trochanteric fractures, is vital for both radiologists and clinicians since it has been well documented that they invariably have an intertrochanteric component which may require surgical management. The detection of intertrochanteric or cervical extension of greater trochanteric fractures has been described utilizing MRI but is underestimated with both computed tomography (CT) and bone scan. Therefore, if MRI is unavailable or contraindicated, the diagnosis of an isolated greater trochanteric fracture should be met with caution. The importance of avoiding this potential pitfall is demonstrated in the following case of an elderly woman with hip pain and CT demonstrating an isolated greater trochanteric fracture who subsequently returned to the ED with a displaced intertrochanteric fracture.

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.

[Arthroscopy-guided fracture management. Ankle joint and calcaneus].

Science.gov (United States)

Schoepp, C; Rixen, D

2013-04-01

Arthroscopic fracture management of the ankle and calcaneus requires a differentiated approach. The aim is to minimize surgical soft tissue damage and to visualize anatomical fracture reduction arthroscopically. Moreover, additional cartilage damage can be detected and treated. The arthroscopic approach is limited by deep impressions of the joint surface needing cancellous bone grafting, by multiple fracture lines on the articular side and by high-grade soft tissue damage. An alternative to the minimally invasive arthroscopic approach is open arthroscopic reduction in conventional osteosynthesis. This facilitates correct assessment of surgical reduction of complex calcaneal fractures, otherwise remaining non-anatomical reduction might not be fluoroscopically detected during surgery.

Discontinuous finite element formulation for bodies of revolution with application in the prevention of fragile fracture in pressure vessel of PWR reactors; Formulacao de elementos finitos descontinuos para corpos de revolucao com aplicacao na prevencao de fratura fragil em vaso de pressao de reatores PWR

Energy Technology Data Exchange (ETDEWEB)

Benitez Alvarez, Gustavo

1999-08-15

In this work, a hybrid formulation is established for bodies of revolution, based on the equation of Fourier series for the discontinuous finite element method, analogous to the one that exists in the classical finite element method. Furthermore, a methodology to analyse the prevention of fragile fracture in pressure vessel of pressurized water reactors is presented. The results obtained suggest that careful analysis must be made for non symmetric refrigeration. (author)

Development of a statistically-based lower bound fracture toughness curve (Ksub(IR) curve)

International Nuclear Information System (INIS)

Wullaert, R.A.; Server, W.L.; Oldfield, W.; Stahlkopf, K.E.

1977-01-01

A program of initiation fracture toughness measurements on fifty heats of nuclear pressure vessel production materials (including weldments) was used to develop a methodology for establishing a revised reference toughness curve. The new methodology was statistically developed and provides a predefined confidence limit (or tolerance limit) for fracture toughness based upon many heats of a particular type of material. Overall reference curves were developed for seven specific materials using large specimen static and dynamic fracture toughness results. The heat-to-heat variation was removed by normalizing both the fracture toughness and temperature data with the precracked Charpy tanh curve coefficients for each particular heat. The variance and distribution about the curve were determined, and lower bounds of predetermined statistical significance were drawn based upon a Pearson distribution in the lower shelf region (since the data were skewed to high values) and a t-distribution in the transition temperature region (since the data were normally distributed)

A new method for improving the reliability of fracture toughness surveillance of nuclear pressure vessel by neutron irradiated embrittlement

International Nuclear Information System (INIS)

Zhang Xinping; Shi Yaowu

1992-01-01

In order to obtain more information from neutron irradiated sample specimens and raise the reliability of fracture toughness surveillance test, it has more important significance to repeatedly exploit the broken Charpy-size specimen which had been tested in surveillance test. In this work, on the renewing design and utilization for Charpy-size specimens, 9 data of fracture toughness can be gained from one pre-cracked side-grooved Charpy-size specimen while at the preset usually only 1 to 3 data of fracture toughness can be obtained from one Chharpy-size specimen. Thus, it is found that the new method would obviously improve the reliability of fracture toughness surveillance test and evaluation. Some factors which affect the reasonable design of pre-cracked deep side-groove Charpy-size compound specimen have been discussed

Hypercholesterolemia induced cerebral small vessel disease.

Science.gov (United States)

Kraft, Peter; Schuhmann, Michael K; Garz, Cornelia; Jandke, Solveig; Urlaub, Daniela; Mencl, Stine; Zernecke, Alma; Heinze, Hans-Jochen; Carare, Roxana O; Kleinschnitz, Christoph; Schreiber, Stefanie

2017-01-01

While hypercholesterolemia plays a causative role for the development of ischemic stroke in large vessels, its significance for cerebral small vessel disease (CSVD) remains unclear. We thus aimed to understand the detailed relationship between hypercholesterolemia and CSVD using the well described Ldlr-/- mouse model. We used Ldlr-/- mice (n = 16) and wild-type (WT) mice (n = 15) at the age of 6 and 12 months. Ldlr-/- mice develop high plasma cholesterol levels following a high fat diet. We analyzed cerebral capillaries and arterioles for intravascular erythrocyte accumulations, thrombotic vessel occlusions, blood-brain barrier (BBB) dysfunction and microbleeds. We found a significant increase in the number of erythrocyte stases in 6 months old Ldlr-/- mice compared to all other groups (P hypercholesterolemia is related to a thrombotic CSVD phenotype, which is different from hypertension-related CSVD that associates with a hemorrhagic CSVD phenotype. Our data demonstrate a relationship between hypercholesterolemia and the development of CSVD. Ldlr-/- mice appear to be an adequate animal model for research into CSVD.

Reactor pressure vessel embrittlement of NPP borssele: Design lifetime and lifetime extension

International Nuclear Information System (INIS)

Blom, F.J.

2007-01-01

Embrittlement of the reactor pressure vessel of the Borssele nuclear power plant has been investigated taking account of the design lifetime of 40 years and considering 20 years subsequent lifetime extension. The paper presents the current licensing status based on considerations of material test data and of US nuclear regulatory standards. Embrittlement status is also evaluated against German and French nuclear safety standards. Results from previous fracture toughness and Charpy tests are investigated by means of the Master curve toughness transition approach. Finally, state of the art insights are investigated by means of literature research. Regarding the embrittlement status of the reactor pressure vessel of Borssele nuclear power plant it is concluded that there is a profound basis for the current license up to the original end of the design life in 2013. The embrittlement temperature changes only slightly with respect to the acceptance criterion adopted postulating further operation up to 2033. Continued safe operation and further lifetime extension are therefore not restricted by reactor pressure vessel embrittlement

Reactive silica transport in fractured porous media: Analytical solutions for a system of parallel fractures

Science.gov (United States)

Yang, Jianwen

2012-04-01

A general analytical solution is derived by using the Laplace transformation to describe transient reactive silica transport in a conceptualized 2-D system involving a set of parallel fractures embedded in an impermeable host rock matrix, taking into account of hydrodynamic dispersion and advection of silica transport along the fractures, molecular diffusion from each fracture to the intervening rock matrix, and dissolution of quartz. A special analytical solution is also developed by ignoring the longitudinal hydrodynamic dispersion term but remaining other conditions the same. The general and special solutions are in the form of a double infinite integral and a single infinite integral, respectively, and can be evaluated using Gauss-Legendre quadrature technique. A simple criterion is developed to determine under what conditions the general analytical solution can be approximated by the special analytical solution. It is proved analytically that the general solution always lags behind the special solution, unless a dimensionless parameter is less than a critical value. Several illustrative calculations are undertaken to demonstrate the effect of fracture spacing, fracture aperture and fluid flow rate on silica transport. The analytical solutions developed here can serve as a benchmark to validate numerical models that simulate reactive mass transport in fractured porous media.

Clinical Utility of Chest Computed Tomography in Patients with Rib Fractures CT Chest and Rib Fractures.

Science.gov (United States)

Chapman, Brandon C; Overbey, Douglas M; Tesfalidet, Feven; Schramm, Kristofer; Stovall, Robert T; French, Andrew; Johnson, Jeffrey L; Burlew, Clay C; Barnett, Carlton; Moore, Ernest E; Pieracci, Fredric M

2016-12-01

Chest CT is more sensitive than a chest X-ray (CXR) in diagnosing rib fractures; however, the clinical significance of these fractures remains unclear. The purpose of this study was to determine the added diagnostic use of chest CT performed after CXR in patients with either known or suspected rib fractures secondary to blunt trauma. Retrospective cohort study of blunt trauma patients with rib fractures at a level I trauma center that had both a CXR and a CT chest. The CT finding of ≥ 3 additional fractures in patients with ≤ 3 rib fractures on CXR was considered clinically meaningful. Student's t-test and chi-square analysis were used for comparison. We identified 499 patients with rib fractures: 93 (18.6%) had CXR only, 7 (1.4%) had chest CT only, and 399 (79.9%) had both CXR and chest CT. Among these 399 patients, a total of 1,969 rib fractures were identified: 1,467 (74.5%) were missed by CXR. The median number of additional fractures identified by CT was 3 (range, 4 - 15). Of 212 (53.1%) patients with a clinically meaningful increase in the number of fractures, 68 patients underwent one or more clinical interventions: 36 SICU admissions, 20 pain catheter placements, 23 epidural placements, and 3 SSRF. Additionally, 70 patients had a chest tube placed for retained hemothorax or occult pneumothorax. Overall, 138 patients (34.5%) had a change in clinical management based upon CT chest. The chest X-ray missed ~75% of rib fractures seen on chest CT. Although patients with a clinical meaningful increase in the number of rib fractures were more likely to be admitted to the intensive care unit, there was no associated improvement in pulmonary outcomes.

Confinement Vessel Assay System: Design and Implementation Report

International Nuclear Information System (INIS)

Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Mayo, Douglas R.; Gomez, Cipriano D.; 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 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. We have developed a neutron assay system for the purposes of Materials Control and Accountability (MC and A) measurements of the vessel prior to and after cleanout. We present our approach to confronting the challenges in designing, building, and testing such a system. The system was designed to meet a set of functional and operational requirements. A Monte Carlo model was developed to aid in optimizing the detector design as well as to predict the systematic uncertainty associated with confinement vessel measurements. Initial testing was performed to optimize and determine various measurement parameters, and then the system was characterized using 252 Cf placed a various locations throughout the measurement system. Measurements were also performed with a 252 Cf source placed inside of small steel and HDPE shells to study the effect of moderation. These measurements compare favorably with their MCNPX model equivalent, making us confident that we can rely on the Monte Carlo simulation to predict the systematic uncertainty due to variations in response to material that may be localized at different points within a vessel.

Ultrasonographic diagnosis of pancreatic fracture

International Nuclear Information System (INIS)

Woo, Seong Ku; Lim, Jae Hoon; Ko, Young Tae; Choi, Yong Dae; Yoon, Yup; Kim, Soo yong

1983-01-01

The pancreatic fracture, known also as complete transection of the pancreas, is severe injury of the pancreas characterized by complete vertical transection of its body overlying the body of the vertebra. The authors diagnosed three cases of traumatic fracture of the pancreas by ultrasonography and these were confirmed surgically. Ultrasonography disclosed an anechoic fluid collection between the separated parenchyma of the body of the pancreas anterior to the superior mesenteric artery. The remaining pancreas enlarged diffusely and decreased in echogenecity. Associated feature was accumulation of fluid in the lesser sac and the peritoneal cavity

Case report: clinical and postmortem findings in four cows with rib fracture

OpenAIRE

Braun, Ueli; Warislohner, Sonja; Hetzel, Udo; Nuss, Karl

2017-01-01

BACKGROUND: Published reports of rib fractures in adult cattle are limited to the occurrence of chronic rib swellings caused by calluses, which are unremarkable from a clinical standpoint, whereas studies identifying clinical signs of rib fractures were not found in a literature search. This report describes the clinical and postmortem findings in four cows with rib fractures. CASE PRESENTATION: The 13th rib was fractured in three cows and the 11th rib in the remaining cow; three fractures...

Anomalous fracture toughness of irradiated Cr-MoV - Reactor pressure vessel steel

Energy Technology Data Exchange (ETDEWEB)

Ahistrand, R [Imatran Voima Oy (IVO), Helsinki (Finland)

1994-12-31

The base metal Crack Opening Displacement (COD) specimens of the irradiation-induced embrittlement surveillance programme in Loviisa 1 revealed an anomalous behaviour of K{sub JC} compared to the Charpy-V results and to expected results according to standards: about 20% of the COD specimens showed an exceptionally low fracture toughness. Abnormal test specimens were analyzed through fractography, metallography and repeated tests using reconstitution technique: the anomalous behaviour appears to be caused by incorrect pre-fatigue cracking of base metal COD specimens. 7 refs., 9 figs.

Safe surgical technique: intramedullary nail fixation of tibial shaft fractures.

Science.gov (United States)

Zelle, Boris A; Boni, Guilherme

2015-01-01

Statically locked, reamed intramedullary nailing remains the standard treatment for displaced tibial shaft fractures. Establishing an appropriate starting point is a crucial part of the surgical procedure. Recently, suprapatellar nailing in the semi-extended position has been suggested as a safe and effective surgical technique. Numerous reduction techiques are available to achieve an anatomic fracture alignment and the treating surgeon should be familiar with these maneuvers. Open reduction techniques should be considered if anatomic fracture alignment cannot be achieved by closed means. Favorable union rates above 90 % can be achieved by both reamed and unreamed intramedullary nailing. Despite favorable union rates, patients continue to have functional long-term impairments. In particular, anterior knee pain remains a common complaint following intramedullary tibial nailing. Malrotation remains a commonly reported complication after tibial nailing. The effect of postoperative tibial malalignment on the clinical and radiographic outcome requires further investigation.

Pediatric facial fractures: evolving patterns of treatment.

Science.gov (United States)

Posnick, J C; Wells, M; Pron, G E

1993-08-01

This study reviews the treatment of facial trauma between October 1986 and December 1990 at a major pediatric referral center. The mechanism of injury, location and pattern of facial fractures, pattern of facial injury, soft tissue injuries, and any associated injuries to other organ systems were recorded, and fracture management and perioperative complications reviewed. The study population consisted of 137 patients who sustained 318 facial fractures. Eighty-one patients (171 fractures) were seen in the acute stage, and 56 patients (147 fractures) were seen for reconstruction of a secondary deformity. Injuries in boys were more prevalent than in girls (63% versus 37%), and the 6- to 12-year cohort made up the largest group (42%). Most fractures resulted from traffic-related accidents (50%), falls (23%), or sports-related injuries (15%). Mandibular (34%) and orbital fractures (23%) predominated; fewer midfacial fractures (7%) were sustained than would be expected in a similar adult population. Three quarters of the patients with acute fractures required operative intervention. Closed reduction techniques with maxillomandibular fixation were frequently chosen for mandibular condyle fractures and open reduction techniques (35%) for other regions of the facial skeleton. When open reduction was indicated, plate-and-screw fixation was the preferred method of stabilization (65%). The long-term effects of the injuries and the treatment given on facial growth remain undetermined. Perioperative complication rates directly related to the surgery were low.

Significance of reheat cracks to the integrity of pressure vessels for light-water reactors

International Nuclear Information System (INIS)

Canonico, D.A.

1977-01-01

Reheat cracks usually manifest themselves as macroscopic defects, which are centimeters long and deep, and are detectable by the usual nondestructive examination (NDE) procedures or as microscopic grain boundary decohesions (GBD) that are beyond the limit of detection by commercial NDE procedures. This report has concentrated on the significance of the microscopic cracks that may go undetected. The probability that GBD exist in the heat-affected zones (HAZ) of weldments of pressure vessel steels is high; particularly in SA 508 Class 2 weldments. A sample of the HAZ from the prolongation-weldment from the Heavy Section Steel Technology program Intermediate Test Vessel (ITV) No. 4 was examined by the Staatliche Materialprufungsanstalt (MPA). They reported GBD 5 mm (0.2 in.) long. This prompted an examination of the HAZ from the ITV vessel that had been tested to failure at 24 0 C (75 0 F). During testing, the region of the weld which contained the flaw that initiated the failure was strained up to 0.5%. A metallographic examination of this region of the weldment revealed GBD, but none of the size reported by the MPA. Further, there was no evidence that the GBD had extended as a consequence of the tests. Fracture toughness tests were made of the HAZ of welds from ITV-4. The electron-beam welding procedure, which permits more accurate siting of the crack, was used. Fracture toughness values in excess of 220 MPa root m (200 ksi root in.) were obtained at -18 0 C

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

International Nuclear Information System (INIS)

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

1978-03-01

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

Modified tension band wiring of medial malleolar ankle fractures.

Science.gov (United States)

Georgiadis, G M; White, D B

1995-02-01

Twenty-two displaced medial malleolar ankle fractures that were treated surgically using the modified tension band method of Cleak and Dawson were retrospectively reviewed at an average follow-up of 25 months. The technique involves the use of a screw to anchor a figure-of-eight wire. There were no malreductions and all fractures healed. Problems with the technique included technical errors with hardware placement, medial ankle pain, and asymptomatic wire migration. Despite this, modified tension band wiring remains an acceptable method for fixation of selected displaced medial malleolar fractures. It is especially suited for small fracture fragments and osteoporotic bone.

Analysis of the necessity for inserting new surveillance capsule into the Kori Unit 1 RPV to monitor material fracture toughness

International Nuclear Information System (INIS)

Song, Taek Ho

2007-01-01

In association with monitoring of reactor pressure vessel (RPV) fracture toughness, surveillance capsule test specimens have been used to monitor the material property of nuclear reactor vessel. As far as Kori Unit 1 is concerned, 6 capsules were put into the vessel before commercial operation of the plant. Up to now, all the six capsules have been withdrawn to test and monitor the fracture toughness of RPV material. The last capsule has been withdrawn on June this year, and the Kori unit 1 has been shut downed since July 2007 and will be shut downed until December this year for about 6 months, preparing the life extension of the plant to operate the plant 10 more years. With the situation that all the surveillance capsules have been withdrawn, public ask the following question, 'To extend the life of Kori Unit 1 more than 10 years, is it necessary to insert new surveillance capsules into the Kori Unit 1 to monitor RPV material fracture toughness?' In connection with this issue, planning project have been carried out since spring this year. In this paper, it is described that inserting new surveillance capsule into the Kori Unit 1 RPV has some meaning in some public acceptance point of view and is not necessary in material engineering point of view

Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [ORNL

2010-08-01

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

Status report on the use of the CRB for the measurement of fracture toughness of RPV steels

International Nuclear Information System (INIS)

Scibetta, M.; Chaouadi, R.; Van Walle, E.

1998-02-01

A large number of fracture toughness tests were performed in order to assess the use of the circumferentially-Cracked Round Bar (CRB) as a potential method for the measurement of fracture toughness of Reactor Pressure Vessel steels. Test conditions were selected to: (1) characterise fracture toughness in the transition region; (2) study the size effect and loss of constraint; (3) establish the limit of validity of this geometry; (4) investigate the ductile fracture at the upper shelf. In the transition region, the fracture toughness obtained from the CRB over-estimates the actual value as long as the loss of constraint and size effect were not taken into account. In addition, the B1/4 size correction is verified and gives a very good description of the size effect. The application of these corrections allows a good prediction of the normalised fracture toughness up to high levels of fracture toughness.In the upper shelf region, promising results were obtained with this geometry to characterise the ductile crack initiation and propagation

Summary of Reported Whale-Vessel Collisions in Alaskan Waters

Directory of Open Access Journals (Sweden)

Janet L. Neilson

2012-01-01

Full Text Available Here we summarize 108 reported whale-vessel collisions in Alaska from 1978–2011, of which 25 are known to have resulted in the whale's death. We found 89 definite and 19 possible/probable strikes based on standard criteria we created for this study. Most strikes involved humpback whales (86% with six other species documented. Small vessel strikes were most common (<15 m, 60%, but medium (15–79 m, 27% and large (≥80 m, 13% vessels also struck whales. Among the 25 mortalities, vessel length was known in seven cases (190–294 m and vessel speed was known in three cases (12–19 kn. In 36 cases, human injury or property damage resulted from the collision, and at least 15 people were thrown into the water. In 15 cases humpback whales struck anchored or drifting vessels, suggesting the whales did not detect the vessels. Documenting collisions in Alaska will remain challenging due to remoteness and resource limitations. For a better understanding of the factors contributing to lethal collisions, we recommend (1 systematic documentation of collisions, including vessel size and speed; (2 greater efforts to necropsy stranded whales; (3 using experienced teams focused on determining cause of death; (4 using standard criteria for validating collision reports, such as those presented in this paper.

Obturator Artery Injury Resulting in Massive Hemorrhage From a Low-Energy Pubic Ramus Fracture.

Science.gov (United States)

Solarz, Mark K; Kistler, Justin M; Rehman, Saqib

2017-05-01

Pelvic ring fractures are common in the elderly population and are usually a result of low-energy trauma, such as falls from standing. In most cases, low-energy pelvic ring injuries can be treated with appropriate analgesia and early mobilization. Arterial injury resulting in hemodynamic instability from a low-energy pelvic ring injury is rare but, given the poor compliance of vessels in the elderly population, possible. These patients must be carefully monitored after the initial injury. The purpose of this report is to describe an elderly patient who sustained a superior pubic ramus fracture and arterial injury following a low-energy fall from standing that required angiographic intervention. Elderly patients who sustain low-energy or pelvic insufficiency fractures are unlike the younger population with high-energy pelvic fractures and hemodynamic collapse. Elderly patients can have a delayed presentation of arterial injury and require careful physical examination and close monitoring. Additionally, the authors provide a review of the literature for low-energy pelvic fractures. [Orthopedics. 2017; 40(3):e546-e548.]. Copyright 2017, SLACK Incorporated.

The pattern and prevalence of vertebral artery injury in patients with cervical spine fractures

Directory of Open Access Journals (Sweden)

Farzanah Ismail

2013-06-01

Method: A retrospective review of patients who had undergone CTA of the vertebral arteries was undertaken. Reports were reviewed to determine which patients met the inclusion criteria of having had both cervical spine fractures and CTA of the vertebral arteries. Images of patients who met the inclusion criteria were analysed by a radiologist. Results: The prevalence of vertebral artery injury was 33%. Four out of the 11 patients who had vertebral artery injury, had post-traumatic spasm of the artery, with associated thrombosis or occlusion of the vessel. In terms of blunt carotid vertebral injury (BCVI grading, most of the patients sustained grade IV injuries. Four patients who had vertebral artery injury had fractures of the upper cervical vertebrae, i.e. C1 to C3. Fifteen transverse process fractures were associated with vertebral artery injury. No vertebral artery injury was detected in patients who had facet joint subluxations. Conclusion: Patients with transverse process fractures of the cervical spine and upper cervical vertebral body fractures should undergo CTA to exclude vertebral artery injury.

Overview and Evaluation of the NESC Projects for Fracture Assessments of Nuclear Components